Wylfa Newydd Project Radioactive Substances Regulation Environmental Permit Application

Transcription

1 Wylfa Newydd Project Radioactive Substances Regulation Environmental Permit Application

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3 Contents 1 Introduction General Introduction Horizon (The Applicant) Summary of the Wylfa Project Scope of this Application Other Applications and Submissions Development Consent Order Regulatory Justification Generic Design Assessment Nuclear Site Licence Environmental Permits Article Document Structure Cross-reference of Application Content with NRW Requirements Abbreviations The Power Station UK ABWR Basic Function Main Structures at the Power Station relevant to the Application The Reactor Building The Turbine Building The Control Building Cooling Water System Heat Exchanger Building Filter Vent Building Radioactive Waste Building Lower Activity Waste Management Facility Service Building ILW Storage Facility and Spent Fuel Storage Facility Radioactive Drain Transfer System and Discharge Tunnels Environmental Survey Laboratory Development of the Reference Design Reference Design for Wylfa Newydd Location and Setting Page i of xii

4 2.4.1 Proposed Boundary of the Permitted Premises Environmental Setting Wylfa Newydd Project Lifecycle Phases Development Construction Commissioning Generation Decommissioning Radioactive Substances Activities at the Power Station Sources of Radioactivity Corrosion Products Fission Products Noble Gases Halogens: Isotopes of Iodine Soluble and Insoluble Fission Products Activation Products Tritium Argon Carbon Nitrogen Chlorine Systems Associated with Radioactive Waste Gaseous Radioactive Waste Management Systems Heating, Ventilation and Air Conditioning System Off-Gas System Liquid Radioactive Waste Management Systems Reactor Water Clean-up System Fuel Pool Cooling and Clean-up System Suppression Pool Clean-up System Condensate Water Clean-up System LCW, HCW and CAD Systems Solid Radioactive Waste Management System Lower Activity Waste Management Facility Wet-solid LLW Processing System Wet-solid ILW Processing System Page ii of xii

5 ILW Storage Facility Spent Fuel Storage Facility Summary of Solid Radioactive Wastes and Treatment Radioactive Waste Discharge Points and Disposal Routes Gaseous Radioactive Waste Liquid Radioactive Waste Solid Radioactive Waste Off-site Transfer of Waste LLWR Acceptance in Principle Disposability of ILW, HLW and Spent Fuel Variation of Discharges During Reactor Cycle Stages Operating Techniques Disposal of Radioactive Waste Source Term Data Source Terms Radionuclides Considered Expected Event Identification of Representative Event Fuel Pin Failure Quantification of Gaseous and Aqueous Discharges Gaseous Discharges Discharge from a Single UK ABWR over a Rolling 12 Month Period Discharge from a Single UK ABWR for a Single Expected Event Discharge from the Rw/B over a 12 Month Period Discharge from the Site-specific Waste Facilities Best Estimate of the Total 12 Month Rolling Gaseous Discharges Aqueous Discharges Best Estimate of the Total 12 Month Rolling Aqueous Discharges Setting Discharge Limits and Quarterly Notification Levels Discharge Limits Headroom Factors Quarterly Notification Levels Discharge data for the Human Dose Assessment Identification of Significant Radionuclides Significant Radionuclides Summary of Discharge Estimates, Limits and QNLs Page iii of xii

6 5.7 Comparison of Power Station Discharge Data with Operational Experience Page iv of xii Gaseous Discharges Noble Gases (including Argon-41) Carbon Tritium Iodine Aqueous Discharges Tritium Solid Waste Radioactive Waste Arisings Immediately following Shutdown Radioactive Waste Arisings during the Normal Operations Phase Solid Intermediate Level Radioactive Waste Solid Low Level Radioactive Waste Estimate of Solid Radioactive Waste Arisings from Decommissioning Monitoring Objectives, Standards and Guidance Objectives of the Sampling and Monitoring Systems Standards and Guidance EU BS ISO 2889: ISO 10780: BS EN :2004 and BS EN : MCERTS BS EN ISO/IEC 17025:2005 and ISO 11929: Radioactive Substances Regulation Environmental Principles Technical Guidance Notes In-process Monitoring Gaseous Radioactive Effluent In-process Sampling and Monitoring Off-Gas System HVAC System Turbine Gland Steam System Off-gas and Mechanical Vacuum Pump Exhaust Liquid Radioactive Effluent In-process Sampling and Monitoring HCW Sampling and Monitoring Discharge Monitoring

7 6.3.1 Gaseous Discharge Monitoring R/B Stack Other Discharge Sources Aqueous Discharge Monitoring Flow Contamination Sampling Radiation Detection Instrumentation Solid Radioactive Waste Monitoring and Characterisation Independent Sampling Gaseous Sampling Aqueous Sampling Environmental Monitoring Guidance on the Design of Environmental Monitoring Programmes Baseline Radiological Monitoring Monitoring Undertaken by the Operators of the Existing Power Station Magnox Terrestrial Monitoring Programme Magnox Marine Monitoring Programme Radioactivity in Food and the Environment (RIFE) Environmental Monitoring for the Power Station Collate Site Information and Assess Site Impact Establish Objectives Determine What to Monitor, Where, How and How Often Determine Analysis and Reporting Requirements Communicate Results and Maintain Records Radiological Assessment Human Dose Assessment Guidance Dose Limit and Dose Constraints Assessment of Doses due to Planned Continuous Discharge Methodology Input Data Habits Data and Assumptions Results Assessment of Doses due to Planned Short-term Discharges Page v of xii

8 Methodology Input Data Results Assessment of Doses due to Direct Radiation Methodology Input Data Habits Data and Assumptions Results Representative Person and Compliance with Source Constraint Assessment of Collective Dose to Populations Methodology Habits Data and Assumptions Results Assessment of Build-up of Radionuclides in the Environment Methodology Land Uses and Likely Future Changes Results Assessment of Total Dose Methodology Results Uncertainty and Variability within the Public Dose Assessment Conclusions on the Human Dose Assessment Non-human Dose Assessment Guidance Methodology and Data Inputs Assessment Approach Site Layout and Surrounding Habitats Calculation of Radionuclide Concentration in the Environment Representative Organisms Assessed Results Terrestrial Habitat Marine Habitat Freshwater Habitat Combined Impacts Results and Conclusions Page vi of xii

9 7.2.5 Uncertainty and Variability in the Non-human Biota Dose Assessment Management Arrangements Management Prospectus and Company Manual EP-RSR Compliance Compliance Requirements Additional Guidance and Arrangements Horizon s Approach to Developing Compliance Development Phase Construction Phase Commissioning Phase Generation Phase Decommissioning Phase Development of Key Compliance Areas Compliance Management Organisational Readiness The Nuclear Baseline Learning Experience Records Radioactive Waste Adviser Control of Operations Management of Design Managing the EPC Contract Maintenance Waste Management Monitoring Arrangements Notifications and Reporting Assurance Quality Assurance Incidents and Emergency Response Decommissioning Forward Work Plan References Page vii of xii

10 List of Tables Table 1.1 List of Sections in this EP-RSR Application... 8 Table 1.2 Signposting of NRW Requirements within the Application... 9 Table 2.1 Main Buildings and Structures on the Power Station Site Table 3.1 Generation Mechanism for Significant Corrosion Products Table 3.2 Generation Mechanism for Significant FP and ActP Table 3.3 Decay Schemes of FP Noble Gases to 4 th Generation Daughters Table 3.4 Decay Scheme of Halogens to Noble Gas Daughters Table 3.5 Mechanisms for Production of C Table 3.6 Radioactive Waste and Spent Fuel Streams Arising from the Power Station Table 3.7 R/B Stack Characteristics and Discharge Parameters Table 3.8 Gaseous Discharge Outlets Reference Numbers Table 5.1 List of Radionuclides Considered for the UK ABWR s Gaseous Release Table 5.2 Additional Radionuclides Potentially arising from the Site Waste Facilities Table 5.3 List of Radionuclides Considered for the UK ABWR s Aqueous Release Table 5.4 Monthly Gaseous Discharges from the OG System (D Rn,gas,OG) Table 5.5 Monthly Gaseous Discharges from the HVAC System (D Rn,gas,HVAC) Table 5.6 Monthly Gaseous Discharges from the TGS (D Rn,gas,TGS) Table 5.7 Annual Gaseous Discharge from a Single UK ABWR (D Rn,gas,GDA) Table 5.8 Gaseous Discharges for a Fuel Pin Failure Expected Event (D Rn,gas,EE) Table 5.9 Gaseous Discharges from the Rw/B (D Rn,gas,Rw/B) Table 5.10 Annual Gaseous Discharges for the Site Waste Facilities (D Rn,gas,SSwaste) Table 5.11 Best Estimate Total 12 Month Rolling Gaseous Discharges (D Rn,gas,total) Table 5.12 Annual Aqueous Discharges from the HCW Sample Tanks Table 5.13 Best Estimate Total 12 Month Rolling Aqueous Discharges (D Rn,liquid,total) Table 5.14 Headroom Factors by Radionuclide Group Table 5.15 Headroom Factors by Radionuclide Table 5.16 Annual Gaseous Radionuclide Limits used in the Dose Assessment Table 5.17 Annual Aqueous Discharge Limits used in the Dose Assessment Table 5.18 Basis for Establishing Radionuclide Groups for Discharge Limit Setting Table 5.19 Summary of the Gaseous Annual Estimates, Discharge Limits and QNLs Table 5.20 Summary of the Aqueous Annual Estimates, Discharge Limits and QNLs Table 5.21 Radioactive Waste Arisings during the Normal Operations Phase Table 5.22 Waste Stream Metadata Page viii of xii

11 Table 5.23 Radioactive Waste Arisings Spent Nuclear Fuel Table 5.24 Radioactive Waste Arisings Sludge (Crud) ILW Table 5.25 Radioactive Waste Arisings Powder Resin and Sludge (crud) ILW Table 5.26 Radioactive Waste Arisings Control Rods ILW Table 5.27 Radioactive Waste Arisings Reactor Components ILW Table 5.28 Radioactive Waste Arisings Bead Resin LLW Table 5.29 Radioactive Waste Arisings Concentrates (Sludge) LLW Table 5.30 Radioactive Waste Arisings HVAC Filters LLW Table 5.31 Radioactive Waste Arisings Spent Filter Media LLW Table 5.32 Radioactive Waste Arisings Heterogeneous LLW Table 5.33 Radioactive Waste Arisings Radioactive Oils and Oily Wastes Table 5.34 Radioactive Waste Arisings Radiologically Contaminated Land Table 5.35 Decommissioning Radioactive Waste Summary Table 6.1 LCW, HCW and CAD Sampling and Monitoring Locations Table 6.2 Items and Purpose for Each Monitoring and Sampling Point HCW System Table 6.3 Discharge Flow Characteristics of the Unit 1 Reactor Stack Table 6.4 Objectives of the Environmental Monitoring Programme for the Power Station 155 Table 7.1 Assessment Parameters for the Farming Family (Gaseous Discharges) Table 7.2 Habit Data for the Fishing Family (Aqueous Discharges) Table 7.3 Farming Family Habit Data (Gaseous and Aqueous Discharges) Table 7.4 Fishing Family Habit Data (Aqueous and Gaseous Discharges) Table 7.5 Habit Data for the Magnox Worker (Gaseous Discharges) Table 7.6 Dose to the Farming Family by Exposure Pathway (Gaseous Only) Table 7.7 Dose to the Fishing Family by Exposure Pathway (Aqueous Only) Table 7.8 Dose to the Farming Family by Exposure Pathway (Aqueous and Gaseous) Table 7.9 Dose to the Fishing Family by Exposure Pathway (Aqueous and Gaseous) Table 7.10 Dose to the Magnox Worker (Gaseous Discharges) by Exposure Pathway Table 7.11 Specified Receptor Locations Table 7.12 Estimated Doses at the Residential Location due to the Short-term Release Table 7.13 Estimated Doses Incurred by a Magnox Worker due to a Short-term Release. 179 Table 7.14 Structures Containing Radioactive Sources on Site Table 7.15 Annual Doses to Main Receptors Table 7.16 Annual Doses to Walkers Table 7.17 Annual Doses to Farming and Fishing Families Table 7.18 Annual Dose to the Candidates for the Representative Person Page ix of xii

12 Table 7.19 Collective Dose per Year of Discharge due to Gaseous Discharges Table 7.20 Collective Dose per Year of Discharge due to Aqueous Discharges Table 7.21 Table 7.22 Table 7.23 Table 7.24 Build-up of Activity in the Marine Environment over a 60-year Period (within the Local Compartment) Build-up of Activity in the Terrestrial Environment over a 60-year Period (at the Food Production Receptor) Measured Radionuclide Activity Concentrations in Seawater from OSPAR Monitoring Area Measured Sediment and Seawater Activity Concentrations for the Wylfa Area in 2014 and Table 7.25 Predicted Combined Dose Table 7.26 Dose Contributions to the RIFE Representative Person Table 7.27 Summary of the Total Doses for the Power Station Table 7.28 Local European Designated Sites with Marine, Terrestrial and Freshwater Qualifying Features Table 7.29 Authorised Limits for the Existing Power Station Table 7.30 SRS-19 Parameter Values for Small Lake Table 7.31 Reference Organisms Modelled for each Proxy Habitat Table 7.32 Terrestrial Habitat ERICA Results (Dose Rate and Risk Quotient) Table 7.33 Terrestrial Habitat R&D128 Results (Dose Rate and Risk Quotient) Table 7.34 Combined (Total) Terrestrial Habitat Results (Dose Rate and Risk Quotient) Table 7.35 Marine Habitat ERICA Results (Dose Rate and Risk Quotient) Table 7.36 Freshwater Habitat ERICA Results (Dose Rate and Risk Quotient) Table 7.37 In-combination Total Dose Rates for European Designated Sites in the Vicinity of the Power Station Table 9.1 Summary of Forward Work Plan Commitments Table 10.1 Schedule of References Page x of xii

13 List of Figures Figure 1.1 Overview of Corporate Structure... 2 Figure 2.1 Schematic of the UK ABWR Figure 2.2 Layout of the Power Station Site Figure 2.3 Cooling Water Outfall Location Figure 2.4 Location of Wylfa Newydd Site Figure 2.5 Proposed Boundary of the Permitted Premises Figure 2.6 Geographic Areas of the Project Figure 3.1 Primary Pathway from the RPV to the Downstream Systems Figure 3.2 Overview of Radioactive Waste Management Processes Figure 3.3 Summary of Gaseous Effluent Generation, Treatment and Disposal Figure 3.4 Process Diagram for the Off-Gas System Figure 3.5 Outline of the Reactor Water Clean-up System Figure 3.6 Outline of the Fuel Pool Cooling and Clean-up System Figure 3.7 Outline of the Suppression Pool Clean-up System Figure 3.8 Outline of Condensate Water Clean-Up System Figure 3.9 Overview of the LCW, HCW and CAD Systems Figure 3.10 Outline of the LCW System Figure 3.11 Outline of the HCW System Figure 3.12 Outline of Waste Stream for Wet-Solid Waste Figure 3.13 Radioactive Waste Management Strategies and Waste Hierarchy Figure 3.14 Radioactive Discharge Routes during Operation, Start-up and Shutdown Figure 3.15 Radioactive Discharge Routes to the Environment during an Outage Figure 5.1 Discharges during Normal Operations and Expected Events Figure 5.2 GNF Historical Fuel Reliability Performance Figure 5.3 Power Station Gaseous Noble Gas and Ar-41 Discharge vs BWRs Figure 5.4 Power Station Gaseous Carbon-14 Discharge vs BWRs Figure 5.5 Power Station Gaseous Tritium Discharge vs BWRs Figure 5.6 Power Station Gaseous Iodine Isotope Discharge vs BWRs Figure 5.7 Power Station Aqueous Tritium Discharge vs BWRs Figure 6.1 Overview of Monitoring Locations for Gaseous Effluent Figure 6.2 Outline of HCW and Monitoring/Sampling Points Figure 6.3 Sketch of the Reactor Stack Figure 6.4 Stack Sampling System Configuration Page xi of xii

14 Figure 6.5 Aqueous Effluent Monitoring Locations Figure 7.1 Relative Positions of Buildings and Receptors Figure 7.2 Walker Exposure Locations for Direct Radiation Assessment Figure 7.3 Illustration of the OSPAR Monitoring Areas Figure 7.4 RIFE Total Historical Dose Methodology Figure 9.1 Indicative Programme for the Delivery of the Forward Work Plan Commitments Page xii of xii

15 1 Introduction 1.1 General Introduction 1. This document is the technical submission which has been produced in support of an Application for an Environmental Permit for the disposal of radioactive waste (hereafter referred to as EP-RSR) from a new nuclear power station (Wylfa Newydd Power Station hereafter referred to as the Power Station ) which will be built on the north coast of Anglesey. During the operation of the Power Station radioactivity will be disposed of in a controlled and strictly regulated way to the atmosphere, to the sea, and to appropriately permitted locations away from the site. The EP-RSR will allow these disposals to be made. 2. This document has been prepared in line with Schedule 23 of the Environmental Permitting (England and Wales) Regulations 2016 (hereafter referred to as EPR16) [RD1], known as the Radioactive Substances Regulation (RSR), and is being submitted to Natural Resources Wales (NRW) in order to supplement the information provided in completed application forms RSR-A, RSR-B3 and RSR-F (the completed forms are presented in Appendix A). 3. This Application describes the Wylfa Newydd Project (the Project), the nuclear reactor technology which will be employed, and the Power Station s principal plant and systems. It identifies the sources of radioactivity, and the techniques that will be adopted to handle and treat gaseous, aqueous and solid radioactive waste. It demonstrates how the design of the Power Station will optimise the protection of people and the environment (through the use of Best Available Techniques (BAT)), and minimise the volume and activity of radioactive waste arising (in accordance with the UK Strategy for Radioactive Discharges, 2009 [RD2]). 4. This Application also proposes limits for the discharge of gaseous and aqueous wastes to the environment, and sets out how these releases will be monitored to ensure on-going compliance with the EP-RSR conditions. It presents an assessment of the potential impact on human health and the environment based upon discharges at these limits, and gives an overview of the arrangements which will be employed at the Power Station to manage the radioactive substance activities. Finally, it presents a Forward Work Plan (FWP) to address those areas where further work is still required. Where the requirements for further work are first introduced, they are identified with the tag [FA RSR-1], [FA RSR-2], and so on. The requirements are reproduced in Section Horizon (the applicant) recognises that there will be design modifications throughout the development of the Project. The grant of an EP-RSR in response to this Application does not of itself foreclose on these design changes, nor does it fix the technology described hereunder. This is because the demonstration of BAT is a continuous process. Only in the case where a significant change in the permitted discharges is envisaged would Horizon address this by applying for a variation to the granted EP-RSR. 6. This document contains information about the United Kingdom Advanced Boiling Water Reactor (UK ABWR) technology which has been taken from documentation produced by Hitachi-GE Nuclear Energy Limited (HGNE) for the Generic Design Assessment (GDA) process (described in Section 1.5.3). Where the GDA information has been reproduced it has been highlighted with grey shading. Page 1 of 255

16 1.2 Horizon (The Applicant) 7. This Application is being made by Horizon Nuclear Power Wylfa Limited, the company responsible for developing and subsequently operating the Power Station. It will develop the Project using technology purchased from HGNE, a joint venture between Hitachi Limited and General Electric Corporation. 8. Horizon Nuclear Power Wylfa Limited (company registration number ) is headquartered at Sunrise House, 1420 Charlton Court, Gloucester Business Park, Gloucester, GL3 4AE and has a site office in Anglesey. It is owned by Horizon Nuclear Power Wylfa Holdings Limited. Along with Horizon Nuclear Power Oldbury Limited, and the parent company Horizon Nuclear Power Limited, it forms part of the Horizon group of companies. These companies are UK registered and owned by the Japanese corporation Hitachi Limited. The corporate structure is outlined in Figure 1.1. Figure 1.1 Overview of Corporate Structure KEY Companies currently within the Horizon Group Hitachi, Ltd. Horizon Wylfa the Site Licence and Environmental Permit Applicant Horizon Services the Company holding contracts with employees Hitachi Nuclear Projects Development Europe Limited The Horizon Group Horizon Nuclear Power Limited Horizon Nuclear Power Oldbury Limited Horizon Nuclear Power Wylfa Holdings Limited Horizon Nuclear Power Wylfa Limited (HORIZON) Horizon Nuclear Power Services Limited (HORIZON SERVICES) Page 2 of 255

17 1.3 Summary of the Wylfa Project 9. Horizon Nuclear Power Wylfa Limited is planning to construct and operate the Power Station on the island of Anglesey, Wales. It will be located to the west of the village of Cemaes and to the south of the existing Magnox power station (the Existing Power Station ). The Power Station will have two UK ABWRs and will be capable of generating 2.7 gigawatts of electricity. 10. Principal construction activities will start after the major permissions required to build the Power Station have been granted. Once construction of the first reactor has reached an advanced stage, it will be commissioned (expected to last two years) to ensure all systems and processes operate as intended. The first reactor will then become operational. This will be followed by the second reactor approximately sixteen months later. The expected operational life of each reactor is 60 years. 11. At the end of operations the Power Station will be decommissioned. This will involve the removal of buildings and structures from the site and the management of all spent fuel and radioactive wastes. The higher activity wastes (HAW) will be safely stored on site until a Geological Disposal Facility (GDF) becomes available. Once the buildings have been dismantled and the fuel and wastes removed, the Wylfa Newydd site will be cleared and delicensed. 1.4 Scope of this Application 12. This Application is for the controlled disposal of radioactive waste from the Power Station, which includes the discharge of gaseous radioactivity to the atmosphere via a number of authorised outlets, the discharge of aqueous radioactivity to the sea via the Power Station s cooling water return, and the disposal of solid radioactive waste to appropriately permitted off-site facilities. It will also allow Horizon to receive radioactive material onto the Power Station site and will thereby enable participation in the National Arrangements for Incidents involving Radioactivity (NAIR) and RADSAFE schemes. 13. This Application does not cover: The keeping (accumulation) and use of radioactivity on site. These activities are covered by the nuclear site licence (NSL) (see Section 1.5.4); Operation of an incinerator (there will not be an incinerator on site); Disposal of solid radioactive waste on the Power Station site; Radioactive doses to Horizon workers on the Power Station site, including those involved in construction and commissioning activities, resulting from the Power Station s normal operation; Radioactive doses to members of the public from accidents releases/events. These activities are covered by the NSL; or, Radioactive discharges from the Reactor Building (R/B) and Turbine Building (T/B) made via the Filter Vent Building in an emergency situation. These activities are covered by the NSL. Page 3 of 255

18 14. Management of nuclear safety, including accident and emergency scenarios, is not within the scope of this Application 1. This is managed by the Office for Nuclear Regulation (ONR), the UK s statutory regulator of nuclear safety, security, and conventional health and safety at nuclear sites, through the issuing of a NSL. 15. The limits for gaseous and aqueous discharges proposed in this Application apply to the commissioning and operation of the Power Station. Decommissioning will result in different gaseous and aqueous discharges. A variation to the EP-RSR will therefore be required in the future in order to accommodate this activity. 1.5 Other Applications and Submissions 16. In addition to the radioactive substances activities addressed by this Application, Horizon requires a number of other permissions, licences and consents related to the construction and operation of the Power Station. The key applications and submissions are described in the sections below. Further information can be found on Horizon s website ( 17. On 23 rd June 2016 the United Kingdom (UK) public voted in a referendum to leave the European Union, and the UK Government has since confirmed that it intends to negotiate the UK s exit under Article 50 of the Lisbon Treaty. However, exit is unlikely to be before the beginning of 2019, and it is not clear what the status of the UK s legal framework will be following the event. Therefore, for the purposes of this Application, it has been assumed that the current relevant legislative requirements, as described below, will remain in force for the foreseeable future Development Consent Order 18. A nuclear power station is a Nationally Significant Infrastructure Project (NSIP) under the Planning Act 2008 [RD3] and its construction and operation must be authorised by a Development Consent Order (DCO) granted by the relevant Secretary of State. Horizon currently anticipates that its application for a DCO for the Power Station will be made in A nuclear power station is also classified as an Environmental Impact Assessment (EIA) development under Schedule 1 of the Infrastructure Planning (Environmental Impact Assessment) Regulations 2017 [RD4]. This means that the Power Station will be subject to a mandatory EIA. The outcomes of this process will be reported in an Environmental Statement (ES), which will accompany the application for a DCO. 20. A number of complementary planning permissions under the Town and Country Planning Act 1990 as amended [RD5] and the Marine and Coastal Access Act 2009 as amended [RD6] will be required to support the development of the Power Station Regulatory Justification 21. The Justification of Practices Involving Ionising Radiation Regulations 2004 (Justification Regulations) [RD7] require that before any new class or type of practice involving ionising 1 The EP-RSR regulates radioactive discharges and off-site doses etc resulting from the normal operation of the Power Station and expected events, not accident and emergency scenarios. Page 4 of 255

19 radiation can be introduced in the UK, the Justifying Authority must first assess it to determine whether it is justified, i.e. whether the economic or social benefits associated with the class or type of practice outweigh the radiological detriments. 22. With the support of Horizon, the Nuclear Industry Association, as the trade association for the civil nuclear industry in the UK, submitted an application for the justification of the UK ABWR practice in December 2013 [RD8]. Following public consultation on both the application and the draft decision, the then Justifying Authority for nuclear energy, the Secretary of State for Energy and Climate Change, made the decision that the class or type of practice is justified under the Justification Regulations. 23. The practice is defined as The generation of electricity from nuclear energy using oxide fuel of low enrichment in fissile content in a light water cooled, light water moderated thermal reactor currently known as UK ABWR designed by Hitachi-GE Nuclear Energy, Ltd. The decision was given effect in a Statutory Instrument (The Justification Decision (Generation of Electricity by the UK ABWR Nuclear Reactor) Regulations 2015) [RD9], which came into force on 12 th February 2015, following approval by both Houses of Parliament Generic Design Assessment 24. GDA is the process by which the nuclear regulators the ONR and the Environment Agency (EA) first assess the safety, security and environmental implications of new nuclear reactor designs (without reference to site-specific issues). The UK ABWR is currently going through GDA the assessment covers one UK ABWR unit with HGNE being the Requesting Party. 25. The GDA documentation prepared by HGNE [RD10] sets out the generic safety, environment and security cases for the UK ABWR design. The main submissions are the Generic Pre-Construction Safety Report (PCSR), the Generic Environmental Permit Application (GEP) and the Conceptual Security Arrangements (CSA). These submissions are underpinned by relevant detailed reference documents. The PCSR sets out the demonstration that the design meets UK safety requirements and that the risks associated with the design are As Low As Reasonably Practicable (ALARP). The GEP describes how the design meets UK environmental requirements and details how BAT has been applied throughout the design of the UK ABWR. The CSA describes how the design meets the UK security requirements. 26. At the end of the assessment process the ONR and EA will issue reports on their findings indicating whether they consider the reactor design to be acceptable. Upon successful completion of the GDA assessments, a Design Acceptance Confirmation (DAC) and a Statement of Design Acceptability (SoDA) for the UK ABWR will be issued by the ONR and the EA respectively. These documents will be valid for a period of 10 years from the date of issue. After this time the UK ABWR design would need to be reviewed and reassessed. 27. NRW is a member of the EA GDA Programme Board and is also participating in the GDA. It will keep stakeholders in Wales updated on the progress of the UK ABWR s GDA. NRW is also participating in the governance of the GDA process, including technical oversight of the GEP programme and participation in the decision making process. This includes endorsement of the SoDA. Whilst the GDA process is generic for England and Wales, it will also underpin the site-specific assessments NRW will make for the Power Station [RD11]. Page 5 of 255

20 28. The EA and NRW undertook a 12 week public consultation (between the 12 th December 2016 and 3 rd March 2017) as part of their assessment of the UK ABWR design. The consultation explained how the regulators had conducted the assessment, described their findings, and sought the public s views on the preliminary conclusions regarding the environmental aspects of the UK ABWR design. It also set out what would happen next. The EA and NRW are now considering all relevant responses in order to finalise their assessment and ultimately, in December 2017, issue HGNE with a SoDA, or an interim Statement of Design Acceptability (isoda), for the UK ABWR design. 29. In determining environmental permit applications for specific sites, NRW will take full account of the work that has been done during the GDA so that efforts are focused on operator-specific and site-specific matters. This includes considering how the operator has addressed any relevant matters arising from GDA [RD12]. 30. GEP documents submitted on 8 th July 2016 have been used as the basis for the GDA information presented in this Application, and therefore act as a design reference (see Section 2.3 for further details). Where GDA information is presented in this document it has been validated as being directly applicable to the site-specific Application for the Power Station and has been incorporated into Horizon s site-specific Power Station design. Future changes to the generic design will also be applicable to its site-specific implementation. 31. It is expected that a number of Assessment Findings will result from the GDA process, and that these will be published by the EA/NRW along with the SoDA, or isoda, in December 2017 as part of the decision document. Section 8 of this Application provides information on Horizon s approach to developing a process to manage and resolve the Assessment Findings. 32. The ONR is currently undertaking an assessment of the UK ABWR Step 4 technical submissions. As part of this assessment, there is the potential for issues to arise which may impact upon the demonstation of BAT for the Power Station. Horizon will continue to monitor the assessments undertaken by the ONR so that any issues which have the potential to affect the environmental impact of the Power Station are appropriately managed [FA RSR-1] Nuclear Site Licence 33. A NSL will be required under the Nuclear Installations Act 1965, as amended [RD13] to install and operate the Power Station. A NSL is described by the ONR in its guidance document Licensing Nuclear Installations [RD14] as follows: Page 6 of 255 The safety of nuclear installations in Great Britain is secured primarily through the nuclear site licence and the conditions attached to it. Any organisation wanting to install or operate a prescribed nuclear installation will need a nuclear site licence. A nuclear site licence is granted for an indefinite period and, providing there are no material changes to the basis on which the licence was granted, it can cover the entire lifecycle of a site from installation and commissioning through operation and decommissioning to site clearance and remediation. 34. The NSL places Horizon under ONR regulation where it will oversee the licensee s control of the safety of the Power Station. This includes activities related to design, construction, installation, commissioning, operation, maintenance, modifications and decommissioning, including the accumulation or storage of radioactive waste.

21 35. Horizon submitted its application for a NSL in March The ONR will now undertake a programme of assessment and intervention to establish whether Horizon can demonstrate it will be in control of all safety related activities on its site. 36. In order to satisfy a number of requirements of the NSL and the EP-RSR, arrangements will be shared where possible. This will include areas such as the site boundaries 2, record keeping, training and managements systems Environmental Permits 37. In addition to the permit to dispose of radioactive waste from the operation of the Power Station the following environmental permits are required under EPR16 for potential discharges to the environment: Water discharge activity during pre-construction and construction (under Schedule 21 of EPR16). Discharges of non-radioactive substances in water will be made during construction of the Power Station; Water discharge activity during operation (under Schedule 21 of EPR16). Commissioning and operation of the Power Station will involve making discharges of non-radioactive aqueous effluent to the Irish Sea; and, Combustion activity during operation (under Schedule 1 of EPR16). Standalone emergency power supply will be required for each of the UK ABWR units at the Power Station. As the aggregated thermal input of the combustion units will be in excess of 50 Megawatts (MW), an environmental permit will be required Article Article 37 of the Treaty Establishing the European Atomic Energy Community (Euratom Treaty) [RD15] stipulates that each Member State shall provide the European Commission (the Commission) with general data relating to any new plan for the disposal of radioactive waste so that the Commission can determine whether the implementation of such a plan is liable to result in the radioactive contamination of the water, soil or airspace of another Member State. The legal duty is on the Member State to make the Article 37 Submission to the Commission, but in normal practice it is the operator of the nuclear facility who drafts the Submission for consideration by the Member State s government prior to formal submission to the Commission for its opinion. 39. Horizon has developed an Article 37 Submission for the Power Station and presented this to the relevant Government Department, for its onward submission to the Commission. Once the Submission has been received by the Commission, a group of experts, as established by Article 31 of the Euratom Treaty, will inform the Commission s assessment of the Submission. 40. The Commission s opinion on the Article 37 Submission is required before the radioactive waste disposals from the Power Station are authorised. A positive opinion on the Article 37 Submission is therefore an essential predecessor of the granting of an EP-RSR. 2 Not all of the site boundary for NSL and EP-RSR requirements is shared. Further information is provided in Section Page 7 of 255

22 1.6 Document Structure 41. This Application is set out in 10 separate sections as described in Table 1.1. An overview of the complete Application is given in the form of a summary, which has been prepared as a separate document. Table 1.1 List of Sections in this EP-RSR Application Section Number Section Content 1 Introduction General background information including the purpose and scope of this Application, its context with respect to other submissions and applications, and a cross-reference of the Application content against NRW requirements. 2 The Power Station An overview of the Power Station and its Reference Design, a high-level description of how the UK ABWR works, and a summary of the station s main plant and systems. 3 Radioactive Substances Activities at the Power Station 4 Operating Techniques 5 Disposal of Radioactive Waste Identification of the sources of radioactivity in the UK ABWR, and a detailed description of the plant and processes involved in the generation, treatment and disposal of gaseous, aqueous and solid waste. Also, information on radioactive waste discharge points and disposal routes are included in this section. Demonstration that BAT has been applied to the design of the plant, systems and processes planned for the Power Station in order to prevent, and where that is not possible, minimise the volume and activity of radioactive waste generated and discharged in a manner that minimises its impact. The Wylfa Newydd Power Station BAT Case is included as Appendix C to this Application. Quantification of the gaseous, aquesous and solid waste disposals including estimated volumes and activities. Proposed limits and notification levels for the discharge of gaseous and liquid wastes are included in this section. 6 Monitoring Description of the proposed measurement and sampling arrangements, techniques and systems at the Power Station, including methods for inprocess monitoring arrangements, as well as those for final discharges and disposals. It also presents information on the proposed environmental monitoring programme that will be deployed once operations at the Power Station have commenced. 7 Radiological Assessment 8 Management Arrangements Description of the dose assessments undertaken for members of the public and non-human species in the vicinity of the Power Station. Information about Horizon s management systems and the arrangements that will enable compliance with the EP-RSR and a high standard of environmental performance. The Company Manual and Management Prospectus are included as appendices to this Application. 9 Forward Work Plan Details of the activities that will be undertaken following submission of this Application to achieve compliance with the EP-RSR conditions. 10 References A listing of the Application references. Page 8 of 255

23 1.7 Cross-reference of Application Content with NRW Requirements 42. A correlation between the requirements of the NRW application forms (completed forms are contained in Appendix A) and the information presented in this Application is set out in Table 1.2. Table 1.2 Signposting of NRW Requirements within the Application NRW Requirement Reference Location in this Application Site reference number RSR-A, Q1 Application form RSR-A Appendix A. About you RSR-A, Q2 Application form RSR-A Appendix A. Applications from an individual RSR-A, Q3 Not applicable. Applications from an organisation of individuals RSR-A, Q4 Not applicable. Applications from companies or limited liability partnerships (LLPs) RSR-A, Q5 Application form RSR-A Appendix A. Applications from public or other corporate bodies RSR-A, Q6 Not applicable. Your address RSR-A, Q7 Application form RSR-A Appendix A. Contact details RSR-A, Q8 Application form RSR-A Appendix A. What is the name and address of the premises where you intend to carry out a radioactive substances activity Is a nuclear site licence under section 1 of the Nuclear Installations Act 1965 needed for the premises? Please provide a plan of the site, marking the site boundary in green RSR-A, Q9a Application form RSR-A Appendix A. RSR-A, Q9b Application form RSR-A Appendix A. See Section RSR-A, Q9c See Section Figure 2.5. Consultation RSR-A, Q10 Application form RSR-A Appendix A. Justification status RSR-A, Q11 Application form RSR-A Appendix A. See Section Your ability as an operator management systems RSR-A, Q12 See Section 8. Existing site contamination RSR-A, Q13 Not applicable. Other applications RSR-B3, Q1 Application form RSR-B3 Appendix A. Horizon s other key applications with respect to the Project are described in Section 1.5. The applications for other permits specifically under EPR16 are listed in Section What activities are you applying for? RSR-B3, Q2a Application form RSR-B3 Appendix A. A high level description of the Power Station is given in Section 2. A technical description of the radioactive substance activities which are being applied for is provided in Section 3. Is a submission to the European Commission under Article 37 of the Euratom treaty required for these activities? RSR-B3, Q2b Application form RSR-B3 Appendix A. See Section Page 9 of 255

24 NRW Requirement Reference Location in this Application Provide a technical description of your activities. RSR-B3, Q2c The overall function of the Power Station is described in Section 2.1. An overview of how the station works is given in Section 2.2. The main plants, systems and processes relevant to the activities are also described in Section 2.2. The systems and processes which have a bearing on the handling and treatment of radioactive waste are described in Section (gaseous), Section (aqueous) and Section (solid). The sources of radioactivity at the Power Station are identified in Section 3.1. The Power Station s radioactive waste discharge points and disposal routes are described in Section 3.3. The variation of discharges during the reactor cycle stages is discussed in Section Further information on the management of radioactive waste throughout the station s lifecycle is provided in Section 5 (see below). Operating techniques. RSR-B3, Q3 The demonstration of BAT for the Power Station is presented in Appendix C of this Application. The demonstration of BAT follows the Claims-Arguments- Evidence approach where the Claims mirror the requirements specified in RSR-B3. Provide quantitative estimates for normal operation. RSR-B3, Q4a Estimates of discharges of gaseous wastes are presented in Section and estimates for aqueous wastes in Section Estimates are provided for individual significant radionuclides and groups of radionuclies. Estimates of solid waste disposals are given in Section 5.8. The estimates are given by category (HLW, ILW, LLW). Provide your proposed limits. RSR-B3, Q4b Limits for gaseous and aqueous discharges are given in Section 5.6. A summary of discharge estimates, limits and quarterly notification levels for significant radionuclides and groups of radionuclides is also presented in Section 5.6. On site incineration will not be carried out. Provide a description of the sampling arrangements, techniques and systems for measurement and assessment of discharges and disposals of radioactive waste. RSR-B3, Q5a In-process monitoring arrangements are described in Section 6.2 and discharge monitoring arrangements in Section 6.3. The demonstration of monitoring BAT is presented in Appendix C of the Application (the Wylfa Newydd BAT Case). Provide a description of your environmental monitoring programme. RSR-B3, Q5b Environmental monitoring is described in Section 6.6. Page 10 of 255

25 NRW Requirement Reference Location in this Application Provide a prospective dose assessment at the proposed limits for discharges and for any on-site disposal. Provide an assessment of the impact of the radioactive discharges and on-site disposals on non-human species. Provide an assessment of the impact on people and non-human species of the environmental studies RSR-B3, Q6a RSR-B3, Q6b RSR-B3, Q6c Addressed as follows: Annual dose to most exposed members of the public for aqueous discharges Section Annual dose to most exposed members of the public for gaseous discharges Section Annual dose to the most exposed members of the public for all discharges from the facility Section Annual dose from direct radiation to the most exposed member of the public Section Annual dose to the critical group for the facility Section Potential short-term doses, including via the food chain Section A comparison of the calculated doses with the relevant dose constraints (taking account of any historical and future discharges from other facilities) Section An assessment of the build-up of radionuclides in the local environment Section Collective dose, truncated at 500 years, to the UK, European and world populations Conclusions are given in Sections There are no discharges to groundwater so this aspect is not assessed. Justification for the models used, data inputs, assumptions, etc are presented in the Radiological Assessment Technical Appendices (Appendices E-Q). Addressed as follows: Terrestrial Habitat Section Marine Habitat Section Freshwater Habitat Section Combined Impacts Section Conclusions are given in Section Not applicable. Receipt of radioactive waste RSR-B3, Q7 The Power Station may receive radioactive material as a result of Horizon s participation in NAIR or RADSAFE. In this instance the NRW guidance is that no further details are required. Radioactive material RSR-B3, Q8 Not applicable. Mobile radioactive apparatus for environmental studies RSR-B3, Q9 Not applicable. Charges and declarations RSR-F Application form RSR-F. 1.8 Abbreviations 43. A list of abbreviations is presented in Appendix B. Page 11 of 255

26 [This page is intentionally blank] Page 12 of 255

27 2 The Power Station 44. This section presents an overview of the Power Station, its operation, and its principal plant and systems. The information is intended to provide context to the detailed description of radioactive substances activities given in Section 3, and thereby addresses individual requirements of part 2c of Natural Resources Wales (NRW) application form RSR-B3 [RD16]. Initially, the operating principles of the UK Advanced Boiling Water Reactor (UK ABWR) are outlined after which the functions of the Power Station s main operational areas (those relevant to this Application) are explained. Information is then provided on the UK ABWR reference design and the reference design for the Power Station. Outline details are given on the Power Station s location and environmental setting (more detailed information can be found in Appendix E). Finally, information is provided on the Lifecycle Phases associated with the Project. 2.1 UK ABWR Basic Function 45. The Power Station will have two UK ABWRs along with supporting infrastructure. A simple schematic diagram of a UK ABWR nuclear power station of the type proposed at the Wylfa Newydd site is shown in Figure In the UK ABWR the heat produced inside the reactor core by the thermal nuclear fission process is removed by water which boils inside the Reactor Pressure Vessel (RPV), turning into high pressure steam (the water coolant also acts as a neutron moderator, enabling the nuclear reaction to be sustained). The steam passes through separators and dryers above the reactor core where water droplets are removed. It is then fed directly to the steam turbine generator which produces electricity. 47. The used steam exiting the turbine flows into condensers located beneath the low pressure turbines where it is cooled and converted back to water. It is then heated and pumped back into the reactor as feed water to repeat the cycle. Radioactive fuel is periodically removed from the reactor and stored in the Spent Fuel Pool (SFP) prior to dry storage in casks pending long term disposal. Page 13 of 255

28 Figure 2.1 Schematic of the UK ABWR Page 14 of 255

29 2.2 Main Structures at the Power Station relevant to the Application 48. The main plant and structures of each of the UK ABWR generating units are shown in Figure 2.2 and listed in Table 2.1. It is highlighted that the layout shown in the figure is for illustrative purposes only and may change. Figure 2.2 Layout of the Power Station Site Table 2.1 Main Buildings and Structures on the Power Station Site Building/Structure Number Building/Structure Number Reactor Building 1 Turbine Building 2 Control Building 3 ILW Storage Facility and Spent Fuel Storage Facility 4 Heat Exchanger Building 5 Service Building 6 Radioactive Waste Building 7 Lower Activity Waste Management Facility 8 Cooling Water Intake Structure 9 Cooling Water Seal Pit 10 Cooling Water Discharge Outfall 11 Filter Vent Building 12 Environmental Survey Laboratory Not shown Radioactive Drain transfer System and Discharge Tunnels Not shown Page 15 of 255

30 2.2.1 The Reactor Building 49. The Reactor Building (R/B) is a reinforced concrete structure that forms the secondary containment around the Reinforced Concrete Containment Vessel (RCCV) the UK ABWR s steel-lined primary containment vessel within which is located in the RPV see Figure 2.1. The R/B also contains the major portions of the Nuclear Steam Supply System (NSSS), the steam tunnel, SFP, the refuelling area, the essential power, the non-essential power, the Emergency Core Cooling System (ECCS), the R/B Heating, Ventilation and Air Conditioning (HVAC) system and other support systems. The SFP is used for the storage of new nuclear fuel prior to being loaded in the reactor, and the storage of spent fuel immediately after it has been removed from the reactor. 50. Each R/B has an exhaust stack directly on top of it. Unit 1 R/B main stack provides the discharge point for the unit s Off-Gas (OG) System and HVAC System, and the Radioactive Waste Building (Rw/B). Unit 2 R/B main stack provides the discharge point for the second unit s OG System and HVAC System only (i.e. the ventilation from the Rw/B is via the Unit 1 R/B stack only). 51. Inside the RPV there are 872 fuel assemblies, each of which is about 4.5 m in height. Each fuel assembly consists of 92 fuel rods, and each rod comprises a tube of zirconium alloy metal containing hundreds of enriched uranium dioxide fuel pellets. Each fuel pellet is approximately 1 cm diameter by 1 cm long. The fuel assemblies are held within the RPV along with control rods and monitoring instrumentation. Water passes around the exterior of the fuel rods turning into steam as it takes away the heat generated within the fuel by the fission reaction The Turbine Building 52. The Turbine Building (T/B) houses all equipment associated with the power conversion and auxiliary systems. This includes part of the Main Steam System (MS), the turbine and generator, the main steam condenser, the condensate and feedwater pumping and reheating equipment, and the OG system. 53. Steam generated within the RPV is passed through the multi-stage turbine causing the turbine shaft to rotate. This in turn drives the generator. The steam ultimately passes out of the turbine and into the condenser, where it is condensed against water in the Circulating Water (CW) System, and returned to the RPV via the feedwater pumping and re-heating equipment. 54. The OG system processes gases from the steam that do not condense in the condenser (e.g. noble gases). The system includes plant and processes such as filters and charcoal filters to reduce radioactivity in the gaseous phase prior to discharge to the environment via the R/B stack The Control Building 55. A Control Building (C/B) is situated between each R/B and its associated T/B. It contains the main control room for the generating unit, as well as some of the electrical switchgear and support systems needed to supply electrical power to the Power Station's auxiliary systems. The main steam tunnel from the R/B to the T/B is located on the ground floor of the C/B. Page 16 of 255

31 2.2.4 Cooling Water System 56. There are two cooling water systems for each UK ABWR unit: CW which provides cooling water to the steam turbine condenser; and, Service water (SW) which provides cooling water for the Reactor Building Service Water system (RSW), and the Turbine Building Service Water system (TSW). 57. Each cooling water system operates as a once through direct cooling system, with seawater abstracted from the sea, passed through the main condenser or heat exchanger, and then discharged back to the sea. The sea provides the ultimate heat sink (UHS) for the Power Station. Both UK ABWR units draw their cooling water requirements from a single intake structure located at Porth-y-pistyll, and share a common cooling water outfall structure in Porth Wnal. 58. Seawater for the CW systems of both UK ABWR units is abstracted through eight drum screens into a common compartment (the discharge bay). Seawater for the SW systems is abstracted through band screens located in pairs on each side of the drum screens. Seawater for the SW system of Unit 1 is drawn through one pair of band screens to collection pits in the Heat Exchanger Building (Hx/B), and the seawater for Unit 2 is drawn through the other pair. The seawater for the RSW and TSW systems is supplied from these collection pits. 59. After passing through the main steam condenser the seawater is discharged via a ball strainer pit to the seal pit. Each UK ABWR unit has a single seal pit, which also receives the discharges from the RSW and TSW systems, process liquid effluents discharged from the Power Station, and surface water runoff arising from within the inner security surface water (these effluents are covered by the Water Discharge Activity Environmental Permit Application referred to in Section 1.5.5). From the seal pit, the seawater (with process liquid effluent, and surface water runoff) is discharged by gravity through a cooling water discharge tunnel to a shared cooling water outfall structure (see Figure 2.3). 60. Discharges from the cooling water system are those arising from the CW, RSW and TSW systems. These systems will not receive drainage from process areas within the Power Station, and consequently should not themselves have any radioactive contaminants present. However, while BAT has been applied to minimise radioactivity in aqueous wastes that are generated in the UK ABWR (see Section 4), some radioactivity will ultimately be discharged to the environment. All of the radioactive aqueous effluent will be discharged with the cooling water via the cooling water outfall. Further details on the discharge of radioactive aqueous waste are given in Section 3 and Section The cooling function of the CW, TSW and RSW systems means that the seawater discharged from them, via the cooling water return, will be warmer than the seawater abstracted at the cooling water intake. At power operation the cooling water flow rate will result in the cooling water being up to 12 C warmer than the receiving seawater at the point of discharge. 62. The position of the cooling water outfall structure is shown in Figure 2.3. The cooling water and radioactive effluent will be discharged adjacent to the Existing Power Station outfall and will re-use the channel already cut into the sea bed. The outfall is being built on individual plots of land owned by the Nuclear Decommissioning Authority (NDA), the Crown Estate and Horizon. Commercial discussions regarding land ownership between Horizon, the NDA and the Crown Estate are on-going. Page 17 of 255

32 Figure 2.3 Cooling Water Outfall Location Heat Exchanger Building 63. The Hx/B contains portions of R/B and T/B cooling water systems (RSW, TSW, RCW and Turbine Building Cooling Water System (TCW)). The Hx/B is located close to the sea water intake to ensure proximity to the cooling water source Filter Vent Building 64. Radioactive gases arising within the R/B and T/B in an emergency situation are routed through filtration and monitoring equipment in the Filter Vent Building. This enables the radioactivity of such gases to be reduced prior to being released into the environment. Discharges from this facility are not within the scope of this Application. Page 18 of 255

33 2.2.7 Radioactive Waste Building 65. There is one Rw/B for both generating units, which will be available during Commissioning, Generation and Decommissioning Phases. It contains equipment associated with the collection and processing of aqueous radioactive waste arisings, and the packaging of wetsolid waste, including: Low Chemical Impurity Waste System (LCW); High Chemical Impurity Waste System (HCW); Controlled Area Drains System (CAD); Wet-solid Intermediate Level Waste (ILW) Processing System; and, Wet-solid Low Level Waste (LLW) Processing System. The facility vents gaseous discharges via the main stack on the R/B for Unit 1. There are also treated aqueous effluent discharges from the HCW and CAD to sea via the cooling water outfall. There are no aqueous discharges to the environment from the LCW system. 66. The Wet-solid ILW and LLW Processing Systems collect and process materials such as bead resins, evaporator concentrates (sludges), powder resins and sludge (crud). Wetsolid ILW and LLW stored in the storage tanks are discharged to the Wet-solid ILW and Wet-solid LLW Processing Systems via a short shielded pipe where the waste is immobilised by encapsulation in a cement based grout. There are not expected to be aqueous radioactive discharges from the encapsulation process Lower Activity Waste Management Facility 67. The Power Station is provided with a single Lower Activity Waste Management Facility (LAWMF) serving both generating units, which will be available for use at the start of operations. The facility receives, processes and packages dry LLW generated across the Power Station prior to disposal. It has a small stack (3 m above the roof ridge) which releases gaseous discharges to atmosphere Service Building 68. A Service Building (S/B) serves as the main entrance and exit to the R/B, C/B, Rw/B and T/B (both units), and is the route through which maintenance, operations, chemistry, plant engineering and other support personnel access the generating unit. Health physics personnel are located in the S/B in order to control these activities within radiationcontrolled areas. An active laboratory for the handling and analysis of radioactive materials is located within a radiation controlled area in the S/B ILW Storage Facility and Spent Fuel Storage Facility 69. There are a number of other radioactive waste buildings that will serve the whole Power Station, namely: Spent Fuel Storage Facility (SFSF); and, ILW Storage Facility. Page 19 of 255

34 70. Spent fuel will be stored in the SFPs (one SFP per reactor) for a period of up to 10 years. After this period it will be transferred to the SFSF for storage for up to 140 years prior to disposal to the Geological Disposal Facility (GDF). There will be one shared SFSF for the two generating units. The design of the facility is in development and yet to be confirmed. However, it will be designed to accommodate the lifetime arisings of spent fuel that will be generated, and will be located in the south west corner of the site. 71. Ultimately, the spent fuel will require repackaging prior to disposal and, as such, it is anticipated that a separate repackaging facility will be required adjacent to the SFSF after The SFSF will also receive HLW casks from the R/B. As a result of decay storage, the HLW will be re-categorised to ILW prior to repackaging and disposal to the GDF. 72. There will be one shared ILW Storage Facility for the two generating units. The purpose of the ILW Storage Facility is to receive ILW packages from the Wet-solid ILW Processing System for storage prior to disposal to the GDF. 73. The SFSF and ILW Storage Facility will not be required until approximately 10 years after the start of operations. This is because: a) the spent fuel and HLW will be cooled in the SFPs for approximately 10 years prior to transfer to the storage facilities, and b) the first campaign of ILW processing will not commence until approximately 10 years after the start of operation. During this period the ILW will be stored in tanks in the Rw/B. 74. No aquesous discharges are anticipated from these facilities during normal operations. Gaseous discharges from the facilities will be negligible, although a small stack may be required from each facility to aid ventilation Radioactive Drain Transfer System and Discharge Tunnels 75. The radioactive drain transfer system is used to transfer waste water collected in the radiation controlled areas in the R/B, T/B and S/B into collection tanks installed in the Rw/B. The radioactive drain subsystem comprises sump tanks, sump pumps, piping, valves, and appropriate instrumentation. Prior to release to the environment via the cooling water outfall (transported along the discharge tunnels), aqueous discharges will be sampled in order to confirm that they meet required discharge criteria Environmental Survey Laboratory 76. An Environmental Survey Laboratory (ESL) will be constructed off the Power Station site to support the performance of environmental monitoring. Further details of the proposed environmental monitoring programme are provided in Section 6. Discharges from this facility are not within the scope of this Application. 2.3 Development of the Reference Design 77. The ABWR was developed primarily in Japan and the USA and was based on an evolution of conventional Boiling Water Reactor (BWR) technology. The development was started in 1978 by Japanese electric utilities and plant manufacturers, including Hitachi Limited in Japan and General Electric Company in the US, in collaboration with various international partners. 78. Hitachi-GE Nuclear Energy Limited (HGNE) has completed the design and construction scope of four ABWR units which have been operational in Japan. The units are: Page 20 of 255

35 Units 6 and 7 of Kashiwazaki-Kariwa Nuclear Power Plant of TEPCO (commenced commercial operation in 1996 and 1997 respectively); Unit 5 of Hamaoka Nuclear Power Plant of Chubu Electric Power Co (commenced commercial operation in 2005); and, Unit 2 of Shika Nuclear Power Plant of Hokuriku Electric Power Company (commenced commercial operation in 2006). HGNE is also involved in the on-going construction of the Shimane 3 and Ohma ABWRs in Japan. 79. The UK ABWR derives from the design of the ABWR. The design reference for the UK ABWR will be the standard design of the first ABWR (Kashiwazaki-Kariwa Units 6 & 7) incorporating further improvements and optimisation from the subsequent ABWR plants and changes made during Generic Design Assessment (GDA). 80. The UK ABWR has been designed with the aim of simplifying the design and operation of the plant compared to the original BWR technology, enhancing the safety and reliability of Structures, Systems and Components (SSCs), and minimising the amount of material that is ultimately treated as waste. Some of the major UK ABWR improvements and differences relative to previous BWRs include: Improvement of safety and reliability; Improvement of capacity factor; Reduction of radiation dose to which workers are exposed; and, Improvement of operability and maintainability. Improvements such as those highlighted above enhance the environmental performance of the reactor as operations are made even more predictable, straightforward and safe Reference Design for Wylfa Newydd 81. The reference design for the Power Station is largely based on the GDA Generic Environmental Permit Application (GEP) documents which were submitted to NRW and the Environment Agency (EA) on the 8 th July The principal aspects of the Power Station design which differ from the UK ABWR design assessed as part of GDA are as follows: The Power Station comprises two reactors where only a single reactor design was assessed at GDA; The Power Station does not have a laundry and the associated drainage system; The Power Station has a Rw/B shared between both reactors instead of one Rw/B for each UK ABWR; The Power Station has a S/B shared between both reactors instead of one S/B for each UK ABWR; The height of the stack on each R/B is 75 m (above ground level); and, The locations of the cooling water intake and outfall have been established. Page 21 of 255

36 2.4 Location and Setting 82. The Power Station will be developed on the island of Anglesey (see Figure 2.4) to the west of the village of Cemaes and south of the Existing Power Station as shown in Figure 2.5. The geographical centre of the Power Station site is at º latitude, º longitude, Ordnance Survey (OS) grid reference SH Figure 2.4 Location of Wylfa Newydd Site Proposed Boundary of the Permitted Premises 83. The area indicated in Figure 2.5 represents the limit of the site on which the radioactive substances activities described in this Application will take place. The majority of the proposed boundary is contiguous with that for which a Nuclear Site Licence (NSL) application has been submitted. The exception to this is the area to the north of the site covering the cooling water outfall pipes and ultimately the discharge point to sea. The boundary runs 5 m either side of the exterior of the outfall pipes and along the mean high water mark. Page 22 of 255

37 Figure 2.5 Proposed Boundary of the Permitted Premises Environmental Setting 84. The Wylfa Newydd Development Area (WNDA) the area of land which will include the Power Station site and the surrounding areas which will be used for its construction and operation is shown in Figure 2.6, along with the National Policy Statement Area site for Wylfa 3 and the EP-RSR site boundary. 85. The WNDA is bounded to the north by the coast and the Existing Power Station. To the east, it is separated from Cemaes by a corridor of agricultural land. The A5025 and residential properties define part of the south-east boundary, with a small parcel of land spanning the road to the north-east of Tregele. To the south and west, the WNDA abuts agricultural land, and to the west it adjoins the coastal hinterland of Cemlyn Bay, where the WNDA includes part of the property known as Cestyll Garden. 86. The area of the Power Station site (only those parts within the permanent security fence demarcating the operational power station) is about 80 hectares (including about 10 hectares of marine area enclosed by the breakwaters). The size of the WNDA is about 380 hectares. 3 The Wylfa site was identified by the UK Government as being potentially suitable for the deployment of a new nuclear power station as part a National Policy Statement for Nuclear Power Generation EN-6 (NPS EN-6). Further information on NPS EN-6 can be found at Page 23 of 255

38 Figure 2.6 Geographic Areas of the Project 87. The Existing Power Station, which started operating in 1971, sits directly to the north of the proposed siting of the Power Station. The Existing Power Station has two Magnox reactors and associated turbo-generators; Reactor 2 was shut down in 2012 and Reactor 1 in The process of decommissioning the site has now commenced [RD17]. 88. Land within and surrounding the WNDA is predominantly agricultural, used for grazing sheep or cattle, contained by hedgerows and crossed by a network of roads, rural lanes, watercourses and overhead electricity infrastructure. The local coastline is used for various recreational activities including walking, dog-walking, bird-watching, picnicking, water sports and other beach activities. A number of public rights of way, including the Wales Coast Path and the Copper Trail (national cycle route) cross the WNDA. However, these will ultimately need to be diverted for reasons of security and safety. 89. There are a number of sites nearby which are subject to ecological conservation designations (both statutory and non-statutory) of international, national and local importance. The most notable of these are the Tre'r Gof and Cae Gwyn sites of special scientific interest (SSSIs) which are, respectively, within and adjacent to the WNDA. Cemlyn Bay to the west forms part of the Anglesey Terns Special Protection Area (SPA) and the Cemlyn Bay Special Area of Conservation (SAC). Page 24 of 255

39 2.5 Wylfa Newydd Project Lifecycle Phases 90. The Project will comprise five Lifecycle Phases: Development; Construction (including Construction Testing and Pre Operational Testing); Commissioning (including Start-up Testing); Generation; and, Decommissioning. These lifecycles are described below. It should be noted that the phases are not always discrete, and will at times occur in parallel, for example during the staggered construction and commissioning of the two units Development 91. The Development Phase will include the following activities: Site preparation and clearance; Procurement of Long Lead Items (LLI) 4 ; Establishment and oversight of the Engineering Procurement and Construction (EPC) Contract; Application for major permissions; and, Development of an organisation capable of holding the major permissions and undertaking nuclear safety-related construction in the subsequent Lifecycle Phase. 92. The Phase will end with the release of the Hold Point 5 enabling the first structural concrete pour for the reinforced concrete Base Slab (First Nuclear Concrete (FNC)) on the Nuclear Island, the undertaking of which marks the commencement of the Construction Phase Construction 93. The Construction Phase will start with the concrete pour for the reinforced Base Slab (FNC) on the Nuclear Island. The site will see significant change during this Phase as the Power Station is constructed. The procurement of remaining Power Station plant items will be undertaken, and all plant items will be transported to site, tested for acceptance, and installed. 94. Construction activities will include Construction Testing and Pre-operational Testing. Construction testing includes examinations of the Power Station components, examples of which include insulation and resistance cable tests, pipework and component flushing and pressure tests, instrument calibration, and Control and Instrumentation loop tests. Pre- 4 As part of Horizon s oversight of procurement of material and fabrication of LLIs, it is expected that the commencement of factory acceptance testing will be undertaken in the Development phase, prior to acceptance, receipt and storage of the first LLI. 5 Hold Points are applied to activities (such as construction and commissioning) to constrain them from proceeding until the organisation has demonstrably achieved readiness to do so. They are applied where there is a significant change in safety risk (e.g. nuclear safety, non-nuclear safety, the environment, security etc) or where there is foreclosure of options by carrying out a practically irreversible activity such as FNC. A Hold Point is released by the appropriate releasing authority, and may also be subject to regulatory assessment and permission. Page 25 of 255

40 operational Testing, sometimes referred to as inactive commissioning, involves the conduction of a series of tests on the Power Station systems including the testing of filtration and treatment systems, and monitoring equipment. 95. Facilities relevant to the processing and disposal of radioactive waste will be constructed although this will not include the ILW Storage Facility or SFSF as these facilities, as previously noted, will not be required until approximately 10 years after the start of generation. 96. It is anticipated that some radioactive wastes may be disposed of from the site during the Construction Phase. In the event these wastes are generated, they will primarily be radioactive sources that become damaged and require disposal. There is also the potential for the presence of land contaminated with radioactivity from the Existing Power Station. Should such land be encountered the contaminated materials will be treated where necessary and disposed of, as required, by a suitably qualified contractor. Horizon will ensure that sufficient competent resources are in place to oversee these disposals. 97. The Construction Phase will end with the release of the Hold Point for back-energisation of the Standby Auxiliary transformer and the First System Level Testing to enable the commencement of the Commissioning Phase Commissioning 98. The Commissioning Phase will see completion of site construction activities, nuclear fuel being brought to site, first criticality, disposal of the first radioactive wastes generated from the testing of systems in place to treat gaseous, aqueous and solid radioactive waste arisings, and completion of Horizon s capability as a competent operator. 99. Commissioning activities will include Start-up Testing, sometimes referred to as active commissioning. The testing will commence with receipt of the first nuclear fuel assembly at the unit and will end with the final test prior to commercial operations. Once all generating unit systems have successfully completed pre-operational testing nuclear fuel assemblies will be loaded into the core. The reactor will be taken critical and then operated over a series of power levels to demonstrate plant performance under static and transient conditions. During this period electricity will be generated and exported to National Grid s high voltage electricity transmission network Following successful completion of the commissioning activities and release of the Commencement of Commercial Operations Hold Point, commencement of the Generation Phase will be enabled Generation 101. The Generation Phase follows completion of commissioning. Electricity will be generated and exported to the national grid. During this Phase there will be periods of maintenance and refuelling. The Phase will continue until the Power Station ceases generation and disconnects from the grid. Page 26 of 255

41 2.5.5 Decommissioning 102. Decommissioning is the final Lifecycle Phase, and begins when the station is shut down and ceases generating electricity. It is currently anticipated that this Phase will include the removal of all buildings and facilities, the on-site storage of higher activity waste (HAW) and spent fuel pending disposal in the UK GDF, remediation of the land, if required, and the justification for the surrender/removal of all licenses and permits, for example the NSL and EP-RSR. Page 27 of 255

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43 3 Radioactive Substances Activities at the Power Station 103. This section provides a technical description of the radioactive substance activities which will be carried out at the Power Station, and thereby addresses the requirements of part 2c of Natural Resources Wales (NRW) application form RSR-B3 [RD16]. Initially, an explanation is provided of the mechanisms by which gaseous, aqueous and other radioactive wastes are generated in the Power Station (the means by which the waste generation is minimised are detailed in Section 4). Descriptions are then given of the systems which are in place to treat the waste arisings in order to minimise the radioactivity and volume of waste that is subsequently released to the environment. The discharge points and disposal routes for the treated radioactive waste streams are identified. Finally, the likely variation in the discharges during the reactor cycle stages are described. Quantified estimates of radioactive waste emissions and discharges are given in Section Sources of Radioactivity 104. Operation of the Power Station will lead to the generation of radionuclides that will then transfer into the water/steam recirculating through the Reactor Pressure Vessel (RPV) and be transported through the reactor systems. These radionuclides can be grouped into three main categories: Corrosion products (CP) that subsequently become activated; Fission products (FP) (including noble gases) and Actinide products (ActP); and, Activation products (AP). The sections below map the generation and movement of these radionuclides from the reactor core through the treatment and handling systems described in Section Corrosion Products 105. Metallic impurities such as iron and cobalt are generated by the corrosion of structural materials. These stable CPs are deposited on the fuel cladding surface and exposed to neutrons. Consequently, they become activated. Cobalt-58 and cobalt-60 are also formed by neutron activation of nickel-58 and cobalt-59 in reactor components and insoluble and soluble metal sludge and particulate in the reactor water. The generation mechanism for significant CPs is shown in Table 3.1. The production of CPs is affected by water chemistry, zinc injection, feed-water iron concentration and material selection. Further information in this area is provided in Section 4. Page 29 of 255

44 Table 3.1 Generation Mechanism for Significant Corrosion Products Significant Corrosion Products Cobalt-58 Cobalt-60 Zinc-65 Chromium-51 Manganese-54 Iron-59 Production Mechanism Nickel-58 (n, p) Cobalt-58 Cobalt-59 (n,γ) Cobalt-60 Zinc-64 (n,γ) Zinc-65 Chromium -50 (n,γ) Chromium-51 Iron-54 (n,p) Manganese-54 Iron-58 (n,γ) Iron-59 Note. In the table, n signifies neutron capture, α signifies alpha radiation, γ signifies gamma radiation and p signifies proton release 106. Most CPs will remain in the reactor water phase as carry-over into steam is low. The CPs in the water will circulate through the Reactor Water Clean-up System (CUW) see Section and will accumulate in the CUW Filter Demineralisers (FDs). Powder resins from these demineralisers will ultimately be transferred to the Radwaste building (Rw/B) for onward processing CPs that do carry over into the reactor steam are assumed to partition into the water condensate in the down stream hotwell (the reservoir which receives the condensate) and be carried forward to the Condensate Water Clean-up System see Section These CPs will accumulate in the system filters and demineralisers, and like the above will be transferred to the Rw/B as sludge (crud). CPs that are transferred into the Spent Fuel Pool (SFP) will accumulate on the Fuel Pool Cooling and Clean-up System (FPC) FDs see Section As with the CW FDs, the products will be transferred to the Rw/B as powder resin Fission Products 108. FPs can be split into four sub-groups: noble gases, halogens, soluble FP, and insoluble FP. It is highlighted that the majority of the FPs are retained within the fuel assemblies and are managed as spent fuel. However, some FPs can be released into the reactor water through two principal mechanisms: FPs can pass into the reactor water in the event of fuel pin failures, typically small cracks in the fuel cladding; and, Activation of tramp uranium present on the external surface of the fuel. The tramp uranium is present either due to fissile isotopes which have washed out of defective fuel and plated-out on the outside surfaces of the fuel cladding, or uranium impurities that are left on the fuel cladding from the manufacturing process. The generation mechanisms for significant fission and actinide products are shown in Table 3.2. Page 30 of 255

45 Table 3.2 Generation Mechanism for Significant FP and ActP FP Noble gas (Xe, Kr) Halogens (I2) General Nuclear Reaction U-235 (n) FP1 + FP2 + ns Insoluble FP Soluble FP (Cs, Sr, Ce) ActP Nuclear Reaction Noble Gases U, Pu, Am, Np Example: U-238 (n,β) Np-239 Note. In the table, n signifies neutron capture and β signifies beta radiation Noble gas FPs that pass into the reactor water partition entirely into the reactor steam phase. These isotopes do not condense in the condenser so they are routed to the OG system where they are retained by charcoal adsorbers to enable radioactive decay prior to discharge in gaseous form via the Reactor Building (R/B) stack see Section In their path through the steam systems, noble gas FPs have the opportunity to decay. The majority of daughter products are water-soluble alkali metals and these partition into the condensate where they are carried forward to the Condensate Water Clean-up System. The decay schemes are given in Table 3.3. The daughter products from longer lived metastable parents are abated via the final charcoal adsorber beds in the OG System. Page 31 of 255

46 Table 3.3 Decay Schemes of FP Noble Gases to 4 th Generation Daughters Parent Noble Gas in Steam/Gas Kr-85m (4.48 h) Kr-87 (76.3 min) Kr-88 (2.84 h) Daughter Nuclides (Partition to Water) 2 nd 3 rd 4 th Kr-85 Rb-87 Rb-88 Kr-89 (3.15 min) Rb-89 Sr-89 Y-89m Kr-90 (33.2 s) Rb-90m Rb-90 Sr-90 Sr-90 Y-90 Rb-90 Sr-90 Y-90 Kr-91(8.57 s) Rb-91 Sr-91 Y-91 Y-91m Xe-133m (2.19 day) Xe-135 (9.14 h) Xe-135m (15.3 min) Xe-133 Cs-135 Xe-135 Cs-135 Cs-135 Xe-137 (3.82 min) Cs-137 Ba-137m Xe-138 (14.08 min) Cs-138 Xe-139 (39.86 s) Cs-139 Ba-139 Xe-140 (13.6 s) Cs-140 Ba-140 La Halogens: Isotopes of Iodine 111. Iodine-131 is formed by fission of the fuel and tramp uranium. It can migrate into the steam phase although some remains in the reactor water. Other isotopes of iodine formed by the fission process are mainly iodine-132, iodine-133, iodine-134 and iodine-135. The iodine isotopes can have three main forms: particulate (caesium iodide, CsI), elemental iodine (molecular iodine vapour, I 2), and organic iodine (many forms are possible but usually considered to be methyl iodide, CH 3I) Iodine that is carried over into the reactor steam phase is mainly in inorganic form. It partitions into the condensate and, due to its high boiling point, only a small amount remains in a non-condensable phase. That which does not condense is routed to the OG System and held up by charcoal adsorbers to allow for radioactive decay prior to discharge via the R/B stack Iodine that remains in the aqueous phase builds up in the reactor water and fuel pool water and is circulated through the CUW and FPC where it is filtered through demineralisers. Volatile iodine in the vapour from the reactor and fuel pool water is discharged via the heating, ventilation and air conditioning (HVAC) system. Page 32 of 255

47 114. In the aqueous system, the hold-up of halogens in filters, demineralisers or tanks, results in the generation of their daughter noble gases. The decay schemes are given in Table 3.4. It is assumed these daughter noble gases remain partitioned in the water phase at the outlet of the demineralisers, so pass to the downstream water systems. Table 3.4 Decay Scheme of Halogens to Noble Gas Daughters Parent Nuclide Noble Gas (Daughter Nuclide) I-131 (8.02 day) Xe-131m I-133 (20.8 h) I-134m (3.6 min) I-135 (6.5 h) Br-88 (16.5 s) Xe-133 Xe-133m Xe-134m Xe-135 Xe-135m Kr-87 Kr Soluble and Insoluble Fission Products 115. Carry-over of soluble and insoluble (particulate) FPs into the steam is low. Radionuclides such as strontium-89, strontium-90, caesium-134 and caesium-137 that remain in the aqueous phase build up in the reactor water and fuel pool water and are circulated through the CUW and FPC as described above Activation Products 116. The neutron activation of the reactor water will also produce radioactive species. Of particular interest are Carbon-14, Argon-41 and Tritium; the main production mechanisms for these nuclides, along with those for Nitrogen-16, Nitrogen-17, and Chlorine-36, are described in the following sub-sections Tritium 117. Tritium, also known as hydrogen-3 (or H-3), is produced by a number of mechanisms, including activation of naturally occurring hydrogen-2 (deuterium) in the reactor water, ternary fission of the fuel, and from boron in the control rods during the neutron absorption process. Activation of hydrogen-2 in the reactor water is the main production route. Tritium is also generated, to a lesser extent, as a ternary fission product in the fuel and in the boron carbide matrix of the control rods. However, for these two examples, the escape of tritium into the coolant requires defects in the fuel pin and control blade cladding materials Tritium is predominantly present as tritiated water and is assumed to partition equally between the steam and water phases. Tritium that becomes entrained into the steam is routed to the condenser and subsequently to the OG System. In the OG System molecular hydrogen and oxygen are combined see Section The hydrogen (and therefore the tritium) is converted to water which is routed to the Condensate Storage Tank (CST) Page 33 of 255

48 from where it is returned for reuse within the plant. Gaseous residual uncombined tritium is discharged to atmosphere via the R/B stack Tritium in the reactor water is unaffected by demineralisation or filtration. There will be some loss of tritium as a result of evaporation from the SFP and the CST. The evaporated tritiated water will be discharged to atmosphere via the HVAC system and the Turbine Gland Steam (TGS) system. Following steam leaks from the system some tritiated water will also be routed to drains and transferred to the appropriate radioactive liquid effluent treatment system (there is no abatement of tritium prior to discharge via these systems) Argon Argon-41 is generated by the activation of argon-40 from entrained air in the coolant during power operation. This radionuclide is completely transferred to the main steam as it is a noble gas. The argon-41 is routed to the OG System where it is retained by charcoal adsorbers in order to enable radioactive decay prior to discharge via the R/B stack. The charcoal delay beds offer seven hours of hold up time for argon-41 which achieves a reduction in radioactivity of 1/14 th of its activity prior to abatement. The daughter products from the decay of argon-41 are assumed to partition into the condensate and are carried forward to the Condensate Water Clean-up System Carbon The principal source of carbon-14 is the thermal neutron reaction with oxygen-17 in the reactor coolant. Carbon-14 can also be produced by neutron activation of nitrogen-14 dissolved in the reactor coolant. This source contributes a small fraction of the inventory to the annual production of carbon-14 due to the low concentration of nitrogen-14 in the reactor coolant (less than 1 ppm by weight). Furthermore, carbon-14 can be generated from carbon released into the coolant by the corrosion of structural carbon and organic carbon (carbon-13) within the reactor water. The production mechanism for C-14 is shown in Table 3.5. Table 3.5 Mechanisms for Production of C-14 Parent Nuclide Source Generation Route N-14 Dissolved gas in reactor water N-14 (n,p) C-14 O-17 Reactor water O-17 (n,α) C-14 C-13 Impurity in metal Organic carbon C-13 (n,ɣ) C Conditions in the upper part of the reactor core are oxidising and most carbon-14 exists as 14 CO 2 and is transferred to the reactor steam. For the purposes of quantifying carbon-14 in gaseous discharges, OPEX data suggests that in the condensate systems the CO 2 remains predominantly in the non-condensable phase, i.e. in the headspace of the main condenser from where it is extracted into the OG System. Carbon-14 is ejected via the R/B stack as CO 2. There is no abatement of carbon-14 in the OG System. Page 34 of 255

49 123. For the purposes of quantifying carbon-14 in solid waste arisings, it is assumed that there is some accumulation of carbon-14 in the demineralisers dealing with reactor water (CUW) and condensate (Condensate Water Clean-up System). Between 1% and 6% of the carbon-14 generated in the reactor water is disposed of as solid waste Nitrogen Nitrogen-16 is formed by neutron activation of oxygen-16 in the reactor water [O-16 (n,p) N-16]. Under normal water chemistry (NWC) conditions, most of the nitrogen-16 is retained in water soluble NO 3 and thus exists primarily in the coolant. Only a small percentage is transferred to the steam phase. Conversely, under hydrogen water chemistry and online noble chemistry (HWC + OLNC) conditions described in Section 4 the additional hydrogen present allows formation of the more volatile NH 3 and/or NO. Being more mobile, these compounds are carried over into the steam Most of the nitrogen-16 activity in the steam phase will migrate to the OG System although as the half-life is short (7.12 seconds) the nitrogen-16 will have decayed and the concentration will be low. Due to the short half-life, under normal operation, nitrogen-16 will not be discharged to the environment in either the gaseous or aqueous effluents Chlorine The principal generation pathway for chlorine-36 is by neutron activation of naturally occurring chlorine (Cl-35) present in the reactor coolant. Sources of Cl-35 within the coolant include leaching of chlorine from the clean-up resin system, ingress of seawater via defective condenser tubes, and ingress from the surrounding atmosphere during plant outage. The chlorine-36 will migrate to the condensate water and will not be found in gaseous form in the reactor steam. Although its concentration will be low, as a consequence of the very long half-life (c. 300,000 years), it is expected that chlorine-36 will be transferred to the Rw/B and will be associated with solid wastes such as the spent resin from the demineraliser. 3.2 Systems Associated with Radioactive Waste 127. The sources of radioactivity in the reactor water/steam, and the pathway of the reactor water/steam through the downstream systems is illustrated in Figure 3.1. The systems that are in place to treat gaseous, aqueous and solid radioactive waste arisings in order to minimise the radioactivity and volume of waste that is subsequently released to the environment are shown in Figure 3.2 and described thereafter. Page 35 of 255

50 Figure 3.1 Primary Pathway from the RPV to the Downstream Systems Page 36 of 255

51 Figure 3.2 Overview of Radioactive Waste Management Processes Main Condenser Equipment Drains CF Cartridge Filters Packaged Spent Filter Cartridges Floor Drains Demineraliser CD Backwash Sludge receiver tank CAD Sump tank Floor Drains Reactor Equipment Drains CUW Filter Demineraliser Spent Resin/Crud reciever tank Spent Fuel Control Rods Reactor Components Multi Purpose Canister (MPC) in Transfer Cask Non-Fuel Waste Canister (NFWC) in Transfer Cask Spent Filter Media LAW Characterisation Direct Disposal LLW & VLLW Combustible Waste Packaging for Transport Packaging for Transport Dispatch for Direct Disposal (LLW/VLLW) Dispatch for Incineration LCW Sump Tank SPCU Fuel Pool LCW Sump tank CAD Sump tank Turbine Building Drywell LCW Sump Drywell HCW Sump Supression Pool Fuel Pool FPC Filter Demineraliser Spent Powder Resin and Crud HVAC Filters LAW Segregated Heterogeneous LAW (from all Buildings) Characterisation Characterisation Compactable Waste Metal Waste (suitable for recycling) Packaging for Transport Packaging for Transport Dispatch for Supercompaction Dispatch for Metal Recycling Outage Only Spent HVAC Filters Spent Bead Resin Spent Cartridge Filter Media Filter Backwash Sludge Reactor Building Potential LAW (oils/contaminated land) Decontam If Req d Decontam Secondary Wastes No current treatment route Lower Activity Waste Management Facility Out of Scope Conventional Waste Management Spent HVAC Filters Equipment drains Floor Drains Radioactive Effluent Management Process Condensate Storage Tank Powder Resin Storage Tank Supernate recycling Cement Powders Wet Solid ILW Process THISO HLW Storage Cask Spent Fuel Storage Cask LCW Spent Fuel Storage Facility LCW Collection tank Hollow fibre filters Demineraliser LCW Sample tank Filter Backwash Sludge Filter Crud Storage Tank ILW Process Tank In drum solidification in 3m 3 Drum 3m 3 Drum Final Waste Package HCW Collection tank HCW Drummed Spent Filter Media (LAW) Evaporator Demineraliser Spent Bead Resin HCW Sample Tank Bead Resin Storage Tank Supernate recycling Pre-mixed Cement Grout & Waste Cement Powders 3m 3 Drum 3m 3 Drum Final Waste Package ILW Storage Facility CAD Evaporator Concentrate Sludge CONW Storage Tank LLW Process Tank Fill and Solidify in Container Grouted Third Height ISO Container CAD Collection Tank Sample tank LCW Sump tank HCW Sump tank Wet Solid LLW Process Radioactive Waste Building Laboratory Drains Decontamination Showers Discharge to Environment Process Key Spent Fuel Wet Solid ILW Dry Solid LLW - HVAC Contaminated Effluent CAD Sump tank HCW Sump tank Dry Solid ILW Wet Solid LLW Dry Solid LLW Clean or very low contamination effluent Service Building Outage only Off-site Disposition Page 37 of 255

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53 3.2.1 Gaseous Radioactive Waste Management Systems 128. Gaseous radioactive waste arising in the Power Station will be treated and processed in the following systems: HVAC System; and, The OG System. These systems are shown in Figure 3.3 and are described in the sections below. The figure corresponds to Unit 1 as it includes the discharge from the Rw/B. It is highlighted that the Unit 2 R/B (i.e. main) stack does not receive ventilation from the Rw/B. Figure 3.3 Summary of Gaseous Effluent Generation, Treatment and Disposal Heating, Ventilation and Air Conditioning System 129. The purpose of the radioactive HVAC System (made up of a number of individual HVAC systems) is to: Prevent the uncontrolled discharge of radioactive substances; Provide a pleasant working environment for workers; Ensure optimal working conditions for plant and equipment; and, Deliver safety-related functions to protect workers in the event of a release of radioactivity. Page 39 of 255

54 130. The air pressure in facilities handling radioactive substances is typically maintained by the system at a lower level than atmospheric pressure to ensure that air flows into the facility from the external environment. This prevents the uncontrolled discharge of any radioactive substances through doors, windows and gaps in the building fabric. The system is configured so that air flows into it from areas of lower contamination risk to areas of higher contamination risk, thus minimising the spread of contamination Separate HVAC sub-systems are employed to provide independent functionality to the following buildings/areas where radioactive contamination is present: R/B; Turbine building (T/B); Rw/B, including the Wet-solid ILW (Intermediate Level Waste) and Wet-solid LLW (Low Level Waste) processing facilities; Service Building (S/B); and, Lower Activity Waste Management Facility (LAWMF) The R/B, T/B and the controlled area inside the Rw/B are the three main areas with the potential to generate gaseous radioactive waste streams. Examples of contributors to the radioactivity that may arise and be routed to the HVAC system include airborne particulates from the filter backwash receiver tank in the T/B, and evaporated water vapour from the SFP in the R/B. The S/B contains the change room facilities and the active laboratory, and thus has a limited potential to generate gaseous radioactive waste Separate HVAC sub-systems are also employed in the following areas for environmental control only: ILW Storage Facility; Emergency Diesel Generator Electrical Equipment Area; Heat Exchanger Area; Instrument and Control Power Supply Panel Area inside Control Building; Main Control Room (MCR); Back-up Building Emergency Control Room (ECR); and, Back-up Building diesel generator electrical equipment area With the exception of those serving the S/B, ILW Storage Facility and the LAWMF, each of the above sub-systems from contaminated areas will discharge to the environment via the R/B stack. The ILW Storage Facility and LAWMF HVAC systems will have their own individual standalone stacks. Discharges from both the ILW Storage Facility and LAWMF will be abated via high efficiency particulate air (HEPA) filters prior to release to the environment. HVAC for areas which are outside the radiologically controlled areas, discharge air directly to atmosphere locally without any filtration The HVAC system has been designed to ensure that appropriate abatement, in the form of HEPA filters, will be provided on those sub-systems serving areas with the potential to generate gaseous radioactive waste. The HEPA filter type chosen for each system will depend on the environmental and flow-rate conditions anticipated. The filter types will be confirmed as part of detailed design of the HVAC systems HEPA filters will be changed, where practicable, based on performance determined using continuous measurement of differential pressures or on measured dose rate. This means Page 40 of 255

55 that the filters will not be changed at a predefined frequency which can result in the generation of additional volumes of solid waste Where practicable filters will be used to their design capacity, within the facility Safety Case and in line with the waste management strategy. HEPA filters that are removed from the HVAC system at the end of their life will ultimately be managed and disposed of as solid waste. Further information on the solid radioactive waste management system is given in Section Off-Gas System 138. As described above, radioactive materials in the reactor water can be carried into the steam phase. This steam is transferred to the turbine and auxiliary turbine steam systems 6 and then to the condenser where it is condensed and returned to the reactor water along with any condensable gases present Some of the radionuclides that are carried in the steam do not condense in the condenser. These radionuclides are carried by the Steam Jet Air Ejector (SJAE), which is used to maintain the vacuum in the condenser, and require treatment and disposal as gaseous radioactive waste. The treatment is carried out in the OG System The purpose of the OG System is to: Safely recombine flammable gases (hydrogen and oxygen) which are generated by radiolytic decomposition of the reactor cooling water in order to reduce the possibility of a hydrogen explosion; and, Minimise and control the release of radioactive gases into the atmosphere by delaying the off gas stream thereby enabling short lived radioactive isotopes to decay and filtering out particulate matter The OG System comprises the following principal plant items: SJAE; Preheater; Recombiner; Condenser; Cooler Condenser; Refrigeration Facility; Charcoal Adsorber (Delay Beds); Filter; Blower; Blower after cooler; Delay Bed Heating Ventilation Handling Unit; Ejector; and, 6 Steam from the reactor is also routed to the turbine auxiliary steam and the turbine bypass system. The turbine auxiliary system supplies driving steam to the reactor feedwater pump turbine and the SJAE, and also supplies heating steam to the moisture separator reheaters and the gland steam evaporator. This steam is ultimately routed to the condenser and returned to the reactor as water. Page 41 of 255

56 TGS Filter. The system is illustrated in Figure 3.4. Figure 3.4 Process Diagram for the Off-Gas System 142. Air that leaks into the condenser, other non-condensable gases and the radiolytic H 2 and O 2 produced in the reactor the off-gas is extracted by the SJAE. A two-stage SJAE system is employed within the OG system, with an SJAE condenser between the two stages. The off-gas is then heated in the OG Preheater in order to prevent formation of water droplets that might adversely affect the catalytic performance of the downstream OG Recombiner The OG Recombiner is used to recombine the H 2 and O 2 present in the off-gas by catalytic reaction. This maintains the H 2 concentration below its lower flammability limit. The resulting steam is condensed in the OG Condenser, reducing the volume of the off-gas and cooling it to nearly room temperature. The cooling source is supplied by the R/B Cooling Water System (RCW). The moisture in the off-gas is then further reduced in the OG Cooler Condenser to ensure that the performance of downstream OG Charcoal Adsorber is not compromised. The cooling source for the OG Cooler Condenser is the OG Refrigeration Facility. Page 42 of 255

57 144. The OG Adsorber comprises four 25%-capacity charcoal beds in series. Its purpose is to retain isotopes of xenon for a period of approximately 30 days and isotopes of krypton for approximately 40 hours. Given the half-lives of krypton-85m, krypton-87, krypton-82, xenon-133 and xenon-135, this significantly reduces the amount of these radionuclides discharged to the environment. The use of the charcoal adsorbers contributes to a reduction in the amount of radioactive krypton and xenon gas discharged to the environment to 1/30,000. Experimental evidence is also available to show that iodine and methyl iodide discharges are significantly reduced using charcoal adsorbers A heating ventilation handling unit is used to control the temperature of the OG Charcoal Adsorber Room to within 25±5 C and to ensure the performance to the Charcoal Adsorbers is not compromised. The OG Filter removes particulate matter from the off-gas after it has been processed through the OG Charcoal Adsorber During start-up, a negative pressure is maintained within the OG Charcoal Adsorber by the OG Blower, which also ejects the off-gas to atmosphere via the R/B stack. During power operation and shutdown this function is carried out by the OG Ejector. The OG Blower is, though, available as a back-up to the OG Ejector during power operation and shutdown The OG process equipment is located within the T/B to minimise the length of piping needed to carry the gas from the main condenser, and housed in a reinforced-concrete structure to provide adequate shielding The discharge from the Turbine Gland Exhauster is also shown in both Figure 3.3 and Figure 3.4. The TGS uses water extracted from the CST to produce steam that is subsequently used in the turbine gland seal. Around 98% of the steam in the TGS is condensed in the Steam Condenser and reused in the plant. The residual steam, containing tritium and iodine, is extracted by the exhauster and routed to the R/B stack where it is mixed with discharges from the OG system and HVAC systems prior to being discharged to the environment. Prior to discharge a HEPA filter on the TGS extract removes particulate matter Radiation monitors (and several in-process radiation monitors which monitor gross radiation) are provided on each discharge line in the OG system to monitor the release of the gaseous radioactivity (OG Charcoal Adsorber discharge line, Gland Steam Exhauster 7 and Mechanical Vacuum Pump 8 discharge line). Further information on discharge monitoring is given in Section The OG system generates charcoal waste from the charcoal adsorbers and waste HEPA filters. The charcoal will be treated as Dry Solid LLW at the end of station life. HEPA filters will also be treated as Dry Solid LLW when replaced Liquid Radioactive Waste Management Systems 151. Liquid radioactive waste arising at the Power Station is treated in the following systems: Reactor Water Clean-up System (CUW); Fuel Pool Cooling and Clean-up System (FPC); 7 The gland steam exhauster is part of the turbine gland steam system which supplies steam to the turbine shaft to seal it and prevent air leaking into the condensers. 8 The mechanical vacuum pump is used to create the vacuum in the condenser that draws steam from the turbines. Page 43 of 255

58 Suppression Pool Clean-up System (SPCU); Condensate Water Clean-up System; Low Chemical Impurity Waste System (LCW); High Chemical Impurity Waste System (HCW); and, Controlled Area Drains System (CAD). These systems are shown in Figure Reactor Water Clean-up System 152. The purpose of the CUW is to maintain the quality of reactor water within a predetermined range in order to prevent or where not possible, minimise: Corrosion of equipment and piping in the reactor primary system; Adhesion of impurities to the fuel surface (which decreases heat transfer efficiency); and, Radioactive contamination of the reactor primary system and related equipment There is a CUW system for each of the reactors at the Power Station. An outline of each system is shown in Figure 3.5. Each CUW system continuously draws water from the reactor and passes it through FDs. The demineralisers contain powdered resin which removes both soluble and insoluble impurities (i.e. nuclides in ionic form) from the reactor water, and the filter removes insoluble impurities (i.e. nuclides in particulate form). OPEX data suggests that radionuclides removed include Ag-110m, Co-60, Cr-51, Fe-59, Mn-54, Zn-65 and Sr Once treated, the water is returned to the reactor in the main feed line. The quality of the water passing through the system is continuously monitored to ensure that the characteristics are within defined parameters and that the system is performing as expected In the unlikely event that the characteristics of the liquid fall outside of the defined parameters, the liquid may be passed through the FD again so that the characteristics of the liquid meet the defined parameters. During start-up and shutdown operations, any excess reactor water can be transferred to the LCW to enable an operator to manage the reactor water level of the system. Following treatment, this water is then returned to the CST and is then available to be recycled back into the reactor water circuit The use of the resin allows the radioactivity removed from the reactor water to be disposed of as solid waste once the resin is spent. To ensure that the CUW FD is working effectively, its differential pressure and outlet conductivity are monitored. When the differential pressure or conductivity exceeds a set value, the filter demineraliser will be backwashed A backwash operation involves the filters being water washed, gravity drained and air scoured to remove insoluble impurities. The resin will be discharged to the CUW Backwash Receiver Tank which is emptied into the CUW/FPC Sludge Storage Tank in the Rw/B. The powdered resin will then be processed in the Wet-solid ILW Processing System as described in Section Page 44 of 255

59 Figure 3.5 Outline of the Reactor Water Clean-up System Fuel Pool Cooling and Clean-up System 158. The SFP is constructed from reinforced concrete and fitted with a stainless steel lining. Within the pool, fuel assemblies (including both spent and new fuel, as well as spent control rods) are stored in spent fuel storage racks, boronated stainless steel square tubes which are immersed in demineralised water. The SFP is located adjacent to the reactor and is separated from the reactor well by the SFP gates. Page 45 of 255

60 159. The FPC provides two functions: Removal of decay heat from the spent fuel stored in the SFP; and, Removal of impurities from the fuel pool water An outline of the FPC is shown in Figure 3.6. It is provided with heat exchangers for the removal of decay heat, and FDs for the removal of impurities. These are shared with the SPCU described Section Figure 3.6 Outline of the Fuel Pool Cooling and Clean-up System 161. The SFP water is processed through the FPC and returned to the SFP. The system maintains the quality of the water in the pool by removing the following impurities: Airborne impurities that are deposited into the pool; Impurities arising from the surface of the fuel and components stored in the pool; CPs and FPs transferred from the core during refuelling; and, Residual chemicals used for cleaning or flushing water after pool cleaning. Page 46 of 255

61 162. In order to ensure that the FPC FD is working effectively, differential pressure and outlet conductivity are monitored. When the differential pressure or conductivity exceeds a specific value, the now spent powder resins are discharged to the CUW Backwash Receiver Tank in Rw/B. The spent resins will be processed in the Wet-solid ILW Processing System Suppression Pool Clean-up System 163. The Suppression Pool (S/P) forms part of the primary containment for the reactor in the R/B (this can be seen in Figure 2.1). The S/P supplies a number of systems that introduce water into the reactor pressure vessel during outages, surveillance operations and accident conditions. The S/P can also supply water to the SFP if it requires topping up An outline of the SPCU is shown in Figure 3.7. Its primary function is to provide purifying water treatment for the S/P on an ad hoc basis. The system transfers S/P water through the FPC FD and returns it back to the S/P thereby maintaining the water quality within specified parameters. There is one SPCU for each reactor, located in the R/B. Figure 3.7 Outline of the Suppression Pool Clean-up System Page 47 of 255

62 Condensate Water Clean-up System 165. The Condensate Water Clean-up System, an outline of which is presented in Figure 3.8, uses filtration and demineralisation to treat condensed water that has passed through the turbines and auxiliary turbine systems as steam. Once treated, the water is returned to the reactor in the main feed line. There is one Condensate Water Clean-up System for each reactor at the Power Station and they are located in the T/Bs Filters (condensate filters) are used to remove any solid matter that could either damage the fuel assemblies or become activated in the reactor core. When the differential pressure across the filters increases above a specified value the filters are backwashed to remove insoluble impurities. The crud which is generated is stored in backwash receiver tanks prior to transfer to the Filter Crud Sludge Storage Tanks in the Rw/B for eventual processing via the Wet-solid ILW Processing System. Backwashing of the filters is carried out to maximise filter life time. At the end of life the filters are removed, monitored and prepared for transfer to the LAWMF A condensate demineraliser is used to remove ions that have become activated or have the potential to become activated in the reactor core and deposit on internal surfaces of pipes and vessels, as well as non-radiological contaminants. This system allows radioactivity in the condensate to be disposed of as spent bead resins. The spent bead resins will be processed in the Wet-solid LLW Processing System. Figure 3.8 Outline of Condensate Water Clean-Up System 168. To ensure the condensate demineraliser is working effectively, the conductivity of the outlet is monitored. During outage a check (break through capacity) is performed on the outlet conductivity to determine whether the resins are operating to the required levels. If these levels are not met, the resins are discharged to the LLW Spent Resin Storage Tank in the Rw/B. Page 48 of 255

63 LCW, HCW and CAD Systems 169. The LCW, HCW, and CAD systems are designed to segregate at source, collect and treat the various streams of radioactive and potentially radioactive waste water generated during various modes of UK ABWR reactor and turbine plant operation: start-up, normal operation, shutdown, re-fuelling and maintenance The Primary Circuit and Fuel Pool (i.e. the plant areas containing water that comes into direct contact with irradiated fuel) are operated as far as is practicable as closed loop systems. For this reason, the LCW and HCW systems are designed so that any water leaks and any water drained from the Primary Circuit or Fuel Pool during the various modes of plant operation is captured and appropriately treated so as to remove both soluble and insoluble impurities. This provides high purity water that is normally recycled for use in the Primary Circuit and/or Fuel Pool. Recycled/treated water is only occasionally discharged to the environment when the recycled volumes exceed Primary Circuit and Fuel Pool water make-up requirements Both generating units share the LCW, HCW and CAD systems, which are housed in the Rw/B. Treatment of the segregated liquid waste is determined based on the properties of the waste. A further subsystem, the Radioactive Drain Transfer System is used to transfer waste water collected in the controlled areas in the R/B, Rw/B, T/B and S/B into collection tanks housed in the Rw/B. All of these subsystems are shown in Figure 3.9 and described further below. Radioactive Drain Transfer System 172. The Radioactive Drain Transfer System is used to transfer waste water, collected in the controlled areas in the R/B, T/B and S/B, into collection tanks in the Rw/B. The system comprises sump tanks, sump pumps, piping, valves, and appropriate instrumentation. In general, waste water is segregated and collected at source as follows: Equipment Drains. These collect radioactive or potentially radioactive waste water from the Primary Circuit system and equipment in the R/B and T/B (including from reactor blow downs during outages), from the SFP system and equipment, and also from systems and equipment in the Rw/B. This waste water is generally expected to have low levels of impurities and, therefore, is normally automatically pumped from the equipment drain sumps to the LCW Treatment System Collection Tank. Chemical Drains. These drains collect the chemical waste generated at the laboratory in the S/B, and water from the condensate demineraliser drains. The waste water generated at the laboratory is collected into the Drain Sump and is automatically pumped to the HCW system. Waste water drained from the condensate demineraliser is routed into the HCW collection tanks. CAD. These collect waste water from other systems (e.g. local HVAC systems) in the Radiation Controlled Areas (RCA) in the R/B and T/B. The waste water collected from the RCAs is potentially contaminated. This water is automatically transferred to the CAD Collection Tank. Page 49 of 255

64 Figure 3.9 Overview of the LCW, HCW and CAD Systems Page 50 of 255

65 Low Chemical Impurity Waste System 173. The LCW Treatment System is housed in the Rw/B and is one of two systems (the other being the HCW system) which are used to treat radioactively or potentially radioactively contaminated waste water. The treated waste waters are returned to the CST for reuse The LCW system is designed to allow the efficient treatment of relatively large volumes of waste water containing low levels of both insoluble and soluble impurities. This liquid effluent is differentiated from that handled by the HCW system in that it has a lower level of chemical impurities, low suspended solids and therefore lower conductivity An outline of the LCW treatment stages is given in Figure Inputs to the system are shown in Figure 3.9. The system processes liquid effulent from: Equipment drains (which includes T/B LCW sump, CUW blow down, reactor well drain, R/B LCW sump etc.); Equipment blow down water (waste generated only in the case of periodic inspection); and, SFP, FPC and CUW. Figure 3.10 Outline of the LCW System 176. During normal operations, the maximum flow rate is expected from decant water from the Filter Crud Storage Tank. The volume of water processed by the system increases during periods of outage. The maximum expected flow rate during outage is from the Reactor Clean-up System blow down LCW is collected in one of four collection tanks, located in the Rw/B. One collection tank will be receiving effluent while a second one is discharging to the downstream treatment process. The third and fourth tanks will receive effluent in turn. All four tanks will be operational during outage when maximum flow is expected Processing within the LCW System is by filtration followed by demineralisation: two filters are used to remove insoluble impurities in the LCW, following which, to remove soluble impurities, the water is passed through two parrallel mixed bed demineralisers packed with bead type ion exchange media Treated water is collected in sample tanks, where a representative sample of the water is analysed to confirm it meets the criteria for re-use in the reactor. If the treated water does not meet the appropriate criteria, it can be routed back to the LCW Collection Tank and the treatment process repeated (potentially multiple times) until the criteria are met. Once the treated water has been confirmed to meet the appropriate criteria it is sent to the CST for reuse as reactor Primary Circuit or Fuel Pool make-up water The acceptance criteria for the CST will be controlled in accordance with the Water Quality Specification (WQS). This will include parameters such as conductivity, ph and total organic carbon (TOC). There is no radioactivity criterion for water treated in the LCW system (i.e. for reuse via the CST), as the system is operated solely on the level of insoluble Page 51 of 255

66 and soluble impurities within the water. Recirculation and re-use (via the CST) ensures that there is no discharge to sea other than via the HCW system The LCW System generates a number of waste streams: When the differential pressure across a filter increases above a certain point, the filter is backwashed (water washed, gravity drained and air scoured) to remove the particulate. The crud generated is transferred to the Filter Crud storage tanks within the Rw/B before being transferred and processed through the Wet-solid ILW Processing System. The filters are removed from the filter vessel, monitored and prepared for transfer to the LAWMF; and, The demineraliser resins are changed when the outlet conductivity exceeds a set value. The spent bead resin is discharged from the demineraliser vessel and held in the spent bead resin storage tanks before being transferred and processed through the Wet-solid LLW Processing system. The LCW bead resin has the potential to be cross-boundary LLW/ILW waste at arising and is discussed in greater detail as part of Wet-solid Waste Storage below. Further information on cross-boundary LLW/ILW waste is provided below under Wet-solid Waste Storage heading. High Chemical Impurity Waste System 182. The HCW system is also located in the Rw/B and is the second system used for the treatment of radioactively or potentially radioactively contaminated waste water. There is one HCW system for both generating units. The HCW system is designed to allow the efficient treatment of waste water containing high levels of both insoluble and soluble impurities An outline of the HCW treatment stages is given in Figure 3.11 and inputs to the system are shown in Figure 3.9. The system processes liquid waste from: Rw/B HCW sump; Chemical drains; and, Equipment blow down water (waste generated only in the case of periodic inspection). Figure 3.11 Outline of the HCW System 184. During normal operations, the maximum expected flow rate to the HCW is from the S/B HCW sump. During outage, the main source of liquid effluent sent to the HCW system is the condensate demineraliser bottom drain At the start of the HCW system, liquid is collected in two collection tanks. Each tank has sufficient capacity for periods of outage. The collection tanks act as a pair: one is filled whilst the other discharges into the downstream treatment process. Page 52 of 255

67 186. The HCW system contains: An Evaporator for the removal of impurities from the contaminated waste water by distillation; A Demineraliser for the subsequent removal of solubles; and, HCW Sampling Tanks. Effluent is treated in the HCW system in batches. Filtration is unnecessary as the evaporator retains solid matter in the concentrate Within the HCW treatment process the contaminated water is initially evaporated to form a concentrated sludge. The evaporated water is collected in the vapour phase, condensed and routed to the HCW Distilled Water Tank. The concentrated sludge is transferred from the Evaporator to the Concentrated Waste Tank, prior to being transferred for treatment in the Wet-solid LLW Processing System The water collected in the HCW Distilled Water Tank is then passed through a mixed bed Demineraliser packed with bead type ion exchange media. This removes any contaminants which could potentially be carried over from the Evaporator, and ensures that the concentration of fission and activation products in the water is sufficiently low in the event that the HCW is required to be disposed of to the environment Treated water is collected in a Sample Tank where a representative sample is analysed to confirm it meets the criteria for re-use in the reactor. If the water does not meet the criteria, it is returned to the HCW Collection tank and the treatment process repeated (potentially multiple times until the criteria are met) Once the treated water meets the criteria, the water is discharged to the CST for reuse as reactor Primary Circuit or SFP make-up water. Only if the treated water volumes exceed Primary Circuit and SFP water make-up requirements is the treated water routed to the main discharge line and discharged to the environment. On these occasions the treated water is first sampled to ensure that residual levels of radioactive and chemical contamination are within set discharge limits/conditions It is important to note that there is no radioactivity criterion for water treated in the HCW system, as the system is operated solely on the level of insoluble and soluble impurities within the liquid effluent. The acceptance criteria for the CST (for reuse as reactor Primary Circuit or SFP make-up water) will be controlled in accordance with the WQS The HCW generates a number of waste streams: The concentrated sludge from the Evaporator mentioned above; and, The bead resins from the HCW demineraliser will be replaced when the outlet conductivity exceeds a set value. The spent bead resin is discharged from the demineraliser vessel and held in the spent resin storage tanks before being transferred and processed through the Wet-solid LLW Processing System. Page 53 of 255

68 Controlled Area Drains System 193. The CAD System collects water from: Local air-conditioning system drains in the R/B and T/B; and, Potentially contaminated drains of various equipment systems in the controlled areas of the R/B and the T/B. Liquid waste in the CAD System is not expected to be radioactive but has the potential to be contaminated The system is comprised of two liquid waste collection tanks, collection pumps, piping, valves and measuring and control equipment. The collected waste water is pumped to the CAD Collection Tank where it is sampled to confirm it contains no significant radiological contamination (or unacceptable chemical contamination). Water which meets the discharge criteria will be routed to the main discharge line and discharged to the environment. If the water is found to contain any significant radiological contamination or unacceptable chemical contamination, then the operator routes the water to the HCW system for treatment. Wet-solid Waste Storage 195. The above described water treatment systems generate the following secondary wet-solid wastes: Spent bead (ion exchange) resins from the LCW and HCW, and f(cd) in the T/B; Sludge (crud); Concentrated sludge waste from the HCW; Powder resins from the CUW and FPC in the R/B; and, Condensate Filter sludge (crud) from the T/B. The routing of wet-solid waste arisings is shown in Figure The wet-solid secondary wastes are stored in tanks before being transferred to the Wetsolid LLW Processing System, or Wet-solid ILW Processing System for treatment and packaging (e.g. cementation, etc.) in preparation for either disposal off site (in the case of LLW or VLLW), or for transfer to the on-site storage facility (in the case of ILW) The spent bead demineraliser resins from the CD and LCW systems are identified as potential LLW/ILW cross boundary waste 9 as defined in LLWR Limited (LLWR) guidance [RD18]. Co-storage and processing of the CD, LCW and HCW bead resin arisings requires the bead resin storage tank inventory to be managed as potential LLW/ILW cross boundary waste. 9 LLW/ILW cross boundary waste can be defined as ILW and LLW with a concentration of specific radionuclides that prohibits or significantly challenges its acceptability at existing and planned future disposal facilities for LLW, that could be practicably managed as LLW (on the basis of radiochemical and physicochemical properties) through application of some treatment process or decay storage. Page 54 of 255

69 Figure 3.12 Outline of Waste Stream for Wet-Solid Waste Page 55 of 255

70 198. Subject to discussion and agreement with the LLWR following their waste enquiry process, appropriate treatment and decay storage will render the cross boundary bead resins disposable as LLW. A storage period of approximately eight years is envisaged. However, this will depend on the radioactivity of the resin at arising A robust monitoring system to calculate and measure the condition of the CD and LCW bead resins during operation will be maintained such that the spent resin will be within acceptable parameters based on the waste management facility safety case and disposal Waste Acceptance Criteria (WAC) specified by external parties Solid Radioactive Waste Management System 200. The Power Station has a number of facilities and systems that receive, sort and process all dry and wet-solid LLW, ILW and High Level Waste (HLW) streams resulting from UK ABWR operation. The facilities and systems include the following: LAWMF; Wet-solid LLW Processing System; Wet-solid ILW Processing System; ILW Storage Facility; and, Spent Fuel Storage Facility (SFSF). These facilities are described below Lower Activity Waste Management Facility 201. The LAWMF is a single facility which serves both generating units. Its purpose is to receive and process the following dry LLW generated at the Power Station: Heterogeneous (LLW/VLLW); HVAC filters (LLW); and, Filter Media (LLW) Prior to its transfer to the LAWMF, checks will be performed in line with operational procedures to ensure that the waste complies with the facility WAC. If the waste is shown to be compliant it will be collected and transferred to the facility using the on-site transporter. The waste will then be inspected on receipt to ensure it remains in a compliant state following on-site movement. If a problem is identified with the waste package, it will be quarantined under a non-conformance process whilst a suitable course of action is determined The LAWMF will ensure that packaged waste meets the off-site treatment/disposal supplier WAC (including transport requirements). Wastes will be processed according to the WAC for their subsequent off-site treatment or disposal and placed in the appropriate form of transport container for consignment. The relevant safety and quality assurance checks will be performed and consignment paperwork raised. In some instances wastes will be accumulated in buffer storage areas until such time as there is sufficient waste to make up a transport consignment. Where wastes are accumulated, a dose assessment will be undertaken to ensure that any defined dose limits are not exceeded and are shown to be ALARP. Page 56 of 255

71 204. The LAWMF will also receive the following immobilised wastes from the Wet-solid LLW Processing System, packaged into Third Height International Organization for Standardization containers (THISOs) containers in the Rw/B: Bead resin from the LCW and HCW demineralisers and the condensate demineraliser in the T/B; and, Concentrates (sludges). The containers will be consigned from the LAWMF for direct disposal to LLWR Any potentially radioactive liquid effluent generated in the LAWMF will be low volume, low radioactivity from personnel hand washing and emergency shower facilities in radiological controlled (designated areas). The water will be collected, monitored and transferred by suitable means to either the LCW or HCW The LAWMF is provided with a HVAC system which exhausts to atmosphere via a small stack which discharges at a height of 3 m above the roof ridge Wet-solid LLW Processing System 207. The Wet-solid LLW Processing System is housed in the Rw/B and serves both generating units. The purpose of the system is to process the following wet-solid LLW received from the Rw/B storage and waste tanks: Bead resin (LLW); and, Concentrated sludge (LLW). The waste will be processed by cement immobilisation directly into THISOs that conform to LLWR specifications The treatment process uses a waste/cement formulation that is proven to produce a compliant, disposable LLW waste form. The waste is mixed with cement based grout and transferred to the THISO container. The waste/cement mixture cures in the THISO. Once the wastes have been processed into THISOs, the containers are transported to the LAWMF It is anticipated that wet-solid LLW will be processed on a campaign basis and the frequency of the campaigns will be confirmed in due course. The Rw/B storage and waste tanks have the capacity to store arisings of cross boundary LLW/ILW bead resin for an appropriate decay period prior to processing. Concentrated sludge will be accumulated and stored based on campaign periods anticipated to be approximately ten years The Rw/B HVAC system discharges gaseous emissions from the system via the R/B stack Wet-solid ILW Processing System 211. The Wet-solid ILW Processing System is housed in the Rw/B and serves both generating units. The purpose of the system is to process the following wet-solid ILW received from the Rw/B storage tanks: Sludge (crud) (ILW); and, Powder resins (ILW). Page 57 of 255

72 212. The resins and sludges are transferred to each facility via pipelines as dilute slurries. The waste is then immobilised in cement in 3 m 3 drums using in-drum mixing of waste and cement The packages are transferred to the ILW Storage Facility and will ultimately be disposed of to the Geological Disposal Facility (GDF). It is currently assumed by Horizon that a GDF will not be available to receive ILW from the Power Station until completion of the Decommissioning Phase ILW Storage Facility 214. There is a single ILW Storage Facility at the Power Station. Its purpose is to store ILW packages received from the Wet-solid ILW Processing System prior to their disposal to the GDF. The storage facility is designed to hold all of the processed ILW generated in the operating lifetime of the ABWR, i.e. 60 years, and to control environmental conditions such that any risk of degradation of the packaged wastes by corrosion or similar processes during the storage period is minimised, so far is as reasonably practicable Radioactive gaseous emissions are not expected from the facility, although trace discharges of radioactivity cannot be ruled out. HEPA filtration of the exhaust air is thus included as a precautionary measure should any waste package containment failure occur The only expected liquid arisings from the ILW storage Facility is from the dehumidifiers in the HVAC system. This effluent is disposed of by the HCW system Spent Fuel Storage Facility 217. The design of the SFSF is in development and yet to be confirmed (it is likely that the construction of the facility will be deferred beyond the construction of the first reactor). It will be located in the south west corner of the site. At present it is assumed there is no requirement for a ventilation stack and any gaseous radiological discharge from this facility will be negligible As well as storing spent fuel, the facility will also store HLW casks received from the R/B prior to their disposal to a GDF. Following a period of up to ten years storage in the SFP, HLW waste control rods and reactor components will be stored in this facility until the end of operations. The wastes will then be will be retrieved, size reduced, re-packaged and conditioned in suitable waste containers in a future Dry Solid ILW Processing Facility immediately prior to disposal to the GDF Summary of Solid Radioactive Wastes and Treatment 219. A summary of the solid radioactive waste streams that will arise from the operational stage of the Power Station and the current preferred processing arrangements for these wastes is given in Table 3.6. Whilst not expected as part of normal operations, oil and radiologically contaminated land may be generated as a result of unplanned incidents. Therefore, these waste types have been included in the table for completeness. Page 58 of 255

73 Table 3.6 Radioactive Waste and Spent Fuel Streams Arising from the Power Station No. Title Description Category Form Preferred Processing and Disposal Arrangements 1 Heterogeneous The wastes include metals, hard and soft wastes, inert wastes and organic materials (including cellulosics), specifically plastics, paper, card, wood, glass, building materials, insulation, motors, cables and pipes, miscellaneous filters and strainers. LLW (including VLLW) Solid Heterogeneous LLW is managed in compliance with waste hierarchy application and disposed of in accordance with BAT, which will be case-specific. Wastes are segregated at source in compliance with Horizon s internal WAC, and characterised to assist in determining the appropriate disposal route. Waste packaging is in compliance with the WAC of the receiving facility and typically includes the standard range of LLW containers used in the UK. Metals are segregated for metal melting treatment where reasonably practicable. Soft wastes are segregated for volume reduction by compaction (either on-site or off-site) or incineration (off-site only) where reasonably practicable. Wherever reasonably practicable (and demonstrated to be BAT) wastes may be sentenced as VLLW and transferred to appropriately permitted landfill sites. 2 HVAC Filters Arising from filter changing in air treatment facilities from exhausts from Reactor (R/B), Turbine (T/B), Radwaste (RW/B) and Service (S/B) Buildings as well as waste treatment facilities. LLW Solid On removal the filter module is loaded into suitable containers which are transferred to the LAWMF. From there they are sent for disposal to LLWR either directly or where reasonably practicable via a UK treatment facility for volume reduction or incineration. Where reasonably practicable (and demonstrated to be BAT) filters may be sentenced as VLLW and transferred to an appropriately permitted facility. 3 Bead resin Arising from the CD, LCW and HCW demineralisers; Styrene divinylbenzene copolymer matrix. LLW (identified as cross boundary wastes) Wet Wastes are initially transferred to tanks in the Rw/B for decay storage prior to processing. Wastes are then transferred to the Wet-solid LLW Processing System for cementation within THISO containers, followed by transport to an appropriately permitted disposal facility. 4 Concentrates (Sludges) Arising from the HCW evaporator comprises particulate and dissolved species. LLW Wet Transfered to the Wet-solid LLW Processing System for cementation within THISO containers, followed by transport to an appropriately permitted disposal facility. Page 59 of 255

74 No. Title Description Category Form Preferred Processing and Disposal Arrangements 5 Spent Filter media Filters are used to remove insoluble impurities in the condensate water clean-up system and LCW system and are therefore effectively wet solid LLW. However, as they are not in a particulate form they are managed as a dry solid waste. 6 Sludge (crud) Arising from backwashing of various filters from the CF and the LCW systems. 7 Powder resin Arising from the CUW and FPC filter demineralisers; cross linked polystyrene matrix. Contains particulate corrosion product. LLW Solid On removal the filter module is loaded into suitable containers. These are transferred to the LAWMF, followed by disposal to LLWR, either directly or where reasonably practicable via a UK treatment facility for volume reduction or incineration. ILW Wet Co-packaged with the spent ILW powder resin and solidified in cement within 3 m 3 unshielded stainless steel drums. ILW Wet Co-packaged with the ILW sludge (crud) and solidified in cement within 3 m 3 unshielded stainless steel drums. 8 Control rods The UK ABWR has two types of control rods, hafnium rods for operational control of the reactor, and boron carbide rods used for shutdown of the reactor. Spent hafnium control rods will be generated at a rate of about 5 per year for the lifetime of the reactor. Boron carbide rods have an expected life of 40 years and will therefore need to be replaced after this period. HLW (ILW at disposal) Solid Initial storage in the SFP followed by packaging in suitable containers for transport to the SFSF for long term storage. The control rods will be retrieved, size reduced, re-packaged and conditioned in suitable waste containers in a future Dry Solid ILW Processing Facility immediately prior to disposal to the GDF. 9 Damaged Fuel Channels (Channel Boxes) Zircaloy box which surrounds the fuel bundle. Approx. 4.3 m long and cm square. Individual channel boxes may arise on occasion during the Generation Phase if damaged or distorted. It is assumed that in these instances the fuel assembly would be loaded to a replacement channel box. HLW Solid Storage with spent fuel in the SFP, followed by packaging in suitable containers for transport to the SFSF. The fuel channels (channel boxes) will be retrieved, size reduced, re-packaged and conditioned in suitable waste containers in a future Dry Solid ILW Processing Facility immediately prior to disposal to the GDF. 10 Reactor Components These activated metallic components arise from the operation of the reactor and include the Start-up Range Neutron Monitor (SRNM) System (including dry tubes), Local Power Range Monitor (LPRM) systems, Traversing in core Probes (TIPs) and blade guides. HLW (ILW at disposal) Solid The preferred strategy for managing activated metals is a period of decay storage in the SFP followed by packaging in suitable containers for transport to the SFSF. The reactor components will be retrieved, size reduced, repackaged and conditioned in suitable waste containers in a future Dry ILW Processing Facility immediately prior to disposal to the GDF. Page 60 of 255

75 No. Title Description Category Form Preferred Processing and Disposal Arrangements 11 Spent Fuel Used fuel assemblies along with Fuel Channels (Channel Boxes). 12 Oil It is possible that during the operational life of the plant accidental spillages or leakages of oil could become radiologically contaminated. For the purposes of this Application it is assumed that any radioactive oil would be contaminated (not activated) and of low volume. Spent fuel Solid Initial decay storage in the SFP followed by packaging into suitable containers and shielded concrete overpacks for transport in a heavily shielded cask to SFSF for long term storage. The spent fuel will be disposed of to purpose built spent fuel disposal facilities within the GDF. While the R/B SFP is operational, it is assumed that inspection/repackaging of spent fuel would occur in the SFP facility. The decontamination liquid waste generated as a result of decontamination of the casks is collected in the LCW sump in the R/B via drainage piping and is then transferred to LCW system for treatment. The same process applies to all wastes stored in HLW casks. LLW Wet The preferred strategy for the processing of contaminated oil is incineration if it is within the receiving facility s WAC. If the waste is outside the acceptance criteria consideration will be given to filtration and/or decay storage. If radioactively contaminated oils need to be disposed of they will be converted to a solid form. In this event Horizon would engage with industry best practice through consultation with NDA/RWM to determine the most appropriate method. Solidified oils would be disposed of to the most appropriate permitted disposal site in line with a demonstration of BAT and application of the waste hierarchy. 13 Radiologically Contaminated Land Examples of where radiologically contaminated soils may arise as a waste product include: If there are areas of contamination resulting from the historic use of the site or the Existing Power Station; and, As a result of spills or other accidental releases during operation. LLW (Including VLLW) Solid The preferred strategy for treatment of radioactively contaminated soil is, upon excavation, to characterise the waste and package it directly into containers suitable for disposal to an appropriately permitted UK off-site disposal facility. Wherever reasonably practicable, radioactively contaminated soil will be removed, packaged by SQEP contractors and dispatched as it arises. Page 61 of 255

76 3.3 Radioactive Waste Discharge Points and Disposal Routes Gaseous Radioactive Waste 220. Radioactive gaseous waste will be discharged to atmosphere via the following two principal routes: Unit 1 R/B stack; and, Unit 2 R/B stack Unit 1 R/B stack serves to discharge radioactive waste that has been extracted via the following systems during normal operations: Unit 1 R/B HVAC system; Unit 1 T/B HVAC system; Unit 1 OG system; and, Rw/B HVAC system. Unit 2 R/B stack serves to discharge radioactive waste that has been extracted via the corresponding systems in Unit 2, although it does not handle ventilation from the Rw/B HVAC system. The characteristics of Unit 1 and Unit 2 R/B stacks are summarised in Table 3.7. Table 3.7 R/B Stack Characteristics and Discharge Parameters Parameter Unit 1 - Values Unit 2 - Values Stack height (above ground level) 75 m 75 m Stack diameter 3.15 m 3.15 m Stack exit velocity (summer) 31.1 m 3 /s 23.9 m 3 /s Stack exit velocity (winter) 30.1 m 3 /s 23.2 m 3 /s Volumetric flow rate (summer) 242 m 3 /s 186 m 3 /s Volumetric flow rate (winter) 235 m 3 /s 181 m 3 /s Exhaust temperature (summer) 39.9 C 39.9 C Exhaust temperature (winter) 30.6 C 30.6 C Relative humidity (summer) 29.5% 29.5% Relative Humidity (winter) 9.6% 9.6% 222. There are four additional discharge points: S/B: The HVAC discharges are released to the atmosphere via a vent on the S/B roof; SFSF: At present it is assumed there is no requirement for a ventilation stack and that any radiological discharge from this facility will be negligible; ILW Storage Facility stack: Radioactive gaseous discharges are not expected from the facility. However, a small stack may be required 3 m above the ridge line to aid ventilation of the facility to control environmental conditions. Should any discharges be released these will be negligible; and, Page 62 of 255

77 LAWMF stack: Gaseous discharges will be made from this stack due to the generation of airborne radionuclides during LLW processing operations. References for each of the above gaseous radioactive waste outlets are given in Table 3.8. Table 3.8 Gaseous Discharge Outlets Reference Numbers Parameter Unit 1 R/B Stack Unit 2 R/B Stack S/B Vent SFSF Stack ILW Storage Facility Stack LAWMF Stack Potential Fugitive Discharges Reference No. RA11 RA21 RA01 RA02 RA03 RA04 RA As described earlier in the section, air pressures in facilities handling radioactive substances are typically maintained at a lower level than atmospheric pressure to ensure that air flows into the facility from the external environment. This prevents the uncontrolled discharge of any radioactive substances through doors, windows and gaps in the building fabric. In spite of this, the potential for fugitive release of gaseous radioactivity from such openings cannot be completely ruled out. It is therefore proposed that openings giving rise to potential fugitive discharges are defined as disposal outlet reference RA Liquid Radioactive Waste 224. Liquid radioactive waste will be discharged via the main cooling water outfall located adjacent to the Existing Power Station outfall, to the south west of Wylfa Head. This is the only discharge route for liquid radioactive waste during normal operations. The cooling water tunnels which feed into the main cooling water outfall are defined as references WW11 (Unit 1) and WW21 (Unit 2) Solid Radioactive Waste 225. The waste classifications and the waste strategies which have been developed by Horizon to manage its solid radioactive waste are summarised in Figure Subject to the application of BAT (which will be frequently reviewed) and pending the waste from the Power Station meeting the relevant WAC for the permitted receiving facility, the preferred transfer routes for different waste types are: Off-site transfer of metallic waste for treatment and recycling; Off-site transfer of waste for incineration; Disposal of LLW (VLLW) at a suitably permitted landfill, where the above routes have been shown not to be BAT; Disposal of LLW at LLWR where the above routes have been shown not to be BAT. This waste may be compacted prior to disposal if compaction is demonstrated to be BAT for the waste; and, Page 63 of 255

78 Disposal of LLW via the routes available through the LLWR waste treatment services framework 10. These transfer routes are described further in the sections below The wastes will be required to comply with strict WAC of the site receiving the waste for processing and/or disposal. These WAC will be derived from the limits and conditions imposed by that site s environmental permit. Horizon will ensure that necessary contractual and procedural arrangements are agreed with the relevant waste service providers to ensure that wastes are transferred in compliance with the relevant WACs and will only transfer radioactive wastes to sites that are suitably authorised to receive such wastes It should be noted that an active laundry will not form part of the Power Station design. Radiation protection clothing will be sent off-site for cleaning It has been assumed by Horizon that spent fuel generated through the operation of the Power Station will not be reprocessed. Horizon s IWS considers that spent nuclear fuel, in the form that it is recovered from the reactor and placed into storage, is not waste. However, once it has been recovered from storage and packaged into its final disposable form yet to be defined by Radioactive Waste Management Limited (RWM) it will be regarded as HLW and will be disposed of to the GDF For spent nuclear fuel and Higher Activity Waste (HAW), Horizon s strategy includes unimmobilised storage with processing to a final disposable form to occur at a later stage. This will enable Horizon to keep its disposal options open in relation to the GDF and possible future routes for HAW Off-site Transfer of Waste Transfer of Metallic Waste for Treatment and Recycling 230. The disposal of metallic waste will preferentially be by transfer to an appropriately permitted metals treatment facility. Metals treatment facilities will use a range of techniques such as size reduction, shot-blasting, and melting to enable recycling of radioactive metallic waste. Secondary waste arisings from the process, such as shot blast media or slag from metal melting, will be further treated or disposed of via the most appropriate route directly from the metals treatment facility. No radioactive materials will be returned to the Power Station Materials such as carbon steel, stainless steel, aluminium, brass, copper and lead as well as other less common metals from the UK nuclear industry have been successfully treated in the past. Metallic waste treatment can achieve up to a 98% volume reduction of the original waste consigned. 10 LLWR provides customers with a range of characterisation, logistics and waste services through an agreed common contractual arrangement. This is known as the waste services framework. Page 64 of 255

79 Figure 3.13 Radioactive Waste Management Strategies and Waste Hierarchy 232. On site segregation will be employed to ensure the most effective use of the metals recycling route for metallic waste arising from the Power Station in order to optimise the volumes of metallic waste that can be recycled. Page 65 of 255

80 233. There are currently two major metallic waste treatment services in the UK: Metal Recycling Facility, Workington, Cumbria; and, Tradebe Inutec, Winfrith, Dorset. However, it should be noted that the services on offer in the UK are limited to surface decontamination. There are also additional service providers outside of the UK which will be considered when determing the optimised disposal route. Transfer of Waste for Incineration 234. The disposal of combustible waste will preferentially be by off-site incineration. Incineration is a mature technology which has been utilised for radioactive waste management for a number of years. Combustible wastes are treated by incineration techniques to reduce their volume There are a number of routes within the supply chain and via the LLWR waste treatment services framework which are capable of managing wastes from the nuclear industry. These routes utilise advanced high temperature incineration technology and operate under strict limits defined and monitored by the environment agencies. Where practicable the operators of incinerators seek to use the heat generated in their processes to generate electricity from waste Incineration will be applied to a number of waste streams arising from the Power Station including heterogeneous LLW such as contaminated personal protective equipment (PPE), monitoring swabs, plastics, paper and combustible wastes such as plastic sheets, paper, wood, cloth and oil There are currently four major incineration services in the UK as listed below. There are also additional service providers outside of the UK. Fawley Thermal Treatment Centre, Southampton; Kent High Temperature Incinerator, Sandwich in Kent; Colnbrook in Berkshire; and, Ellesmere Port in Cheshire On site segregation will be employed to ensure the most effective use of the incineration route for combustible waste arising from the Power Station in order to optimise the volumes of combustible waste that can be incinerated. Transfer of LLW (VLLW) for Disposal at a Permitted Landfill 239. The disposal of LLW (VLLW) may be by transfer to a suitably permitted landfill. Where the waste is not suitable for metal recycling or incineration, it may be demonstrated that disposal to landfill is BAT. This option is generally suitable for lightly contaminated, low risk waste that does not require the same degree of engineering protection provided by the LLWR. This waste typically includes rubble, soil, lagging and glass There are currently three landfills that are permitted to accept disposals of LLW. These are: East Northants Resource Management Facility, Kings Cliffe in Northamptonshire; Clifton Marsh Landfill site, Lancashire; and, Lillyhall landfill site in Cumbria. Page 66 of 255

81 241. Waste from the Power Station that is disposed of to landfill will meet the WAC for the landfill and the standards of the landfill directive [RD19]. In order to make the most effective use of the landfill disposal route, Horizon will ensure that the waste requiring disposal is effectively characterised. Transfer of LLW for Off-site High Force Compaction 242. The disposal of LLW may include high force compaction (also known as super compaction) prior to disposal. This route will be utilised for waste that is not suitable for metal recycling, incineration or landfill. High force compaction with subsequent disposal to a suitably permitted disposal facility may be demonstrated to be BAT for these compactable wastes There is one commercially available high force compactor in the UK: Tradebe Inutec at Winfrith. Waste will be compacted and sent directly to LLWR, no waste will be returned to the Power Station Whilst volumes of waste being routed for high force compaction are declining (as some compactable material is being diverted to incinerators), it is recognised that the ability to treat waste in this way remains an important option to minimise the volume of waste that ultimately requires disposal at LLWR. Therefore, Horizon will utilise the use of high force compaction in those cases where it is demonstrated to be BAT. Transfer of LLW to LLWR for Disposal 245. Finally, the disposal of LLW may be by transfer to LLWR. Disposal to LLWR will be required primarily for processed bead resins, concentrates (sludges) and HVAC filters. Other wastes may also be disposed of via this route if this option for their disposal is demonstrated to be BAT Once waste has been consigned to LLWR, it (or the Service Supplier LLWR sub-contracts to undertake disposal or incineration) will take ownership of the waste and be responsible for disposal of all secondary wastes generated from its processes LLWR Acceptance in Principle 247. LLWR offers a well-recognised waste management service for the whole UK nuclear industry. The service is currently adopted by many operational sites in the UK and also by sites that are being decommissioned The LLWR waste services framework provides access to the disposal routes described in the above sections. As part of the GDA of the UK ABWR a letter was received from the LLWR giving Acceptance in Principle for the LLW described in this Application. Conditions noted in the letter will need to be met at the time of waste arising. These include: New waste enquiry forms will be required to be submitted at the time of waste arising; Treatment/disposal at the LLWR is supported by a suitably underpinned BAT justification; The waste complies with the WAC in force at the time and meets the UK policy for the long term management of solid LLW (<4 GBq/t alpha and <12 GBq/t other radionuclides); The waste stream is adequately characterised; and, The waste stream possesses a UK Radioactive Waste Inventory identifier. Page 67 of 255

82 Disposability of ILW, HLW and Spent Fuel 249. Radioactive Waste Management Limited (RWM) is a wholly owned subsidiary of the Nuclear Decommissioning Authority (NDA) and is responsible for delivering the GDF. It operates a process of disposability assessment for wastes that require disposal in a GDF in order to minimise the risk that the conditioning and packaging of such wastes now will result in packages that are incompatible with geological disposal in the future Horizon is required to obtain a view from RWM as to whether the ILW, HLW (to be disposed of as ILW) and spent fuel arising from the Power Station can be disposed of in line with any plans for a GDF in the UK. For the GDA, Hitachi-GE Nuclear Energy, Limited (HGNE) provided RWM with a suite of information in support of Disposability Assessment Submission for ILW, HLW and spent fuel to allow RWM to undertake its assessments RWM concluded that sufficient information had been provided to produce valid and justifiable conclusions under the GDA Disposability Assessment process. Both spent fuel and ILW from the operation and decommissioning from a UK ABWR should be compatible with plans for transport and subsequent disposal and the assessment process has not identified any significant issues that challenge fundamental disposability Horizon will undertake formulation development work in the future to ensure that the packaged waste volumes are minimised as far as is practicable, and to support RWM in a disposability assessment through a Letter of Compliance (LoC) Submission. Initial stakeholder engagement with RWM has commenced and the plan and timings for LoC submissions will be developed and confirmed at a later date [FA RSR-2] Variation of Discharges During Reactor Cycle Stages 253. This section describes the variation in discharges during the reactor cycle stages. The cycle stages during commercial operation are as follows: Start-up: The process of start-up commences with the reactor in a cold shutdown condition. The plant is heated and finally nuclear power is raised to the point where electrical power is generated; Power operation: The mode during which the plant is at rated power. This is based on an 18 month reactor cycle with 17 months at power 11. This mode has by far the longest duration at approximately 517 days; Hot Stand-by: The condition in which the plant is maintained in high temperature and pressure, and the core is maintained in a sub-critical condition; Shutdown: The shutdown process takes the plant from hot stand-by to a cold shutdown condition (reactor coolant < 100 C) and is the reverse process to start-up; and, Outage: Once the plant is in a cold shutdown condition, re-fuelling operations may begin. This involves the flooding of the reactor cavity, removal of the vessel head and upper internals (steam separator and dryer) and removal of fuel to the SFP. New fuel is introduced and the plant is made ready for start-up. For the UK ABWR, this mode is anticipated to last for approximately 30 days. 11 It should be noted that this Application has assumed an 18 month fuel cycle in line with the information presented in GDA. However, the actual cycle length which will be implemented will be decided by Horizon based upon balancing factors such as fuel burn up and radioactive waste produced. Section 4 provides further information. Page 68 of 255

83 254. The reactor cycle stages have an effect upon the generation of certain radionuclides: Temporary increases in corrosion product activity concentrations are observed in the reactor water circuit during major changes of state of a reactor system, e.g. at shutdown. This is due to deposited radionuclides on the fuel surface becoming disturbed as a consequence of the changing flow rate, reactor pressure, reactor water temperature and the change from reducing to oxidising conditions; The concentration of fission product radionuclides in the reactor water is directly related to neutron flux. The concentration of these products in the reactor water, and hence in off-gas discharges, during power operation bounds the concentration during start-up and outage; Where shutdown is concerned, an additional release of some FP radionuclides is observed (i.e. halogen, alkali metal and noble gas FP). With respect to fuel pin failure and a pinhole leak, this additional release is principally due to the reduction in power output and in turn the reduction of reactor coolant pressure. Where a fuel pin has been found to fail and reactor pressure is reduced, a pressure differential exists between the inside and outside of the fuel pin. As such, this gives rise to a large increase in FP activity release into the coolant and a FP spike is observed; Where fuel pin failure does not occur, a rise in activity at shutdown is also observed for halogen, alkali metal and noble gas FP radionuclides (smaller in magnitude compared to a fuel pin failure case). Therefore, regardless of fuel pin failure, an increase in FP is observed during the shutdown phase for a range of FP radionuclides; The formation of activation products is directly related to the neutron flux within the reactor core. While the concentration of AP varies during the start-up and shutdown phases, their concentrations will be at their maximum during the power operation phase which corresponds to 100% power. Where the outage phase is concerned, power operation is assumed to have ceased and therefore AP cease to be created in the reactor core; Tritium concentrations are dependent upon neutron flux and therefore at the start of commercial operations where there has been no flux the tritium concentration is zero. Once operations begin, the tritium concentration will rise in the reactor water. This level will be diluted (or reduced) during outage by tritium free SFP water, the extent of this dilution will be proportionate to the reactor-sfp water mix. Over the subsequent operating cycles the tritium concentration will increase as any new tritium release is not balanced by tritium decay (half-life of tritium is 12 years). An equilibrium value will be reached where the losses and dilution counteract new release. Under these conditions, tritium concentration reaches a state of equilibrium after four to five years; and, During shutdown and start-up, the reactor core output varies from 0 to 100%. The N-16 and N-17 release rates are proportional to that of the output and therefore the release will follow the power output profile. However, as the half-lifes of the isotopes are very short they will decay rapidly after reactor shutdown Discharges of certain radionuclides will be higher during operations whereas discharges of others will be higher during outage. The radioactivity sources and discharge routes during power operation, start-up and shutdown and during outage are shown in Figure 3.14 and Figure 3.15 respectively. Page 69 of 255

84 Figure 3.14 Radioactive Discharge Routes during Operation, Start-up and Shutdown 256. During operation, start-up and shutdown gaseous discharges are made via the OG, HVAC and TGS systems. Carbon-14 and noble gas are the gaseous radionuclides discharged from the OG System. Negligible amounts of other nuclides such as iodine, tritium and particulates are discharged from the OG System Radionuclides such as iodine, tritium and particulate nuclides are discharged to the environment through evaporation and dispersion via the HVAC system and the TGS system. During all reactor cycle phases, tritium, iodine and other particulate nuclides migrate into the vapour from the SFP and waste system tank (pool) within each building, and are discharged from the stack via the HVAC system. Page 70 of 255

85 Figure 3.15 Radioactive Discharge Routes to the Environment during an Outage 258. The main difference between gaseous discharges during power operations, start-up and shutdown and those at an outage is that there is no discharge via the OG system and the TGS system, but there are additional discharges via the HVAC system (namely from Reactor well, Dryer/Separator (D/S) pool, S/P and Turbine Condenser Hotwell). This is described further in Section During an outage, the turbine and condensers are isolated from the reactor by closure of the Main Steam Isolation Valves. The condensers are let up to ambient atmospheric pressure and the OG system is bypassed. Therefore no discharge is made through the OG system In addition, during outage, tritium, iodine and other particulate nuclides migrate into the vapour from the reactor well and D/S pool after the RPV is opened and from the suppression pool, and these radionuclides are discharged from the stack via the HVAC system. Tritium, iodine and other particulate nuclides migrate into the vapour released from the main condenser hotwell after the turbine casing is opened, and are discharged from the stack via the HVAC system. Page 71 of 255

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87 4 Operating Techniques 261. There is a requirement under part 3 of Natural Resources Wales (NRW) application form RSR-B3 [RD16] that an applicant must describe how the production, discharge and disposal of radioactive waste will be managed to protect the environment and optimise the protection of people. The requirement necessitates the description of the optimisation process, and the identification and justification of the techniques that are proposed as Best Available Techniques (BAT) Horizon has addressed this requirement through the preparation of the Wylfa Newydd EP- RSR BAT Case which is presented in Appendix C. The BAT Case sets out the Claims, Arguments and Evidence which demonstrate that BAT will be applied to all lifecycle stages of the Power Station between design and operation 12, that the environmental performance associated with the practice of generating electricity from the Power Station will be optimised, and that impacts from potentially harmful ionising radiation on members of the public and the environment are being minimised. Six claims have been developed, namely: Claim 1: Eliminate or reduce the generation of radioactive waste; Claim 2: Minimise the radioactivity in radioactive waste disposed to the environment; Claim 3: Minimise the volume of radioactive waste disposed of to other premises; Claim 4: Selecting the optimal disposal routes for wastes transferred to other premises; Claim 5: Minimise the impacts on the environment and members of the public from radioactive waste that is disposed of to the environment; and, Claim 6: Horizon shall apply BAT when characterising and quantifying gaseous and aqueous radioactive waste discharges It is highlighted that the demonstration of BAT is an ongoing activity reflecting technological advances, economic and social factors, and changes in scientific understanding. It is therefore anticipated that the BAT Case will be periodically updated as the Project and Horizon develops. 12 A separate Wylfa Newydd Decommissioning BAT Case will be prepared at the optimum time prior to decommissioning. Decommissioning is therefore not addressed in the current BAT Case. Page 73 of 255

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89 5 Disposal of Radioactive Waste 264. This section presents information on the gaseous, aqueous and solid waste arisings associated with the normal operation of the Power Station. The gaseous and aqueous radioactive wastes generated by the UK ABWR, including those arising from operational fluctuations, trends and events that are expected to occur over the lifetime of the facility, are estimated largely from the Source Term data developed for Generic Design Assessment (GDA). A general description of the source terms is therefore provided as an introduction to the section. Thereafter a detailed description is given of the methodology, input data and assumptions used to estimate the discharges to the environment. The approach to setting discharge limits and quarterly notification levels (QNLs) is also explained, where the latter are levels at which Natural Resources Wales (NRW) would be notified of significant short term changes in plant performance Estimates of annual gaseous and aqueous radioactive discharges are developed and limits are ascribed to each of the radionuclides discharged. It is these discharge limits which are used as input to the dose assessment work set out in Section 7. The significant radionuclides or groups of radionuclides in the discharges are identified, in accordance with NRW guidance, and corresponding QNLs are proposed. Finally, estimates of annual solid waste arisings are presented for low level waste (LLW), for each proposed disposal route, intermediate level waste (ILW), high level waste (HLW) and spent fuel The information presented in this section meets the requirements of parts 4a and 4b of NRW s application form RSR-B3 [RD16]. 5.1 Source Term Data Source Terms 267. Radionuclide activity associated with all aspects of plant operation and lifecycle, including radiation protection, radioactive waste, decommissioning, and discharge assessments, were derived from the Primary Source Term (PST) the concentration of each radionuclide in the reactor water and steam and the Process Source Term (PrST) the radioactive concentrations in the ancillary systems. The PST and PrST are based on modelling and operating experience feedback (OPEX) data from modern generating plants which have comparable/similar design and operational philosophy. For the UK ABWR, relevant data has been gathered from a number of operating modern BWRs as well as the ABWRs operating in Japan Source terms were developed for Best Estimate (BE) and Design Basis (DB) conditions: The BE source term value provides a realistic assessment of the activity relating to a radionuclide and corresponds to a realistic level of radioactivity likely to be observed during the different phases of normal operation of the UK ABWR (Power Operation, Start-up, Shutdown, Outage etc). The BE value is used to inform design decisions such as environmental discharges and radioactive waste disposability assessments; and, The DB source term value provides a conservative assessment of the activity relating to a radionuclide and corresponds to an upper level of radioactivity likely to be observed during the different phases of normal operation of the UK ABWR, as well as during Expected Events. The DB value is used to inform design decisions in the UK ABWR Page 75 of 255

90 that have safety-related functions such as shielding design in the technical area of radiation protection. Both the DB and BE types of source term can represent the activity in the system at any given point and at any given phase of the operating cycle The source term also takes into account the increase in activity brought about by an Expected Event, i.e. a fuel pin failure (see Section ). The key parameter of interest where a fuel pin failure occurs is the f-value, which is defined as the release rate of noble gases, and thus the release rate associated with xenon and krypton. The BE Source Term includes a discrete fuel pin failure Expected Event for calculating optimised discharge amounts. This is described further below Radionuclides Considered 270. The full list of radionuclides expected to be produced in the UK ABWR s PST was generated through use of the Oak Ridge Isotope Generator (ORIGEN) Code, giving a raw output of over 600 radionuclides. This list was then down selected to about 100 nuclides by selecting radionuclides that contribute to 90% of the total dose or activity Two separate radionuclide lists were then generated for gaseous and aqueous discharges. These lists were determined by assessing the 100 nuclides against the following: Considering the practical requirements of the discharge assessment, i.e. which radionuclides are important as significant contributors to dose and activity; and, Incorporating a review of regulatory guidance on which radionuclides should be considered for assessment. The radionuclides considered for the UK ABWR s gaseous release, on the back of the above assessment, are listed in Table 5.1. Subsequently additional radionuclides were identified which could arise from miscellaneous activities at the site waste facilities. These are listed in Table 5.2. The radionuclides considered for the UK ABWR s aqueous release are given in Table 5.3. Table 5.1 List of Radionuclides Considered for the UK ABWR s Gaseous Release Ar-41 Xe-133m Fe-59 Sb-122 Ce-144 I-131 Kr-85 Xe-135 Co-60 Sb-124 Pu-238 I-132 Kr-85m Xe-135m Zn-65 Sb-125 Pu-239 I-133 Kr-87 Xe-137 Sr-89 Cs-134 Pu-240 I-135 Kr-88 Xe-138 Sr-90 Cs-137 Am-241 H-3 Kr-89 Cr-51 Zr-95 Ba-140 Cm-242 C-14 Xe-131m Mn-54 Nb-95 La-140 Cm-243 Xe-133 Co-58 Ag-110m Ce-141 Cm-244 Page 76 of 255

91 Table 5.2 Additional Radionuclides Potentially arising from the Site Waste Facilities Cl-36 Fe-55 Ru-103 Te-123m Ni-59 Nb-94 Ru-106 I-129 Ni-63 Tc-99 Ag-110m Pu-241 Table 5.3 List of Radionuclides Considered for the UK ABWR s Aqueous Release Cr-51 Ni-63 Ru-103 Sb-125 Ce-141 Cm-242 Mn-54 Zn-65 Ru-106 I-131 Ce-144 Cm-243 Fe-55 Sr-89 Ag-110m Cs-134 Pu-238 Cm-244 Fe-59 Sr-90 Sb-122 Cs-137 Pu-239 H-3 Co-58 Zr-95 Te-123m Ba-140 Pu-240 Co-60 Nb-95 Sb-124 La-140 Am Expected Event Identification of Representative Event 272. In preparing discharge estimates it is necessary to take into consideration discharges arising from operational fluctuations, trends and events that are expected to occur over the likely lifetime of the facility [RD21]. Expected Events are those that are anticipated to occur at least once during the lifetime of the plant (frequency >10-2 ). They do not include accidents or activities that are inconsistent with the use of BAT The relationship between discharges during normal operations and those during Expected Events is illustrated in Figure 5.1. The effect on the plant of Expected Events should be minimal, amounting to no more than a small deviation from normal operating conditions, and consequences would be below the basic safety objective (<0.01 msv) without any mitigating actions A number of Expected Events have been identified for the UK ABWR, including: Small gaseous leak from each equipment item; Small gaseous leak from pipe, valve or pump in each system; Small liquid leak from each equipment item; Small liquid leak from pipe, valve or pump in each system; Degradation of equipment which guarantees Decontamination Factor (DF); Fuel cladding damage (fuel pin failure); and, Control rod damage (boron carbide tube cracking). Page 77 of 255

92 Figure 5.1 Discharges during Normal Operations and Expected Events 275. The events were ranked according to their effect on environmental discharges, taking into consideration the magnitude of the release activity. Ranking was made according to the following criteria: Ranking A: It is necessary to estimate the release nuclides, because there is extra release to the environment; Ranking B: It is not necessary to estimate, because there is only small release; and, Ranking C: It is not necessary to estimate, because there is no release. Fuel pin failure was the only Expected Event to be classified as Ranking A. It was therefore chosen as the representative event Fuel Pin Failure Frequency of Failure 276. Operating data provided by a fuel manufacturer, Global Nuclear Fuels Limited (GNF), shows that fuel reliability has improved over the last three decades, reducing from approximately 50 fuel pin failures per million rods for the period 1980 to 1984, to 3.4 failures per million rods for the period 2010 to 2014 (Figure 5.2). The figure also identifies the reasons for the improved fuel reliability The core of a UK ABWR comprises 872 assemblies each of which contains 92 fuel pins. This gives a total of 80,224. The reactor operates on an 18 month cycle (17 months operation followed by a month s outage) and at each outage 26% of the pins are scheduled for replacement. Over the 60 year life of the facility, therefore, 883,200 fuel pins would be replaced. At the above failure rate of 3.4 per million it is anticipated that three pins would have failed, giving a fuel pin failure rate for the UK ABWR of one every 20 years. Page 78 of 255

93 Figure 5.2 GNF Historical Fuel Reliability Performance 5.2 Quantification of Gaseous and Aqueous Discharges Gaseous Discharges 278. A best estimate of the total 12 month rolling gaseous discharges from the Power Station was developed from four component estimates, namely the estimate of gaseous discharges from: A single UK ABWR for each radionuclide / radionuclide group over a rolling 12 month period (D Rn,gas,GDA); A single UK ABWR for a single Expected Event for each radionuclide / radionuclide group (D Rn,gas,EE); The Rw/B for each radionuclide / radionuclide group over a 12 month period (D Rn,gas,Rw/B) 13 ; and, The site-specific waste facilities for each radionuclide / radionuclide group over a 12 month period (D Rn,gas,SSwaste) The Best Estimate, D Rn,gas,total, was then calculated as follows: D Rn,gas,total = (D Rn,gas,GDA x 2) + D Rn,gas,EE + D Rn,gas,Rw/B + D Rn,gas,SSwaste Eq At GDA it was intended that each UK ABWR would receive the gaseous discharge from its dedicated Rw/B (ie. resulting in two Rw/Bs at the Power Station). Each Rw/B discharge formed part of the corresponding HVAC system discharge (the other HVAC components arising from the R/B and T/B). Following an optimisation process, the Power Station is now provided with a single Rw/B) which will serve both UK ABWRs. The single Rw/B gaseous discharge (discharges expected to be double from those stated in Table 5.9) is now identified as a separate stream which is released to the environment via Unit 1 R/B stack. Page 79 of 255

94 Effectively the best estimate is based upon the operation of two UK ABWRs, the occurrence of a single Expected Event over any 12 month period, and the operation of the Rw/B and the site waste facilities Since the two UK ABWRs at the Power Station site will be independent, it has been concluded that the discharges and disposals calculated for one UK ABWR (presented in the GDA) will be applicable for each one of the UK ABWRs at the Power Station. This assumes they will be operated in the same way as has been assumed in the GDA UK ABWR source term data assessments. It is not anticipated that there will be any opportunity to reduce the combined discharges from two units The basis for the four component estimates are described below. The first three estimates D Rn,gas,GDA, D Rn,gas,EE and D Rn,gas,Rw/B were developed from the GDA UK ABWR source term described above. The GDA UK ABWR source term assessment does not include discharges from the waste facilities (Lower Activity Waste Management Facility (LAWMF), ILW Storage Facility and Spent Fuel Storage Facility (SFSF)) which is why the discharges from these facilities were quantified separately (see Section ) Discharge from a Single UK ABWR over a Rolling 12 Month Period 282. The estimate of gaseous discharges to the environment from a single UK ABWR (for each of the radionuclides listed in Table 5.1) over a 12 month period, D Rn,gas,GDA, was based upon estimates of monthly discharges to the R/B stack from the Off-Gas System (OG), HVAC System and Turbine Gland System (TGS). D Rn,gas,GDA = (D Rn,gas,OG + D Rn,gas,HVAC + D Rn,gas,TGS) x 12 Eq. 5.2 where D Rn,gas,OG, D Rn,gas,HVAC and D Rn,gas,TG are the monthly average gaseous discharges from the OG, HVAC System and TGS No measurement data is readily available for these discharges. Therefore the monthly predicted gaseous discharges presented below are based on theoretical calculation of the production of radioisotopes in water, subsequent partitioning and transfer to steam and the performance of the process systems in the relevant discharge streams. As a result, the figures showing the predicted monthly discharge are estimates and somewhat idealised. Off-Gas System (D Rn,gas,OG) 284. The radioactive discharges from the OG arise as volatile non-condensable gases such as noble gases and C-14. They are removed from the reactor water circuit at the Steam Jet Air Ejector (SJAE) and are routed to the OG system, where they are held up, and then discharged from the stack. During outage, the turbine and condensers are isolated from the reactor by closure of the Main Steam Isolation Valves, and the condensers are let up to ambient atmospheric pressure. Therefore, during this time, no discharge is made through the OG system. Page 80 of 255

95 285. The monthly discharges of radionuclides from the OG system have been calculated based on the transport of radioactivity through the reactor, main steam line and the OG system, and are summarised in Table 5.4. The calculations take appropriate account of flow rates, partitioning between water and steam, separation into the OG system, removal in the OG system condenser and charcoal adsorbers. As the OG system is extremely effective in removing tritiated water, iodine and particulates from the gas stream, only negligibly small amounts of the associated radionuclides (i.e. H-3, iodine and radionuclides associated with particulate matter) are present in the discharged gas. For the purposes of this assessment the discharges of these radionuclides are considered to be zero. Table 5.4 Monthly Gaseous Discharges from the OG System (D Rn,gas,OG) Radionuclides Monthly Discharge (D Rn,gas,OG) (Bq/month) Ar E+11 C E+10 Kr-85 Kr-85m Kr-87 Kr-88 Kr-89 Xe-131m Xe-133 Xe-133m Xe-135 Xe-135m Xe-137 Xe E E E E E E E E E E E E+00 HVAC System (D Rn,gas,HVAC) 286. The radioactive discharge from the HVAC system (Units 1 and 2) is aggregated from contributing plant systems serving the R/B and T/B. During start-up, power operation and shutdown, H-3, iodine and particulate nuclides migrate from the water in each plant system into the air. During outage they also migrate into the vapour from: The Reactor Well and Dryer/Separator (D/S) pool after the Reactor Pressure Vessel (RPV) is opened for fuel change; The Suppression Pool (S/P) as the air in the Suppression Chamber (S/C) is ventilated to allow workers to enter the enclosure; and, The main Condenser Hotwell after the turbine casing is opened for maintenance. Page 81 of 255

96 287. In the HVAC systems present in Units 1 and 2, particulates are removed from the air by using High Efficiency Particulate Air (HEPA) filters, and the filtered air is then discharged from the HVAC system via the stack. Volatile non-condensable gases are removed from the reactor water/steam circuit at the SJAE and are discharged via the OG system as described above. As a result, noble gases and C-14 are typically not present in the systems that are served by the HVAC system and so very low concentrations would be expected in the HVAC system discharge. For the purposes of this assessment the discharge of noble gases and C-14 from the HVAC system is assumed to be zero The monthly discharges of radionuclides from the HVAC system have been calculated based on the radioactivity present in the Spent Fuel Pool (SFP), reactor well, D/S pool, S/P, condenser hotwell and various tanks containing reactor water, supernatant or washings from liquid waste systems (Table 5.5). The calculations take account of the transition rates of radioactivity through these systems, including partitioning between water and air, water temperature, evaporation rates, number of contributing vessels and removal by abatement systems The monthly discharge given in Table 5.5, D Rn,gas,HVAC, is based upon the 18 month operating cycle, and is the average of 17 months of operation [streams designated (1)(2)(3) and (1)(2)(3)(4)] and one month of outage [streams designated (4)]. Page 82 of 255

97 Table 5.5 Monthly Gaseous Discharges from the HVAC System (D Rn,gas,HVAC) Rn Monthly Discharge (Bq/month) R/B T/B SFP SFP Reactor Well D/S pool Supp n Pool CUW Filter B kwash Receiver Tank (from CUW) CUW Filter B kwash Receiver Tank (from FPC) Turbine Condenser Hotwell CF B kwash Receiver Tank Average (DRn,gas,HVAC) Operating Mode (1)(2)(3) (4) (4) (4) (4) (1)(2)(3)(4) (1)(2)(3)(4) (4) (1)(2)(3)(4) Ag-110m 5.6E E E E E E E E E E-02 Am E E E E E E E E E E-07 Ba E E E E E E E E E E+01 Ce E E E E E E E E E E+01 Ce E E E E E E E E E E+01 Cm E E E E E E E E E E-04 Cm E E E E E E E E E E-08 Cm E E E E E E E E E E-06 Co E E E E E E E E E E+02 Co E E E E E E E E E E+02 Cr E E E E E E E E E E+02 Cs E E E E E E E E E E+00 Cs E E E E E E E E E E+00 Fe E E E E E E E E E E+01 H-3 6.3E E E E E E E E E E+10 I E E E E E E E E E E+06 I E E E E E E E E E E+06 Page 83 of 255

98 Rn Monthly Discharge (Bq/month) R/B T/B SFP SFP Reactor Well D/S pool Supp n Pool CUW Filter B kwash Receiver Tank (from CUW) CUW Filter B kwash Receiver Tank (from FPC) Turbine Condenser Hotwell CF B kwash Receiver Tank Average (DRn,gas,HVAC) Operating Mode (1)(2)(3) (4) (4) (4) (4) (1)(2)(3)(4) (1)(2)(3)(4) (4) (1)(2)(3)(4) I E E E E E E E E E E+06 I E E E E E E E E E E+05 La E E E E E E E E E E+01 Mn E E E E E E E E E E+02 Nb E E E E E E E E E E+01 Pu E E E E E E E E E E-06 Pu E E E E E E E E E E-07 Pu E E E E E E E E E E-06 Sb E E E E E E E E E E-01 Sb E E E E E E E E E E+01 Sb E E E E E E E E E E+00 Sr E E E E E E E E E E+01 Sr E E E E E E E E E E+00 Zn E E E E E E E E E E+02 Zr E E E E E E E E E E+01 Note. Operating mode: (1) Power operation, (2) Start-up, (3) Shutdown, (4) Outage. Page 84 of 255

99 Turbine Gland System (D Rn,gas,TG) 290. Steam is used to seal the turbine shaft seal parts and the major valve gland parts during reactor operation. The sealing steam is produced by the Gland Steam Evaporator from water taken from the Condensate Storage Tank (CST), and includes H-3, iodine and particulate nuclides present in the water After usage, the majority of the steam is condensed by the Gland Steam Condenser and the particulates are removed by a HEPA filter. The remaining gases are discharged from the TGS via the R/B stack. During outage, the TGS system is not operated and there is thus no discharge through the TGS during this period Volatile non-condensable gases are removed from the reactor water/steam circuit at the SJAE and are discharged via the OG as described earlier. As a result, noble gases and C-14 are typically not present in the CST and so very low concentrations would be expected in the TGS discharge. For the purposes of this assessment the discharge of noble gases and C-14 from the TGS is assumed to be zero The monthly discharges of radionuclides from the TGS have been calculated based on the transport of radioactivity through the CST and the TGS (Table 5.6). The calculations take account of the flow rate of radioactivity through these systems, including partitioning between water and air, water temperature, evaporation rates and removal by abatement systems. Table 5.6 Monthly Gaseous Discharges from the TGS (D Rn,gas,TGS) Radionuclides Monthly Discharges Radionuclides Monthly Discharges (D Rn,gas,TGS) (D Rn,gas,TGS) (Bq/month) (Bq/month) Ag-110m 6.1E-01 I E+06 Am E-05 I E+05 Ba E+02 I E+06 Ce E+02 La E+02 Ce E+02 Mn E+03 Cm E-03 Nb E+03 Cm E-07 Pu E-04 Cm E-04 Pu E-05 Co E+03 Pu E-05 Co E+03 Sb E+00 Cr E+03 Sb E+02 Cs E+02 Sb E+02 Cs E+02 Sr E+02 Fe E+02 Sr E+01 H-3 1.3E+11 Zn E+02 I E+05 Zr E+03 Page 85 of 255

100 Total Discharge from a Single UK ABWR over a Rolling 12 Month Period (D Rn,gas,GDA) 294. The total gaseous discharge to the environment from a single UK ABWR over a 12 month rolling period, D Rn,gas,GDA, was calculated in accordance with Equation 5.2. The results are presented in Table 5.7. Table 5.7 Annual Gaseous Discharge from a Single UK ABWR (D Rn,gas,GDA) Radionuclides Total Discharge (D Rn,gas,GDA) Radionuclides Total Discharge (D Rn,gas,GDA) (Bq/annum) (Bq/annum) Ag-110m 8.2E+00 Kr E+03 Am E-04 Kr E+08 Ar E+12 Kr E+00 Ba E+03 La E+03 C E+11 Mn E+04 Ce E+04 Nb E+04 Ce E+04 Pu E-03 Cm E-01 Pu E-04 Cm E-05 Pu E-04 Cm E-03 Sb E+02 Co E+04 Sb E+04 Co E+04 Sb E+03 Cr E+04 Sr E+03 Cs E+03 Sr E+02 Cs E+03 Xe-131m 1.4E+08 Fe E+03 Xe E+10 H-3 2.5E+12 Xe-133m 1.7E+06 I E+08 Xe E-11 I E+07 Xe-135m 0.0E+00 I E+07 Xe E+00 I E+07 Xe E+00 Kr E+08 Zn E+03 Kr-85m 2.3E+09 Zr E Discharge from a Single UK ABWR for a Single Expected Event 295. Hitachi-GE has undertaken a study to identify and assess the unplanned events that could have a bearing on discharges and that are considered likely to occur during normal operation of a UK ABWR across its planned plant lifetime. As described earlier, the study identified one such expected event, specifically a fuel pin failure Should a fuel pin failure occur, elevated radioactivity, as a result of the escape of fission products from the fuel pin, is detected at the inlet of the OG system and a process is Page 86 of 255

101 initiated to identify and isolate the fuel assembly in which the pin failure has occurred. Once the relevant fuel assembly is isolated, the fission within the fuel and subsequent leakage into the coolant reduces. This identification and isolation process can last up to 14 days Failure of a fuel pin increases the noble gas release ratio ( f-value ). Should a fuel pin failure occur, OPEX shows that a noble gas release ratio of 1.0E+08 Bq/s the Limiting Condition of Operation (LCO) value 14 is appropriate to calculate the associated discharges. The gaseous discharge due to the expected event was quantified by taking account of the escape of volatile fission products into the reactor water and their transport around the steam circuit, separation from the steam at the SJAE and subsequent retention in the OG system. The radionuclides and their associated activities released in the discharged gas per Expected Event, D Rn,gas,EE, is shown in Table 5.8. Table 5.8 Gaseous Discharges for a Fuel Pin Failure Expected Event (D Rn,gas,EE) Radionuclides Kr-85 Kr-85m Kr-87 Kr-88 Kr-89 Xe-131m Xe-133 Xe-133m Xe-135 Xe-135m Xe-137 Xe-138 Activity Discharge (D Rn,gas,EE) (Bq per event) 1.1E E E E E E E E E E E E Discharge from the Rw/B over a 12 Month Period 298. The radioactive discharge from the Rw/B arises from the vents/extracts from the various Rw/B tanks. The monthly discharges by tank and resulting total annual discharge are set out in Table LCOs define the safe operating envelope within which Horizon will operate the Power Station. Page 87 of 255

102 Table 5.9 Gaseous Discharges from the Rw/B (D Rn,gas,Rw/B) Rn Monthly Discharge (Bq/month) Annual Discharge LCW HCW Concent. (Bq/annum) Collection Collection Waste (DRn,gas,Rw/B) Tank Tank Tank Powder Resin Storage Tank Filter Crud Storage Tank Operating Mode Bead Resin Storage Tank (1)(2)(3)(4) (1)(2)(3)(4) (1)(2)(3)(4) (1)(2)(3)(4) (1)(2)(3)(4) (1)(2)(3)(4) Ag-110m 1.4E E E E E E E+00 Am E E E E E E E-05 Ba E E E E E E E+01 Ce E E E E E E E+02 Ce E E E E E E E+02 Cm E E E E E E E-02 Cm E E E E E E E-06 Cm E E E E E E E-04 Co E E E E E E E+03 Co E E E E E E E+04 Cr E E E E E E E+03 Cs E E E E E E E+02 Cs E E E E E E E+02 Fe E E E E E E+03 H-3 1.9E E E E E E E+11 I E E E E E E E+08 I E E E E E E E+06 I E E E E E E E+07 I E E E E E E E+06 La E E E E E E E+02 Mn E E E E E E E+03 Nb E E E E E E E+02 Pu E E E E E E E-04 Pu E E E E E E E-05 Pu E E E E E E E-04 Sb E E E E E E E+00 Sb E E E E E E E+02 Sb E E E E E E E+01 Sr E E E E E E E+02 Sr E E E E E E E+01 Zn E E E E E E E+03 Zr E E E E E E E+02 Note. Operating mode: (1) Power operation, (2) Start-up, (3) Shutdown, (4) Outage. Page 88 of 255

103 Discharge from the Site-specific Waste Facilities 299. The estimate of gaseous discharges from the site-specific waste facilities for each radionuclide / radionuclide group over a 12 month rolling period, D Rn,gas,SSwaste, was based upon the activity of the HEPA filters (from the R/B, T/B, Rw/B and S/B) processed in the LAWMF, the number processed per year, and an assumption that 0.1% of the activity on each filter is discharged to atmosphere. The annual site-specific discharges by filter type and as a total, D Rn,gas,SSwaste, are presented in Table Table 5.10 Annual Gaseous Discharges for the Site Waste Facilities (D Rn,gas,SSwaste) Radionuclides Site-specific Discharge (Bq/annum) R/B T/B Rw/B S/B Total (D Rn,gas,SSwaste) Ag-110m 1.5E E E E E-02 Am E E E E E-07 Ar E E E E E+00 Ba E E E E E-01 C E E E E E-02 Ce E E E E E+00 Ce E E E E E+01 Cl E E E E E-06 Cm E E E E E-05 Cm E E E E E-08 Cm E E E E E-06 Co E E E E E+01 Co E E E E E+02 Cr E E E E E+01 Cs E E E E E+00 Cs E E E E E+00 Fe E E E E E+02 Fe E E E E E+00 H-3 0.0E E E E E+00 I E E E E E-05 I E E E E E+00 I E E E E E-02 I E E E E E-01 I E E E E E-01 Kr E E E E E+00 Kr-85m 0.0E E E E E+00 Kr E E E E E+00 Kr E E E E E+00 Page 89 of 255

104 Radionuclides Site-specific Discharge (Bq/annum) R/B T/B Rw/B S/B Total (D Rn,gas,SSwaste) Kr E E E E E+00 La E E E E E-01 Mn E E E E E+01 Nb E E E E E-04 Nb E E E E E+00 Ni E E E E E-02 Ni E E E E E+00 Pu E E E E E-06 Pu E E E E E-07 Pu E E E E E-07 Pu E E E E E-04 Ru E E E E E-01 Ru E E E E E-01 Sb E E E E E-04 Sb E E E E E+00 Sb E E E E E+00 Sr E E E E E+00 Sr E E E E E+00 Tc E E E E E-01 Te-123m 2.3E E E E E-03 Xe-131m 0.0E E E E E+00 Xe E E E E E+00 Xe-133m 0.0E E E E E+00 Xe E E E E E+00 Xe-135m 0.0E E E E E+00 Xe E E E E E+00 Xe E E E E E+00 Zn E E E E E+01 Zr E E E E E Best Estimate of the Total 12 Month Rolling Gaseous Discharges 300. The best estimate of the total 12 month rolling gaseous discharges from the Power Station, D Rn,gas,total, developed from the above component estimates according to Equation 5.1, is presented in Table It is reiterated that the estimate is based upon the operation of two UK ABWRs, the Rw/B and the site waste facilities, and the occurrence of a single fuel pin failure over the 12 month period. Page 90 of 255

105 Table 5.11 Best Estimate Total 12 Month Rolling Gaseous Discharges (D Rn,gas,total) Radionuclides Total Discharge (D Rn,gas,total) Radionuclides Total Discharge (D Rn,gas,total) (Bq/annum) (Bq/annum) Ag-110m 1.9E+01 La E+04 Am E-04 Mn E+04 Ar E+12 Nb E-04 Ba E+04 Nb E+04 C E+12 Ni E-02 Ce E+04 Ni E+00 Ce E+04 Pu E-03 Cl E-06 Pu E-04 Cm E-01 Pu E-04 Cm E-05 Pu E-04 Cm E-03 Ru E-01 Co E+04 Ru E-01 Co E+04 Sb E+02 Cr E+04 Sb E+04 Cs E+03 Sb E+03 Cs E+03 Sr E+04 Fe E+02 Sr E+03 Fe E+04 Tc E-01 H-3 5.5E+12 Te-123m 2.4E-03 I E-05 Xe-131m 2.9E+09 I E+08 Xe E+11 I E+08 Xe-133m 1.7E+07 I E+07 Xe E-11 I E+07 Xe-135m 0.0E+00 Kr E+09 Xe E+00 Kr-85m 1.0E+10 Xe E+00 Kr E+03 Zn E+04 Kr E+08 Zr E+04 Kr E+00 Page 91 of 255

106 5.2.2 Aqueous Discharges 301. The liquid effulent treatment system which contributes to aqueous radioactive discharges is the HCW system 15. This system is designed to treat water with high levels of soluble and insoluble impurities by passing it through an ion exchange bed and the evaporation process (more detail on the system is presented in Section 3). After confirmation that residual activity levels in the HCW Sample Tank are sufficiently low, the treated water is usually transferred to the CST for reuse within the plant. Whenever the water balance in the Power Station prevents the reuse of the water (i.e. the capacity of the CST is reached), the treated effluent is discharged to the environment The UK ABWR design assessed at GDA had a laundry and associated drainage system the Laundry Drain (LD) system which received liquid effluent from the laundry, emergency showers, and handwashing facilities in controlled areas. The laundry and LD system has since been removed from the design of the Power Station, and liquid effluent arising from emergency showers and handwashing facilities is now routed to the HCW system A best estimate of the total 12 month rolling aqueous discharges from the Power Station, D Rn,liquid,total, was calculated as follows: D Rn,liquid,total = D Rn,liquid,HCW* x 2 Eq. 5.3 where: D Rn,liquid,HCW* is an estimate of the aqueous discharge from the HCW sample tanks of a single UK ABWR for each radionuclide / radionuclide group over a rolling 12 month period; 304. It is noted that the liquid discharges to the HCW sample tanks will vary depending on the operating mode of the contributing systems. However, it is considered that the activity concentration in these tanks will not vary significantly over an operating cycle due to both the capability to recirculate the contents through the treatment systems (to ensure the discharge limits are met) and also the volume of the tanks themselves which will provide a buffering effect to any minor variations in the concentration of the feed water Best Estimate of the Total 12 Month Rolling Aqueous Discharges 305. An estimates of the annual discharges from the HCW sample tanks is given in Table The estimate has taken account of the movement of radioactivity through the system including flow rates and effectiveness of abatement. However, it was assumed that all of the contaminated aqueous wastes processed in the HCW system are discharged to the environment. Therefore, the annual discharge presented in Table 5.12 is the maximum conceivable inventory that could be discharged from the HCW system into the 15 Discharges are also made from the controlled area drain (CAD) system, which is used to collect waste water from other plant and systems in the Radiologically Controlled Areas (RCAs) in the R/B and T/B. However, it is only during an accident that the waste water collected from the RCAs is contaminated, and for this reason this discharge route is not included in this discharge assessment. Page 92 of 255

107 environment. The best estimate of 12 month rolling aqueous discharges from the Power Station, D Rn,liquid,total, calculated according to Equation 5.3, is given in Table Table 5.12 Annual Aqueous Discharges from the HCW Sample Tanks Radionuclides Discharge from HCW Sample Tank Radionuclides Discharge from HCW Sample Tank (D Rn,liquid,HCW*) (Bq/annum) (D Rn,liquid,HCW*) (Bq/annum) Ag-110m 1.2E+00 La E+03 Am E-02 Mn E+04 Ba E+03 Nb E+04 Ce E+03 Ni E+04 Ce E+04 Pu E-01 Cm E-01 Pu E-02 Cm E-04 Pu E-02 Cm E-02 Ru E+03 Co E+04 Ru E+03 Co E+05 Sb E+01 Cr E+03 Sb E+03 Cs E+03 Sb E+03 Cs E+03 Sr E+03 Fe E+05 Sr E+03 Fe E+03 Te-123m 6.4E+00 H-3 2.0E+11 Zn E+04 I E+04 Zr E+04 Table 5.13 Best Estimate Total 12 Month Rolling Aqueous Discharges (D Rn,liquid,total) Radionuclides Total Discharge (D Rn,liquid,total) Radionuclides Total Discharge (D Rn,liquid,total) (Bq/annum) (Bq/annum) Ag-110m 2.3E+00 La E+03 Am E-02 Mn E+04 Ba E+03 Nb E+04 Ce E+04 Ni E+05 Ce E+04 Pu E-01 Cm E-01 Pu E-01 Cm E-03 Pu E-01 Cm E-02 Ru E+03 Co E+04 Ru E+03 Co E+05 Sb E+01 Page 93 of 255

108 Radionuclides Total Discharge (D Rn,liquid,total) Radionuclides Total Discharge (D Rn,liquid,total) (Bq/annum) (Bq/annum) Cr E+04 Sb E+04 Cs E+03 Sb E+04 Cs E+03 Sr E+03 Fe E+06 Sr E+03 Fe E+03 Te-123m 1.3E+01 H-3 4.0E+11 Zn E+04 I E+04 Zr E Setting Discharge Limits and Quarterly Notification Levels Discharge Limits 306. Annual limits are proposed for gaseous and aqueous discharges from the Power Station based on the use of BAT together with the application of a minimum headroom factor necessary to permit normal operation of a facility. The limits are applied on a 12 month rolling basis Discharge limits for gaseous discharges from the Power Station, L Rn,gas,PS, were calculated as follows: L Rn,gas,PS = (D Rn,gas,GDA x HF x 2) + (D Rn,gas,EE x 2) + (D Rn,gas,Rw/B x HF) + (D Rn,gas,SSwaste x HF) Eq. 5.4 where HF is the headroom factor for each radionuclide/radionuclide group. The limit is based upon the operation of two UK ABWRs, the Rw/B and site waste facilities (with the headroom factors applied), and the occurrence of two Expected Events over the 12 month period Discharge limits for the two R/B stacks used as input to the dose assessment presented in Section 7 were calculated on the basis that Unit 1 R/B stack receives 100% of the ventilation from the site-specific waste facilities (along with the ventilation from the Rw/B). The calculation was as follows: L Rn,gas,Stack 1 = (D Rn,gas,GDA x HF) + (D Rn,gas,EE) + (D Rn,gas,Rw/B x HF) + (D Rn,gas,SSwaste x HF) Eq. 5.5 L Rn,gas,Stack 2 = (D Rn,gas,GDA x HF) + (D Rn,gas,EE) Eq. 5.6 Page 94 of 255

109 309. Discharge limits for aqueous discharges from the Power Station, L Rn,liquid,PS, were calculated as follows: L Rn,liquid,PS = D Rn,liquid,HCW* x 2 x HF Eq Headroom Factors 310. The headroom factors used in the determination of discharge limits for the Power Station are given in Table 5.14 and Table The headroom factors were derived as part of the GDA for the UK ABWR [RD20]. The use of the factors reflects NRW s guidance for limit setting [RD21] which suggests that a headroom should be allowed between the actual levels of discharge expected during normal operation and the limits themselves. The headroom factor should be set at the minimum practical level but may take into account that there may be uncertainty about the data presented in this EP-RSR Application due to the technology in operation at the Power Station. Table 5.14 Headroom Factors by Radionuclide Group Radionuclide Group Headroom Factor Iodine 1.7 C Noble gases 2.1 Ar H Particulates 4.1 Table 5.15 Headroom Factors by Radionuclide Radionuclide Group Headroom Factor Radionuclide Group Headroom Factor Ag-110m Particulates 4.1 La-140 Particulates 4.1 Am-241 Particulates 4.1 Mn-54 Particulates 4.1 Ar-41 Ar Nb-94 Particulates 4.1 Ba-140 Particulates 4.1 Nb-95 Particulates 4.1 C-14 C Ni-59 Particulates 4.1 Ce-141 Particulates 4.1 Ni-63 Particulates 4.1 Ce-144 Particulates 4.1 Pu-238 Particulates 4.1 Cl-36 Iodine 1.7 Pu-239 Particulates 4.1 Cm-242 Particulates 4.1 Pu-240 Particulates 4.1 Cm-243 Particulates 4.1 Pu-241 Particulates 4.1 Cm-244 Particulates 4.1 Ru-103 Particulates 4.1 Co-58 Particulates 4.1 Ru-106 Particulates 4.1 Page 95 of 255

110 Radionuclide Group Headroom Factor Radionuclide Group Headroom Factor Co-60 Particulates 4.1 Sb-122 Particulates 4.1 Cr-51 Particulates 4.1 Sb-124 Particulates 4.1 Cs-134 Particulates 4.1 Sb-125 Particulates 4.1 Cs-137 Particulates 4.1 Sr-89 Particulates 4.1 Fe-55 Particulates 4.1 Sr-90 Particulates 4.1 Fe-59 Particulates 4.1 Tc-99 Particulates 4.1 H-3 Tritium 3.8 Te-123m Particulates 4.1 I-129 Iodine 1.7 Xe-131m Nobles 2.1 I-131 Iodine 1.7 Xe-133 Nobles 2.1 I-132 Iodine 1.7 Xe-133m Nobles 2.1 I-133 Iodine 1.7 Xe-135 Nobles 2.1 I-135 Iodine 1.7 Xe-135m Nobles 2.1 Kr-85 Nobles 2.1 Xe-137 Nobles 2.1 Kr-85m Nobles 2.1 Xe-138 Nobles 2.1 Kr-87 Nobles 2.1 Zn-65 Particulates 4.1 Kr-88 Nobles 2.1 Zr-95 Particulates 4.1 Kr-89 Nobles In order to determine the headroom factors for the proposed annual limits, the mean and standard deviation for each of the radionuclide data sets presented at GDA were determined and a confidence interval, or coverage factor k, applied. The coverage factor was based on a one-sided normal distribution (as an upper limit is being determined). The headroom factor was thus determined as follows: Headroom factor = (mean value + k x standard deviation)/mean value = (µ + kσ)/µ 312. A fundamental assumption is that the discharges to the environment are directly proportional to variations in the reactor water or reactor steam values. Horizon considers that this approach, which is based upon operational experience from existing BWRs, is the most credible means of understanding the potential variation in reactor water or reactor steam values and hence the potential variation in discharges to the environment reflect the variation in average activity concentrations Horizon considers that the derivation of headroom factors using a coverage factor corresponding to the 99.9 th percentile value in the datasets would result in a low risk of breaching the proposed annual discharge limit for normal operations (excluding the Expected Event). For such a headroom factor there is a 1/1000 chance that the permitted annual 12 month rolling limit for a particular radionuclide or radionuclide group will be exceeded in a year A headroom factor has not been calculated for the Expected Event (fuel pin failure) as this is considered a discrete occurrence. Page 96 of 255

111 5.3.2 Quarterly Notification Levels 315. NRW may set notification levels on radionuclides either where it wishes to be notified of significant short term changes in plant performance and process control (as shown by the level of discharges) or where it requires the operator to review the techniques used to control discharges. These are typically quarterly notification levels (QNLs) [RD21] NRW normally sets radionuclide-specific QNLs for the site based on the: Typical levels of discharges over a quarter, excluding abnormal events; and, Level of increased discharges that indicates significant deterioration of plant performance and process control to the extent that it is necessary for the operator to review the techniques used [RD21] Horizon has proposed QNLs for each radionuclide or group of radionuclides for which discharges have been quantified. These are based on a quarter of the discharge limits but without Expected Events: QNL Rn,gas,PS = ((D Rn,gas,GDA x HF x 2) + (D Rn,gas,Rw/B x HF) + (D Rn,gas,SSwaste x HF))/4 Eq. 5.8 QNL Rn,liquid,PS = (D Rn,liquid,HCW* x HF x 2)/4 Eq It is concluded, given the data available on the discharges from the UK ABWR, that a QNL of a quarter of the discharge limits is sufficiently conservative. When the Power Station becomes operational, Horizon will be able to review the discharges to determine whether the QNLs require amending with NRW. 5.4 Discharge data for the Human Dose Assessment 319. As part of the Dose Assessment (see Section 7), the radionuclides that form the gaseous and aqueous discharges from the Power Station were modelled (see Section 7.1.3). For the Dose Assessment to be undertaken, the radionuclides present in the gaseous and aqueous discharges were assigned limits based on the rationale provided in Section The gaseous and aqueous discharges values used in the Dose Assessment are given in Table 5.16 and Table It should be noted that the values provided in Table 5.16 and Table 5.17 are not the annual limits for which Horizon is applying for the UK ABWR s significant radionuclides and radionuclide groups. These limits, along with the annual discharges and QNLs, are given in Table 5.19 (gaseous discharges) and Table 5.20 (aqueous discharges). Page 97 of 255

112 Table 5.16 Annual Gaseous Radionuclide Limits used in the Dose Assessment Radionuclide Discharge (Bq/y) Stack 1 Stack 2 Ag-110m 4.6E E+01 Am E E-04 Ar E E+12 Ba E E+04 C E E+12 Ce E E+04 Ce E E+04 Cm E E-01 Cm E E-05 Cm E E-03 Co E E+05 Co E E+05 Cr E E+05 Cs E E+03 Cs E E+03 Fe E E+04 H-3 1.1E E+12 I E E+08 I E E+08 I E E+07 I E E+07 Kr E E+09 Kr-85m 1.0E E+10 Kr E E+03 Kr E E+08 Kr E+00 La E E+04 Mn E E+04 Nb E E+05 Pu E E-03 Pu E E-03 Pu E E-03 Sb E E+02 Sb E E+04 Sb E E+04 Sr E E+04 Sr E E+03 Tc E E+00 Page 98 of 255

113 Radionuclide Discharge (Bq/y) Stack 1 Stack 2 Xe-131m 2.9E E+09 Xe E E+11 Xe-133m 1.8E E+07 Xe E E-11 Xe-135m Xe Xe Zn E E+04 Zr E E+04 Table 5.17 Annual Aqueous Discharge Limits used in the Dose Assessment Radionuclide Bq/y Radionuclide Bq/y Ag-110m 9.4E+00 I E+00 Am E-02 I E+00 Ba E+04 La E+04 C E+00 Mn E+05 Ce E+04 Nb E+05 Ce E+05 Ni E+05 Cm E+00 Pu E+00 Cm E-03 Pu E-01 Cm E-01 Pu E-01 Co E+05 Ru E+04 Co E+05 Ru E+04 Cr E+04 Sb E+02 Cs E+04 Sb E+04 Cs E+04 Sb E+04 Fe E+06 Sr E+04 Fe E+04 Sr E+03 H-3 1.5E+12 Tc E+00 I E+05 Te-123m 5.3E+01 I E+00 Zn E+05 I E+00 Zr E+04 Page 99 of 255

114 5.5 Identification of Significant Radionuclides 321. In line with the NRW guidance on limits setting [RD21], Horizon has established groups of significant radionuclides for which discharge limits will be set. The basis for the selection is shown in Table Table 5.18 Basis for Establishing Radionuclide Groups for Discharge Limit Setting Criterion Specified in NRW Guidance [RD21] Radionuclides that are significant in terms of radiological impact on people (that is, the dose to the most exposed group at the proposed limit exceeds 1 μsv per year). Radionuclides that are significant in terms of radiological impact on non-human species (this only needs to be considered where the impact on reference organisms from the discharges of all radionuclides/groups at the proposed limits exceeds 40 μgy/hour). Radionuclides that are significant in terms of the quantity of radioactivity discharged (that is, the discharge exceeds 1 TBq per year). Radionuclides that may contribute significantly to collective dose (this only needs to be considered where the 500-year collective dose from the discharges of all radionuclides at the proposed limits exceeds 1 mansievert per year). Radionuclides that are constrained under national or international agreements or are of concern internationally. Radionuclides that are indicators of plant performance, if not otherwise limited on the above criteria. For the appropriate generic categories from the RSR Pollution Inventory (e.g. alpha particulate and beta/gamma particulate for discharges to air) to limit any radionuclides not otherwise covered by the limits set on the above criteria. Applicable Radionuclide or Group of Radionuclides Gaseous discharges: C-14 and H-3. Aqueous Discharges: N/A. see Section 7. Gaseous discharges: N/A. Aqueous Discharges: N/A. see Section 7. Gaseous discharges: Ar-41, C-14, H-3. Aqueous Discharges: H-3. see Table 5.19 and Table Gaseous discharges: C-14. Aqueous Discharges: N/A. see Section 7. No additional radionuclides identified as a result of this requirement. Gaseous discharges: Noble gases (enhanced noble gas activity would indicate possible fuel pin failure or failure of the OG system). Iodine is also an indicator of possible fuel pin failure. Aqueous discharges: No additional radionuclides have been identified that would act as indicators of plant performance. Gaseous and Aqueous discharges: Beta/gamma particulate and alpha particulate discharges are likely to be below limit of detection (LoD)*. Note. * Radionuclides within the Beta-gamma particulate in air group for gaseous discharges and the Other (beta-gamma) radionuclide group for aqueous discharges are likely to be discharged from the Power Station at levels that will not be detectable within samples taken for measurement. Both radionuclide groups make contributions to the dose to the most exposed person (the Farming Family infant for gaseous discharges and the Fishing Farming family adult for aqueous discharges) of the order of 10-5 µsv/y to these individuals (see Tables A.69 and A.71 of Appendix O). As these doses are below the threshold of 1 µsv/y used as one of the criteria for setting discharge limits [RD31] it is proposed that discharge limits are not set for either of these radionuclide groupings but that BAT be applied to minimise their discharge to the environment. Details on the respective BAT to be applied for beta-gamma emitters in both effluent streams are provided in the BAT Case (Appendix C). Page 100 of 255

115 322. In summary, the significant radionuclides/radionuclide groups for the UK ABWR are as follows: Gaseous discharges: - Noble gases (excluding Argon-41); - Argon-41; - Carbon-14; - Tritium; and, - Iodine. Aqueous discharges: - Tritium. 5.6 Significant Radionuclides Summary of Discharge Estimates, Limits and QNLs 323. The best estimate annual discharges, annual discharge limits, and QNLs for the UK ABWR s significant radionuclides and radionuclide groups are given in Table 5.19 for gaseous discharges and Table 5.20 for aqueous discharges. The best estimates are based upon the operation of two UK ABWRs, the Rw/B and the site waste facilities, and the occurrence of a single Expected Event over any 12 month period. The discharge limits are also based upon the operation of two UK ABWRs, the Rw/B and site waste facilities, but with headroom factors applied, and the occurrence of two Expected Events over the 12 month period. Table 5.19 Summary of the Gaseous Annual Estimates, Discharge Limits and QNLs Radionuclide Best Estimate Annual Discharges (GBq/y) Annual Limit (GBq) Quarterly Notification Level (GBq) Tritium 5.5E E E+03 Carbon E E E+02 Argon E E E+03 Iodines 6.4E E E-01 Noble Gases (excluding Argon-41) 2.2E E E+01 Table 5.20 Summary of the Aqueous Annual Estimates, Discharge Limits and QNLs Radionuclide Best Estimate Annual Discharges (GBq/y) Annual Limit (GBq) Quarterly Notification Level (GBq) Tritium 4.0E E E+02 Page 101 of 255

116 5.7 Comparison of Power Station Discharge Data with Operational Experience 324. Part 4a of the NRW s application form RSR-B3 requires that discharge estimates are supported with performance data from similar facilities. The proposed discharge levels and annual limits for the Power Station described above have therefore been compared with OPEX which was gathered as part of GDA [RD22] As part of the GDA, the Environment Agency (EA) commissioned a study collating and presenting baseline data and relevant operational information for radiological discharges from existing BWRs around the world [RD22]. The study used data from 46 reactors across 24 sites, including reactors in Europe, the US and Japan, over the period 2005 to These reactors predominantly commenced operation after 1980, and therefore were considered appropriate to establish baseline data for BWR performance In order to enable comparison of the Power Station s discharges with the OPEX data, the 12 month rolling average discharge and proposed annual discharge limits were normalised to GWeh by dividing them by the expected energy generation for the station. For a single UK ABWR, the expected energy generation is calculated by multiplying 1.35 GWe (the output of the UK ABWR) by 8,760 hours (the number of hours in a year), giving 11,826 GWeh per unit per annum. The expected energy generation for the Power Station is 23,652 GWeh, as it comprises two UK ABWRs Gaseous Discharges Noble Gases (including Argon-41) 327. The annual mean and range of gaseous noble gas discharges for BWRs between 2005 and 2013 (from the available data) are shown in Figure 5.3. The discharges include Ar-41 as this was not identified as a separate radionuclide in the source data. The 12 month rolling average discharge and proposed annual discharge limit for noble gases and Ar-41 from the Power Station have therefore been combined, and are also shown in Figure 5.3. All data are normalised to Bq/GWeh. It can be seen that the discharges of noble gases and Ar-41 from the Power Station are comparable with those from other BWRs, i.e. both the discharge estimate and limit are slightly below the mean BWR value for each year shown. Page 102 of 255

117 Radioactivity discharged per unit of energy (Bq/ GWeh) Wylfa Newydd Project - Radioactive Substances Regulation Environmental Permit Application Figure 5.3 Power Station Gaseous Noble Gas and Ar-41 Discharge vs BWRs 1.00E E E E E E E E E Year Boiling Water Reactors (mean with range indicators) Proposed limit Best estimate annual discharge Carbon With respect to C-14, the annual mean and range of gaseous discharges for the BWRs is shown in Figure 5.4 along with the 12 month rolling average discharge and proposed annual discharge limit for the Power Station. It can be seen that the best estimate discharge for the Power Station is within the range of discharges reported by other BWRs, albeit above the mean values. However, the proposed discharge limit for C-14 has been set at a level which is higher than the range of discharges reported for other BWRs The issue regarding why C-14 discharges from the UK ABWR appear high when compared with mean discharges from comparable reactors was addressed at GDA. It was concluded that the estimate of the UK ABWR s annual gaseous discharge of C-14, which is based upon conservative assumptions and the present best estimate source term, is conceivably a maximum. However, the conservatism of the estimate is appropriate for the human and non-human dose assessment, as presented in Section 7. Data acquired following future operation of the UK ABWR over several cycles will enable direct comparison with the BWR performance data. Page 103 of 255

118 Radioactivity discharged per unit of energy (Bq/ GWeh) Radioactivity discharged per unit of energy (Bq/ GWeh) Wylfa Newydd Project - Radioactive Substances Regulation Environmental Permit Application Figure 5.4 Power Station Gaseous Carbon-14 Discharge vs BWRs 1.00E E E E Year Boiling Water Reactors (mean with range indicators) Proposed limit Best estimate annual discharge Tritium 330. Comparable data for gaseous H-3 discharges are presented in Figure 5.5. Once again it can be seen that the best estimate discharge for the Power Station is within the range of discharges reported by other BWRs, although above the mean values. The Power Station s proposed discharge limit for H-3, however, has been set at a level which is generally higher than the range of discharges reported for other BWRs. Figure 5.5 Power Station Gaseous Tritium Discharge vs BWRs 1.00E E E E E E E E Boiling Water Reactors (mean with range indicators) Proposed limit Best estimate annual discharge Page 104 of 255

119 Radioactivity discharged per unit of energy (Bq/ GWeh) Wylfa Newydd Project - Radioactive Substances Regulation Environmental Permit Application Iodine 331. Data for iodine are presented in Figure 5.6. It can be seen that the discharges of iodines from the Power Station are comparable with those from other BWRs, i.e. that both the discharge estimate and limit are generally below the mean BWR value for each year shown. Figure 5.6 Power Station Gaseous Iodine Isotope Discharge vs BWRs 1.00E E E E E E E E E E E Year Boiling Water Reactors (mean with range indicators) Proposed limit Best estimate annual discharge Page 105 of 255

120 5.7.2 Aqueous Discharges Tritium 332. Comparable data for aqueous H-3 discharges are presented in Figure 5.7. It can be seen that the Power Station s best estimate discharge and proposed discharge limit are below the mean and largely within the range of discharges reported for other BWRs. Radioactivity discharged per unit of energy (Bq/ GWeh) Figure E E E E E+05 Power Station Aqueous Tritium Discharge vs BWRs Year Boiling Water Reactors (mean with range indicators) Proposed limit Best estimate annual discharge 5.8 Solid Waste 333. The sub-sections below present estimates of solid waste arisings and disposals during normal operation. These are given by category and physico-chemical characteristics. An indication is also provided of the likely waste arisings during decommissioning (albeit that this Application does not cover the disposal of wastes from that activity). The waste data presented are primarily sourced from the GDA, Radioactive Waste Management Arrangements document [RD23] There is the potential for some radioactive waste to arise prior to normal operations (see Section 2.5). However, it is expected that the majority of radioactive waste will be generated once nuclear fuel is brought onto the site as part of the active commissioning of Unit 1. This is the start point of normal operations from a waste and discharges perspective. The UK ABWR is designed to operate for a period of 60 years and Unit 2 will lag Unit 1 by approximately 16 months so the normal operations period is estimated to cover 62 years from the arrival of first fuel on site. Page 106 of 255

121 5.8.1 Radioactive Waste Arisings Immediately following Shutdown 335. On final shutdown the reactors will be fully defueled and the last batch of spent fuel will remain in the SFP within the R/B for a period of approximately five years. Only once the last of the spent fuel has been removed from the pool, packaged into casks and transferred to the SFSF, will the decommissioning of the reactors fully commence It is anticipated that some post operational clean out (POCO), de-planting and decommissioning will commence in parallel with the five year fuel cooling period, subject to appropriate safety case constraints. During this interim stage numerous systems will cease operation although at present a discrete strategy for the progressive shutdown of systems has not been defined The shutdown process will affect both gaseous and aqueous discharges and the rates and quantities of generation of radioactive wastes. As this aspect remains undefined at present, this Application does not include radioactive discharges or waste arisings during this period. Further work is required to quantify radioactive discharges and wastes for the period immediately following shutdown once a shutdown strategy has been determined Radioactive Waste Arisings during the Normal Operations Phase 338. An overview of the radioactive waste arisings during the normal operations phase is given in Table Further detail is given in the subsequent tables. Quantities are calculated for the Power Station on the basis that the site will comprise two UK ABWR reactors and a number of common facilities It should be recognised that all of the waste quantities and specific activities presented are Source Term estimates based on operational experience, assumptions and calculations. The actual quantities and activities will be dependent on the future operational regime and history of the site and, as such, the values quoted are subject to a degree of uncertainty. Unless specifically identified as cross-boundary waste, the uncertainties associated with specific activity and normal operations are not expected to alter the waste classification The design of the solid radioactive waste processing facilities on the UK ABWR is not yet finalised. Therefore, data on final waste package types and quantities will be developed as the detailed design and the operational philosophy develops. Data is provided for gross raw waste volumes and total activities Horizon is not applying for specific limits on the volume of solid radioactive waste to be disposed of by transfer from the Power Station site or the activity of such solid radioactive waste transfers/disposals. The disposal of solid radioactive waste will be subject to regulatory scrutiny through Horizon s EP-RSR requiring the application of BAT to minimise the activity in waste generated on the premises and to minimise the volume of radioactive waste disposed of. Page 107 of 255

122 Table 5.21 Radioactive Waste Arisings during the Normal Operations Phase Category (Generated) Category (Disposed) Stream Packaging Anticipated Disposal Route Volume (m 3 ) Mass (t) Total Activity (TBq) Activity Concentration (TBq/t) Data High Level Radioactive Waste Spent Nuclear Fuel HLW Spent Nuclear Fuel Spent fuel assemblies packaged for disposal in GDF detail to be defined. GDF (HLW) 1.69 x x x x 10 3 Table 5.23 Intermediate Level Radioactive Waste ILW ILW Sludge (crud) 3 m 3 RWM approved drums containing solidified sludge (crud), powder resin and sludge (crud) and cement waste form. GDF (ILW) x Table 5.24 ILW ILW Powder resin and Sludge (crud) 3 m 3 RWM approved drums containing solidified powder resin and Sludge (crud, sludge (crud) and cement waste form. GDF (ILW) x x 10 1 Table 5.25 HLW ILW Control Rods RWM approved containers containing control rod sections immobilised in a cementitious grout matrix. GDF (ILW) x x 10 3 Table 5.26 Page 108 of 255

123 Category (Generated) Category (Disposed) Stream Packaging Anticipated Disposal Route Volume (m 3 ) Mass (t) Total Activity (TBq) Activity Concentration (TBq/t) Data HLW ILW Reactor Components Low Level Radioactive Waste Homogenous LLW RWM approved containers containing reactor component sections immobilised in a cementitious grout matrix. GDF (ILW) x x 10 3 Table 5.27 LLW LLW Bead Resin Third Height ISO containers (THISO) containing solidified resin and cement waste form. LLWR 1, , x x 10-2 Table 5.28 LLW LLW Concentrates (Sludge) THISO containers containing solidified sludge and cement waste form. LLWR x x 10-5 Table 5.29 LLW LLW/ VLLW HVAC Filters Waste packaging will be in compliance with the WAC of the receiving facility. LLWR or Landfill 2, x x 10-6 Table 5.30 LLW LLW Spent Filter Media Waste packaging will be in compliance with the WAC of the receiving facility. LLWR or Landfill x x 10-3 Table 5.31 Page 109 of 255

124 Category (Generated) Category (Disposed) Stream Packaging Anticipated Disposal Route Volume (m 3 ) Mass (t) Total Activity (TBq) Activity Concentration (TBq/t) Data Heterogeneous LLW LLW/VLLW LLW/VLLW Miscellaneous Combustible/ Noncombustible Waste packaging will be in compliance with the WAC of the receiving facility. Metal melting, Incineration, Compaction, Landfill or LLWR 6,732 3, x x 10-3 Table 5.32 Potential LLW LLW LLW Oils/Oily Waste Determined on a case by case basis through BAT assessment. Incineration N/A N/A N/A N/A Table 5.33 LLW LLW Radioactively Contaminated Land Determined on a case by case basis through BAT assessment. LLWR or Landfill N/A N/A N/A N/A Table 5.34 Decommissioning Solid Radioactive Waste Arisings ILW LLW ILW LLW Mixed see Table 5.35 Mixed see Table 5.35 GDF LLWR N/A N/A N/A N/A Table 5.35 VLLW VLLW Landfill Page 110 of 255

125 342. Detailed information on radioactive waste arisings during the normal operations phase is presented in the tables below. Each of the tables is set out as per Table Table 5.22 Waste Stream Metadata Parameter Information Waste Stream Waste Category Waste Origin Waste Characteristics Annual Arisings for 2 units Proposed Packaging Anticipated Disposal Route Key Radionuclides and specific activity Total Activity for 2 units Total Arisings for 2 units Description of the waste stream. Spent Nuclear Fuel, HLW, ILW, LLW (including VLLW). Information on source(s) of arising/provenance of the waste. Physico-chemical information on the waste. Stated as mass/volume dependent on nature of the waste and information available total calculated annual arising covering normal operation, start-up, shutdown and major maintenance activities for 2 units. Anticipated processing and packaging for treatment/disposal. GDF, LLWR, permitted processing facility, permitted landfill etc. Typically those radionuclides that constitute >90% of total radioactivity, Total specific activity stated in Bq/g. The data presented represents the total specific radioactivity based upon the dominant (key) radionuclides in the waste. Stated as mass/volume dependent on nature of the waste and information available total calculated arising covering the 60 year normal operations period (2 x 60 years) The waste categories are defined as follows: HLW: Decay heat thermal power >2 kw/m 3 ; ILW: Decay heat thermal power <2 kw/m 3, specific activity >12 GBq/t βγ and 4 GBq/tα; LLW: Specific activity <12 GBq/t βγ and 4 GBq/tα; and, VLLW is a sub-set of LLW [RD24]. Page 111 of 255

126 Table 5.23 Radioactive Waste Arisings Spent Nuclear Fuel Parameter Waste Stream Waste Category Waste Origin Waste Characteristics Annual Arisings for 2 units Proposed Packaging Anticipated Disposal Route Key Radionuclides Total Activity for 2 units Total Arisings for 2 units Information Horizon s strategy is to place spent fuel into dry storage after a period of cooling in the SFP. The spent fuel will be stored on site until the GDF technical solution for disposal of spent fuel has been made available (assumed to be provided by RWM). It is assumed that the fuel will be retrieved and repackaged into a disposable form at this point. In line with the UK DEFRA definition of waste [RD25] this step will clearly indicate the intent to discard spent fuel and the definition of spent fuel will then be changed to identify it as a radioactive waste. Due to its radiogenic heat output the spent fuel will be categorised as HLW in accordance with the definition provided in the joint regulatory guidance on the management of higher activity wastes [RD26]. Initially categorised as spent nuclear fuel. Will be re-categorised as HLW once repackaged for disposal. The UK ABWR fuel assembly comprises uranium dioxide ceramic fuel pellets housed in a zircaloy cladding to form fuel rods. A 10 x 10 array of fuel rods is held in a bundle comprising tie plates, spacers and a debris filter. The bundle is housed in a zircaloy channel box to form the fuel assembly. The fuel undergoes fission in the reactor core to produce heat. The FPs and actinides produced during the fission process are considered as waste. In addition to the FPs and actinides, the structures of the fuel assemblies become activated. The spent fuel assembly comprises a 10 x 10 array of zircaloy clad fuel rods held together with steel components and housed in a zircaloy channel box. The fuel assemblies are c. 4.5 m in length and c m square, with a gross mass of 298 kg. Each fuel assembly occupies m 3. It is assumed that 224 spent fuel assemblies per reactor will be replaced at each outage, and outages are assumed to occur once every 18 months. This will result in the generation of 448 spent fuel assemblies every outage. Spent fuel will initially be cooled in the SFP before being transferred to storage casks. It will be repackaged into disposal containers at a later stage once a disposal container design has been specified. It is currently assumed that RWM will provide this specification. It is assumed that, in line with government policy, spent nuclear fuel from new nuclear power stations will not be reprocessed but will be disposed of once disposal facilities are available. It is further assumed that the GDF will provide a disposal route for spent nuclear fuel and that the spent fuel will be categorised as HLW due to its radiogenic heat output. Activity data for spent nuclear fuel is derived from the GDA disposability assessment for wastes and spent fuel [RD27]. Data used for a spent fuel assembly (SFA) is based on average burn-up fuel (50 GWD/tU) after 61 years of cooling. Spent Fuel: Cs-137, Sr-90, Pu-241, Am-241, Pu-238, Pu-240, Co-60; 5.46 x 10 2 Tbq/SFA Total activity (SFAs): 9,600 x 2 x 5.46 x 10 2 = 1.05 x 10 7 Tbq Total SFAs: 224 units per cycle per reactor = 19,200 Gross mass: 19,200 x 298 kg = 5.72 x 10 3 t Gross volume: 19,200 x m 3 = 1.69 x 10 3 m 3 Page 112 of 255

127 Solid Intermediate Level Radioactive Waste 344. Solid Intermediate Level Radioactive Waste (ILW) arising from the Power Station will generally be produced as a result of discrete processes and can therefore be considered as homogenous and possessing consistent physio-chemical characteristics throughout the period of its production. The waste will include both wet-solid and dry-solid wastes The list of solid ILW wastes includes one potential arising relating to generation of redundant fuel assembly channel boxes. This will only occur if a channel box has to be removed due to deformation. In this instance the fuel bundle would be transferred to a replacement channel box and the empty channel box would become a radioactive waste item Technical information on channel boxes is included under spent nuclear fuel (see Table 5.23) since they are part of the fuel assembly. The strategy for their management as a discrete waste stream would be the same as for other activated components (control rods and reactor components). Since this is only a potential arising no estimate is provided for annual or total quantities. Page 113 of 255

128 Table 5.24 Radioactive Waste Arisings Sludge (Crud) ILW Parameter Waste Stream Waste Category Waste Origin Waste Characteristics Annual Arisings for 2 units Proposed Packaging Anticipated Disposal Route Key Radionuclides Total Activity for 2 units Total Arisings for 2 units Information The wastes will arise as a sludge from backwashing of spent filter media in the reactor coolant cleaning circuits, and from the cleaning of the SFP. The sludge will be discharged to a Sludge Storage Tank in the Rw/B and accumulated for short term storage prior to processing. Wet-solid ILW. Crud is made up of CPs (predominantly ferrous) that will arise during reactor operations. The CPs will form as particulates that are entrained in the coolant circuit filtrations systems. The spent filter media provide crossflow ultrafiltration of the coolant and the waste will be generated as sludge through backwashing of the Condensate Filter (CF) and Low Chemical Impurity Waste (LCW) filters. Sludge formed mainly of ferrous corrosion particles with a dry bulk density of approx t/m 3. Particle size distribution is broadly 80% (by volume) <5 µm. The sludge is assumed to arise as an aqueous slurry with a wet density of 1.1 t/m 3. Estimated annual arisings: CF Filter: LCW Filter: Total: 2.4 m 3 /year, 2.64 t/year 0.6 m 3 /year, 0.66 t/year 3.0 m 3 /year, 3.3 t/year Sludges will be transferred to the Wet-solid ILW Processing System where they will be copackaged with the spent ILW powder resin and sludge (crud) and solidified in cement within 3 m 3 unshielded stainless steel drums. Packaging will comply with RWM waste packaging specifications and will be substantiated through RWM s disposability assessment and letter of compliance process. Disposal as ILW to GDF. CF: Fe-55, Co-60, Mn-54; 6.1 x 10 5 Bq/g LCW: Fe-55, Co-60, Mn-54; 7.8 x 10 6 Bq/g CF: 60 x 6.1 x 10 5 Bq/g x 2.64 x 10 6 g = 9.66 x 10 1 TBq LCW: 60 x 7.8 x 10 6 Bq/g x 0.66 x 10 6 g = 3.09 x 10 2 TBq Total: 9.66 x 10 1 TBq x 10 2 TBq = 4.06 x 10 2 TBq Annual arisings are calculated based on equal proportions of material being generated by each unit. Therefore total arisings for the normal operations period is calculated as 60 x annual arisings; CF Filters: LCW Filters: Total: 60 x 2.4 m 3 = 144 m 3, 60 x 2.64 t = t 60 x 0.6 m 3 = 36.0 m 3, 60 x 0.66 t = 39.6 t = m 3, = t Page 114 of 255

129 Table 5.25 Parameter Waste Stream Waste Category Waste Origin Waste Characteristics Annual Arisings for 2 units Proposed Packaging Anticipated Disposal Route Key Radionuclides Total Activity for 2 units Total Arisings for 2 units Radioactive Waste Arisings Powder Resin and Sludge (crud) ILW Information The wastes will arise as a resin/sludge mixture from backwashing of filter demineralisers (FD) in the reactor coolant cleaning circuits. The waste will initially be discharged to receiver tank in the R/B before being transferred and accumulated in a Sludge Storage Tank in the Rw/B for short term storage prior to processing. Wet-solid ILW This material arises from the FD in the CUW and Fuel Pool Clean-up System (FPC). The FPC system also treats water from the Suppression Pool Clean-up System (SPCU). The FD is a cartridge element (stainless steel wedge wire cylinder) pre-coated in powdered resin. The powder resin is a cross-linked polystyrene matrix with a dry bulk density of approx. 0.2 t/m 3. The FD uses a mixed bed Cation/Anion powder coat resin (1:1 ratio). The resin beds also act as filters and will entrain particulate corrosion product (crud). The sludge is assumed to arise as an aqueous slurry with a wet density of 1.1 t/m 3. Estimated annual arisings: CUW FD: FPC FD: Total: 6.2 m 3 /year, 6.82 t/year 2.6 m 3 /year, 2.86 t/year 8.8 m 3 /year, 9.68 t/year The resin and sludge waste will be transferred to the Wet-solid ILW Processing System, where it will be co-packaged with the ILW sludge (crud) and solidified in cement within 3 m 3 unshielded stainless steel drums. Packaging will comply with RWM waste packaging specifications and will be substantiated through RWM s disposability assessment and letter of compliance process. Disposal as ILW to GDF. CUW FD: Fe-55, Co-60, Mn-54; 1.3 x 10 8 Bq/g FPC FD: Fe-55, Zn-65, Co-60, Mn-54; 1.7 x 10 7 Bq/g CUW FD: 60 x 1.3 x 10 8 Bq/g x 6.82 x 10 6 g = 5.32 x 10 4 TBq SPC FD: 60 x 1.7 x 10 7 Bq/g x 2.86 x 10 6 g = 2.92 x 10 3 TBq Total: 5.32 x 10 4 TBq x 10 3 TBq = 5.61 x 10 4 TBq Annual arisings are calculated based on equal proportions of material being generated by each unit. Therefore total arisings for the normal operations period is calculated as 60 x annual arisings; CUW FD: SPC FD: Total: 60 x 6.2 m 3 = m 3, 60 x 6.82 t = t 60 x 2.6 m 3 = m 3, 60 x 2.86 t = t = m 3, = t Page 115 of 255

130 Table 5.26 Radioactive Waste Arisings Control Rods ILW Parameter Waste Stream Waste Category Waste Origin Waste Characteristics Annual Arisings for 2 units Proposed Packaging Anticipated Disposal Route Key Radionuclides Total Activity for 2 units Information The UK ABWR has two types of control rods, hafnium rods for operational control of the reactor, and boron carbide rods for shutdown of the reactor. Each core houses 205 control rods, comprised of 29 hafnium rods and 176 boron carbide rods. The control rods are used to control reactivity in the reactor core and are therefore exposed to the full neutron flux and become activated during their service life. Control rods are initially categorised as HLW due to their high radiogenic heat loading. Following decay storage the waste can be re-categorised as Dry-solid ILW. Spent control rods are removed during reactor outages and are handled underwater from the reactor into the SFP. Due to the limited space in the SFP, and the requirement to retain operational resilience throughout the normal operations period, Horizon s strategy is to remove the control rods and place into dry cask storage. They are of a cruciform cross-section with the neutron absorber material contained in the wings of the cruciform. The cruciform itself is fabricated from stainless steel. Control rods are c.4.1 m in length and c m in cross-section. Hafnium control rods consist of hafnium plates encased in a stainless steel outer, with a maximum gross weight of c.110 kg. Boron carbide control rods are made up of tubes containing boron carbide powder encased in a stainless steel outer, with a maximum gross weight of c.90 kg. Spent hafnium control rods will be generated at a rate of about five per year per reactor for the lifetime of the reactor. Boron carbide rods have an expected life up to 40 years and will therefore need to be replaced after this period. It is therefore assumed that during the normal operations period one full set of boron carbide rods per reactor will be generated as radioactive waste, with a further set generated at end of life as a decommissioning waste arising. Hafnium control rods: 10 x 110 kg = 1.1 t, 4.1 m x m x m = 0.26 m 3 Boron Carbide control rods: zero annual arisings, reported under total arisings as only changed once during the normal operations phase. Following an initial cooling period in the SFP the control rods will be packaged into dry storage casks for decay storage. They will be transferred to an on-site HLW Decay Store where they will be stored until the radiogenic heat load drops to ILW levels. The waste will then be recovered from the casks, size reduced and packaged into suitable waste containers. The waste will be immobilised using a cementitious grout matrix. Packaging will comply with RWM waste packaging specifications and will be substantiated through RWM s disposability assessment and letter of compliance process. Disposal as ILW to GDF. The waste is HLW on arising but will be decay stored and managed as ILW for disposal. Co-60, Ni-63 Hafnium control rods: 2.1 x 10 9 Bq/g Boron carbide control rods: 7.3 x 10 8 Bq/g Hafnium control rods: 2 x 324 x 110 x 10 3 x 2.1 x 10 9 = 1.5 x 10 5 TBq Boron carbide control rods: 2 x 176 x 90 x 10 3 x 7.3 x 10 8 = 2.31 x 10 4 TBq Total: 1.5 x 10 5 TBq x 10 4 TBq = 1.73 x 10 5 TBq Page 116 of 255

131 Parameter Total Arisings for 2 units Information Hafnium control rods (648 hafnium control rods during the normal operations phase): 648 x 110 kg = t, 648 x (4.1 m x m x m) = m 3 Boron carbide control rods (352 boron carbide control rods during the normal operations phase): 352 x 90 kg = t, 352 x (4.1 m x m 2 ) = 91.6 m 3 Total waste quantities for two units (952 control rods during the normal operations phase): = t, = m 3 Page 117 of 255

132 Parameter Table 5.27 Radioactive Waste Arisings Reactor Components ILW Information Waste Stream Waste Category Waste Origin Waste Characteristics Annual Arisings Proposed Packaging Anticipated Disposal Route Key Radionuclides These activated metallic components arise from the operation of the reactor and consist of a number of items which are principally used to monitor in-core reactivity. These will include Start-up Range Neutron Monitoring system (SRNM including dry tubes), Local Power Range Monitors (LPRMs), and Traversing in core Probes (TIPs). Some additional in-core components are only used during outages (e.g. blade guides) and will therefore be less active than items exposed to the full neutron flux. Wherever reasonably practicable opportunities will be taken to segregate and re-classify waste as LLW. Reactor components are housed in stainless steel outer casings and are initially categorised as HLW due to their high radiogenic heat loading. Following decay storage the waste can be re-categorised as Dry-solid ILW. Operational in-core equipment that will arise as waste periodically when items of in-core equipment are replaced. The waste will generally arise during planned reactor outages. The waste consists of a range of standard reactor components that may require periodic replacement. The redundant units will become radioactive waste due to the levels of activation associated with their use in the reactor core. This stream includes the following items: LPRM: 24 kg, 0.02 m 3 SRNM: 24 kg, m 3 TIP: 4 kg, m 3 Neutron Sources: 4 kg, m 3 All items are small diameter ( 40mm), LPRM, SRNM and TIP have an overall length of approximately 16m and are assumed to be cut into 4 equal length pieces of 4m once removed from the reactor. The neutron sources are approximately 4 m in length. Items are replaced periodically but total arisings of this stream are relatively small. Due to the infrequent nature and relatively small quantity of arisings of these wastes they are reported under total arisings only. Following an initial cooling period in the SFP the waste will be packaged into dry storage casks for decay storage. They will be transferred to an on-site SFSF where they will be stored until radiogenic heat load drops to ILW levels. The waste will then be recovered from the casks, size reduced and packaged into RWM approved 3 m 3 boxes. The waste will be immobilised using a cementitious grout matrix. Packaging will comply with RWM waste packaging specifications and will be substantiated through RWM s disposability assessment and letter of compliance process. Disposal as ILW to GDF. The waste is HLW on arising but will be decay stored and managed as ILW for disposal. LPRM: Co-60, Ni-63; 1.3 x 10 9 Bq/g SRNM: Co-60, Ni-63; 3.0 x 10 9 Bq/g TIP: Co-60, Ni-63; 7.2 x 10 6 Bq/g Neutron Source: Body: Co-60, Ni-63; 3.7 x 10 8 Bq/g Source: Ni-63, Cf-252; 16 Gbq/unit Page 118 of 255

133 Parameter Total Activity Total Arisings Information LPRM: 57.6 x 10 6 g x 1.3 x 10 9 Bq/g = 74.9 x 10 3 TBq SRNM: 2.88 x 10 6 g x 3.0 x 10 9 Bq/g = 8.64 x 10 3 TBq TIP: 0.14 x 10 6 g x 7.2 x 10 6 Bq/g = 1.04 TBq Neutron Source: Body: 0.04 x 10 6 g x 3.7 x 10 8 Bq/g = 14.8 TBq Sources - Ni-63 and Cf-252: 10 units x 16 Gbq/unit = Gbq Total: 8.35 x 10 4 TBq For 2 units: LPRM (40 units replaced per year): 40 x 60 x 24 kg = 57.6 t, 40 x 60 x 0.02 m 3 = 48.2 m 3 SRNM (on average 2 units replaced per year 20 units replaced every 10 years): 2 x 60 x 24 kg = 2.88 t, 2 x 60 x m 3 = 1.36 m 3 TIP (on average 0.6 units replaced per year 6 units replaced every 10 years): 0.6 x 60 x 4 kg = 0.14 t, 0.6 x 60 x m 3 = m 3 Neutron Source (on average units replaced per year 10 units replaced per 60 years): x 60 x 4 kg = 0.04 t, x 60 x m 3 = m 3 Total volume: 50.0 m 3 Total Mass: 60.7 t Page 119 of 255

134 Solid Low Level Radioactive Waste 347. Solid Low Level Radioactive Waste (LLW) arising from the Power Station is considered in three forms: Homogenous Waste Arising resultant waste from a specific process which displays consistent physico-chemical characteristics. Includes both wet-solid and dry-solid wastes; Heterogeneous Waste Arising resultant waste from general Power Station activities and has variable physico-chemical characteristics; and, Potential Waste Arising not expected to arise but is included in arrangements to provide capability should it occur as a result of process failures or accidents This Application provides indicative gross raw waste volumes for disposal of LLW. It should be noted that final decisions on choice of waste containers is in part dependent on meeting the WAC of the receiving facility and may therefore be subject to change. Therefore, in the majority of cases detailed information on proposed packaging has not been provided in the individual tables. However, it is likely that containers listed in the LLWR Packaging Services Brochure [RD30] will be used for lower activity wastes (LAW) The list of LLW wastes includes two potential arisings relating to generation of radioactively contaminated oils (or oily wastes) and radioactively-contaminated land. Horizon s strategy is to avoid the generation of these wastes through appropriate containment of radioactive material at all times. However, historical operating experience on nuclear licensed sites indicates that the possibility of producing this form of waste cannot be disregarded. The wastes are therefore included, together with an explanation of Horizon s strategy for managing them should they arise, but no quantification is provided since the intent is to produce none Four homogenous LLW streams are identified: Bead Resin; Evaporator Concentrate (sludge); HVAC Filters; and, Spent Filter Media Only one heterogeneous LLW stream is identified while there are two potential LLW streams: Radioactive Contaminated Oils and Oily Wastes; and, Radiologically Contaminated Land. Page 120 of 255

135 Table 5.28 Radioactive Waste Arisings Bead Resin LLW Parameter Waste Stream Waste Category Waste Origin Waste Characteristics Annual Arisings for 2 units Proposed Packaging Anticipated Disposal Route Key Radionuclides Information The waste is either a styrene divinylbenzene copolymer or cross-linked polystyrene matrix in a fine bead form. The resin is used in reactor coolant clean-up systems to remove soluble impurities (both radioactive and non-radioactive). The purpose of the clean-up systems is to maintain high purity reactor coolant to minimise reactor circuit corrosion and risk of material deposition that could lead to plant operation problems and radiation hotspots. The waste is categorised as being cross boundary ILW/LLW on generation but due to the relatively short half-lives of the principal radionuclides Horizon s strategy is to decay store the waste in tanks in the Rw/B until it can be processed to meet LLWR activity limits and radioactive waste transport dose limits. Wet-solid LLW. Assessed as being cross boundary ILW/LLW based on anticipated Co-60 and Fe-55 loading which means wastes initially exceed 12 Gbq/t limit for beta gamma activity and transport dose rates. Bead ion exchange resins are used in the demineralisers in the LCW and HCW in the Rw/B and the CD in the TB. The resin beds are non-regenerative and will therefore be depleted over time. The spent bead resin is discharged by backwashing the demineralisers and transferred to storage tanks in the Rw/B for decay storage prior to processing. CD Mixed bed anion/cation styrene divinylbenzene copolymer resin in bead form. Average particle size is 0.5 mm and dry bulk density is approx to 0.4 t/m 3 LCW, HCW Cross-linked polystyrene matrix in bead form. Average particle size is 1.0 mm and dry bulk density is approx to 0.4 t/m 3. Wet bead resin is assumed to arise as an aqueous slurry with a wet density of t/m 3. This waste will arise routinely through normal operations and is calculated based on an averaged cycle that includes start-up and shutdown activities in addition to normal operations. CD: LCW: HCW: Total: 2 x 9.6 m 3 /year = 19.2 m 3, 19.2 m 3 x t/m 3 = t 2 x 1.7 m 3 /year = 3.4 m 3, 3.4 m 3 x t/m 3 = 3.96 t 2 x 0.3 m 3 /year = 0.6 m 3, 0.6 m 3 x t/m 3 = 0.70 t 23.2 m 3, t The spent bead resin will be decay stored underwater in tanks in the Rw/B. Once the waste has decayed to LLW levels it will be transferred to the Wet-solid LLW Processing System (within the Rw/B) for processing. The waste will be in-line mixed with a preprepared cementitious grout and poured into THISO containers. LLWR CD: Co-60, Fe-55, Ce-144; 4.3 x 10 4 Bq/g LCW: Co-60, Fe-55, Zn-65, Mn-54, Co-58; 2.1 x 10 5 Bq/g HCW: Co-60, Ce-144, Fe-55; 6.83 x 10 0 Bq/g Total Activity for 2 units Total Arisings for 2 units CD: LCW: HCW: Total: CD: LCW: HCW: Total: 60 x x 10 7 g x 4.3 x 10 4 Bq/g = 5.77 x 10 1 TBq 60 x 3.96 x 10 6 g x 2.1 x 10 5 Bq/g = 4.99 x 10 1 TBq 60 x g x 6.83 x 10 0 Bq/g = 2.87 x 10 2 MBq 1.08 x 10 2 TBq 60 x 19.2 m 3 = 1,152.0 m 3, 60 x t = t 60 x 3.4 m 3 = m 3, 60 x 3.96 t = t 60 x 0.6 m 3 = 36.0 m 3, 60 x 0.70 t = 42.0 t 1,392.0 m 3, t Page 121 of 255

136 Parameter Table 5.29 Radioactive Waste Arisings Concentrates (Sludge) LLW Information Waste Stream Waste Category Waste Origin Waste Characteristics Annual Arisings Proposed Packaging Anticipated Disposal Route Key Radionuclides Total Activity The HCW system utilises an evaporator to remove insoluble solids from various Power Station water systems the system is used for effluent streams that are too high in impurities for treatment by demineralisation. The evaporator concentrate is removed from the evaporator as a dilute sludge and accumulated for storage in a concentrates waste tank in the Rw/B. Wet-solid LLW The HCW system processes effluents from sump tanks, building drainage systems, laboratory drains and decontamination showers. Insoluble solids are concentrated by evaporation and transferred to a storage tank where they are accumulated prior to processing. Dilute sludge (c.6 wt%) made up of concentrated organic and inorganic particulate (rad and non-rad) in an aqueous slurry with a solids loading of approx. 60,000 ppm. The sludge is assumed to arise as an aqueous slurry with a wet density of 1.1 t/m 3. Estimated arisings for two units: 2.0 m 3 /year, 2.2 t/year The waste will be accumulated and stored underwater in tanks in the Rw/B transferred. Once sufficient process volumes have been accumulated it will be transferred to the Wetsolid LLW Processing System (within the Rw/B) for processing. The waste will be in-line mixed with a pre-prepared cementitious grout and poured into THISO containers. LLWR H-3, Fe-55, Co-60; 7.6 x 10 1 Bq/g 60 x 2.2 x 10 6 g x 7.6 x 10 1 Bq/g = 1.00 x 10 1 GBq Total Arisings 60 x 2.0 m 3 /year = 120 m 3 60 x 2.2 t = t Page 122 of 255

137 Table 5.30 Radioactive Waste Arisings HVAC Filters LLW Parameter Waste Stream Waste Category Waste Origin Waste Characteristics Annual Arisings Proposed Packaging Anticipated Disposal Route Information The radiation controlled areas of the nuclear island buildings are served by dedicated ventilation systems. The extracted air from these systems is subject to a number of airborne activity abatement techniques, before discharge to the environment. The filters used to abate airborne particulate are HEPA filters. Dry LLW/VLLW HVAC filters will arise from filter changing in HVAC systems servicing the R/B, T/B, Rw/B and S/B as well as waste treatment facilities. HVAC filters typically comprise of a metallic body and a glass fibre filter matrix. For nuclear ventilation systems these would be expected to conform to industry best practice for ventilation of radioactive areas [RD28]. The filter matrix will entrain a mixture of radioactive and non-radioactive airborne particulates. Specific filter types and sizes are not yet determined but are assumed to be of the most modern cylindrical safe change HEPA filter type advised as industry best practice [RD29]. Based on advice from a filter manufacturer the filtration surface area for this type of filter is 29.0 m 2 with a filter material mass of 89 g/m 2 which gives a total mass of filtration material of kg per filter. This figure has been used to calculate the waste inventory. HVAC filters are assumed to be changed every three years (or 20 times over the normal operations period). Filters will be changed on pre-determined differential pressure limits or on measured dose-rate, noting that filter shelf lives are typically 10 years so filters in some lower duty systems may require replacement less frequently. There are a total of 48 HVAC systems in the UK ABWR design covering the R/B, T/B, Rw/B and S/B in each reactor unit, which comprise a total of 546 individual filters. Each filter is assumed to have dimensions of Ø 520 mm and length 610 mm (calculated overall volume m 3 ) and a gross mass of 18 kg. Each unit has a single reactor building and turbine building. The service building and radwaste building are shared between the two units. Average annual HEPA filter arisings, based on a replacement frequency of three years, are: R/B x2: 312/3 = 104 filters T/B x2: 492/3 = 164 filters Rw/B x1: S/B x1: 72/3 = 24 filters Total 216/3 = 72 filters 1092/3 = 364 filters The total annualised arising for two units is therefore calculated to be: (1092 x 18 kg)/3 = t (1092 x m 3 )/3 = 46.9 m 3 This does not include HVAC filters from waste processing facilities for which designs are not yet finalised. It is assumed that these filters will be of the same type and specification as those stated here. On removal the filter module is loaded into suitable containers (in compliance with the WAC of the receiving facility). These are transferred to the LAWMF, followed by direct disposal to LLWR or where reasonably practicable transported to a UK treatment facility for volume reduction or incineration prior to disposal at LLWR. Where reasonably practicable (and demonstrated to be BAT) filters may be sentenced as VLLW and transferred to appropriately permitted facility. Off-site permitted supercompaction facilities, incineration facilities. permitted landfill, LLWR. Page 123 of 255

138 Parameter Key Radionuclides Total Activity Information R/B: Fe-55, Co-60, Mn-54, Zn-65; 2.1 x 10 2 Bq/g T/B: Fe-55, Co-60, Mn-54; 2.2 x 10-1 Bq/g Rw/B: Fe-55, Co-60, Zn-65, Mn-54; 2.3 x 10 1 Bq/g S/B: Key radionuclides and specific activity are assumed to be the same as T/B. Total activity is calculated by multiplying the waste loading of each filter by the stated specific activity. R/B: T/B: Rw/B: S/B: 2 x 156 x (60/3) x x 10 3 g x 2.1 x 10 2 Bq/g = 3.4 GBq 2 x 246 x (60/3) x x 10 3 g x 2.2 x 10-1 Bq/g = 5.6 MBq 1 x 216 x (60/3) x x 10 3 g x 2.3 x 10 1 Bq/g = 260 MBq 1 x 72 x (60/3) x x 10 3 g x 2.2 x 10-1 Bq/g = 820 KBq Total activity of all filters: 3.6 GBq Total Arisings Total arisings are calculated by multiplying the annual arisings by 60: R/B: 60 x 104 = 6,240 filters, 6,240 x m 3 = m 3, 6,240 x18 kg = t T/B: 60 x 164 = 9,840 filters, 9,840 x m 3 = 1, m 3, 9,840 x 18 kg = t Rw/B: S/B: 60 x 72 = 4,320 filters, 4,320 x m 3 = m 3, 4,320 x 18 kg = t 60 x 24 = 1,440 filters, 1,440 x m 3 = m 3, 1,440 x 18 kg = t Total: 21,840 filters, total volume m 3, total mass t, Page 124 of 255

139 Parameter Table 5.31 Radioactive Waste Arisings Spent Filter Media LLW Information Waste Stream Waste Category Waste Origin Waste Characteristics Annual Arisings for 2 units Proposed Packaging Anticipated Disposal Route Key Radionuclides Total Activity for 2 units Total Arisings for 2 units The filters remove insoluble impurities in the Condensate Water Clean-up system and LCW system. Spent filter media will be managed as a dry solid waste. The filters remove particulate and are periodically backwashed to remove the build-up of sludge (crud). The resultant sludge (crud) is a Wet-solid ILW stream. Dry LLW. Over a period of time the filters will lose efficiency and require replacement. The filter is backwashed and drained prior to removal of the filter elements. There are two filter installations in the plant: LCW and CF filters. The waste item is the individual filter module (a plastic tube with the hollow fibres within). CF Polyolefin fibre LCW Polysulfone fibre bundle. The filter waste is assumed to have a density of 0.5 t/m 3. Filters are anticipated to require replacement every 5 to 10 years. For the purposes of this Application they are assumed to be changed every three years. This equates to annual arisings of: CF Filter: LCW Filter: Total: 4.4 m 3 /year, 2.2 t/year 0.12 m 3 /year, 0.06 t/year 4.52 m 3 /year, 2.26 t/year On removal from the unit, the filters will be loaded into suitable containers (in compliance with the WAC of the receiving facility). These are transferred to the LAWMF, followed by direct disposal to LLWR or where reasonably practicable transported to a UK treatment facility for volume reduction or incineration prior to disposal at LLWR. Volume reduction facilities. Incineration facilities. LLWR. CF Filters: Co-60, Mn-54, Fe-55, Zn-65; 4.1 x 10 3 Bq/g LCW Filters: Co-60, Mn-54, Fe-55, Zn-65; 4.1 x 10 3 Bq/g CF Filters: 60 x 2.2 x 10 6 g x 4.1 x 10 3 Bq/g = 5.41 x 10 2 GBq LCW Filters: 60 x 0.06 x 10 6 g x 4.1 x 10 3 Bq/g = 14.8 GBq Total: 5.56 x 10-1 TBq Total arisings for two units: CF Filters: LCW Filters: Total: 60 x 4.4 m 3 = m 3, 60 x 2.2 t = t 60 x 0.12 m 3 = 7.2 m 3, 60 x 0.06 t = 3.6 t = m 3, = t Page 125 of 255

140 Table 5.32 Radioactive Waste Arisings Heterogeneous LLW Parameter Waste Stream Waste Category Waste Origin Waste Characteristics Annual Arisings Proposed Packaging Anticipated Disposal Route Key Radionuclides Total Activity Information This waste consists of a diverse range of waste materials that are anticipated to arise throughout the normal operations phase as a result of operations and maintenance activities. The wastes will include metals, hard wastes, soft wastes, inert wastes and organic wastes (including cellulosics). Waste materials are expected to include plastics, paper, card, wood, glass, building materials, insulation, motors, cables and pipes, miscellaneous filters and strainers. Dry LLW/VLLW Heterogeneous LLW may be generated anywhere within the active facilities including R/B, T/B, Rw/B, S/B and radioactive waste management facilities. The waste characteristics will be diverse and will include a range of sizes, shapes, masses and densities. It is assumed that wastes will be generated in quantities/sizes that are manageable using standard range of LLW containers used in the UK. Since this waste includes a significant proportion of combustible and soft wastes a mean density of 0.5 t/m 3 is assumed. Estimated m 3 /year for two units based on the following estimated breakdown: Combustible LLW: 62 m 3 Non-combustible LLW: 15.4 m 3 Combustible VLLW: 28 m 3 Non-combustible VLLW: 6.8 m 3 Heterogeneous LLW will be managed in compliance with waste hierarchy application and disposed in line with BAT, which will be case specific. Waste packaging will be in compliance with the WAC of the receiving facility. Should the waste be sent to LLWR, containers specified by LLWR [RD30] will be used. Metals will be segregated for metal melting treatment where reasonably practicable. Metal melting facilities. Volume reduction facilities. Incineration facilities. Permitted landfills. LLWR. Fe-55, Co-60, Mn-54; 4.1 x 10 3 Bq/g m 3 /yr x 0.5 t/m 3 x 60 yr x 10 6 g/t x 4.1 x 10 3 Bq/g = 1.38 x 10 1 TBq Page 126 of 255

141 Parameter Information Total Arisings Combustible LLW: 2 x 60 x 31 m 3 = m 3, m 3 x 0.5 t/m 3 = t Non-combustible LLW: 2 x 60 x 7.7 m 3 = m m 3 x 0.5 t/m 3 = t Combustible VLLW: 2 x 60 x 14 m 3 = 1,680.0 m 3 1,680.0 m 3 x 0.5 t/m 3 = t Non-combustible VLLW: 2 x 60 x 3.4 m 3 = m m 3 x 0.5 t/m 3 = t Total: 2 x 60 x 56.1 m 3 = 6,732.0 m m 3 x 0.5 t/m 3 = 3,366.0 t Note this figure is for raw waste arisings and does not take account of any waste minimisation, size reduction or other treatment/processing activities. Further definition of the design of the Dry LLW processes is required in order to refine final waste quantities and volume reduction factors for discrete processes. Page 127 of 255

142 Table 5.33 Parameter Waste Stream Waste Category Waste Origin Waste Characteristics Annual Arisings Proposed Packaging Anticipated Disposal Route Key Radionuclides Total Activity Total Arisings Radioactive Waste Arisings Radioactive Oils and Oily Wastes Information This waste stream is included to cover any arisings or radioactively contaminated oils or oily wastes and to demonstrate a holistic consideration of radioactive wastes that have the potential to arise from the Power Station. Predicted volumes over the life of the Power Station are currently assumed to be zero. Assumed to be limited to LLW/VLLW levels. No production of radioactively contaminated oils or oily wastes is anticipated from the Power Station. However, it is possible that during the operational life of the plant accidental spillages or leakages of oil could become radioactively contaminated. For the purposes of this Application it is assumed that any radioactive oil would be contaminated (not activated) and of low volume. Sources of oils are predominantly lubricating oils associated with the main turbine and generating sets, and fuel oils associated with standby generating plant. It is recognised that H-3 contamination of turbine lubricating oil could occur but at present this is neither established nor quantified and further assessment is required. Lubricating oils, fuel oils, oil contaminated rags/wipes, oil soaked absorbent materials. Assumed to be zero. Waste packaging would be in compliance with the WAC of the receiving facility. Permitted incineration facilities. Incineration residues will be sent to LLWR by the incineration service provider. Permitted landfill for VLLW. LLWR. Determined on a case by case basis. Assumed to be zero. Assumed to be zero. Page 128 of 255

143 Table 5.34 Parameter Waste Stream Waste Category Waste Origin Waste Characteristics Annual Arisings Proposed Packaging Anticipated Disposal Route Key Radionuclides Total Activity Total Arisings Radioactive Waste Arisings Radiologically Contaminated Land Information This waste stream is included to cover any arisings or radiologically contaminated soil or other ground materials and to demonstrate a holistic consideration of radioactive wastes that have the potential to arise from the Power Station. Predicted volumes over the life of the Power Station are currently assumed to be zero. Assumed to be limited to LLW/VLLW levels. Examples in which radiologically contaminated soils may arise as a waste product include: if there are areas of contamination resulting from the historic use of the site or the adjacent Magnox power station; and, as a result of spills or other accidental releases during the normal operations phase. Radiologically contaminated soil/gravel/rock due to localised spills of radioactive material or identification of historic contamination/accumulation. Assumed to be zero. Waste packaging would be in compliance with the WAC of the receiving facility. Permitted landfill for VLLW. LLWR. Determined on a case by case basis. Assumed to be zero. Assumed to be zero. Page 129 of 255

144 5.8.3 Estimate of Solid Radioactive Waste Arisings from Decommissioning Table 5.35 Decommissioning Radioactive Waste Summary Parameter Waste Stream Waste Category Waste Origin Waste Characteristics Annual Arisings Proposed Packaging Anticipated Disposal Route Key Radionuclides Total Activity Information This waste stream is included to cover any arisings from activities associated with the decommisoning of the Power Station. Wastes being discharged from site during the decommissioning and subsequent quiescent phase will include packaged spent nuclear fuel (as HLW), ILW packages from the operational phase plus decommissioning, and quantities of LLW/VLLW arising as a result of the decommissioning activities. Following the End of Electricity Generation (EoG) at both reactors spent fuel will remain in the fuel cooling pools for up to 10 years. During this period no major decommissioning work will be carried out in many of the main plant buildings as these house systems which support the operation of the SFPs. As many of the aqueous and wet-solid waste processing systems are expected to remain operational during this period, it is assumed that waste continues to be generated at operational levels as long as spent fuel remains in the SFPs. This is likely to be a conservative assumption as, without an operational reactor, the main source of radioactivity is removed. During this period, redundant systems will be de-energised, drained of operating fluids and cleaned out ready for dismantling. Following the removal of spent fuel from the SFP at each reactor unit, that unit s remaining systems will undergo a thorough post-operational cleanout and decontamination, likely to include aggressive chemical decontamination of the RPV and circulating systems. After decontamination, systems will be dismantled with items of removed plant/equipment being transferred to a new decommissioning waste management facility (DWMF) for size reduction, processing and packaging for disposal. The exception to this is expected to be ILW items which will follow a handle once principle and be placed directly into their disposal packages at their place of arising before the package is taken to the DWMF to be grout-filled before emplacement in a newly constructed Decommissioning ILW Storage Facility. Wastes arising from decommissioning of the UK ABWRs at the Power Station site will consist of a large quantity of metal and concrete wastes from dismantling together with arisings of materials associated with decommissioning processes including decontamination resins and slurries from diamond wire cutting of concrete structures and abrasive water jet cutting. Since these activities are all associated with radioactive areas there will be quantities of typical operational soft wastes such as PPE, plastics, swabs and wipes etc. Deplanting of systems will generate significant quantities of lower activity metal wastes (pipework, vessels, pumps etc.). Decommissioning is assumed to take place over a 20 to 25 year period immediately following end of generation. No annualised breakdown is provided at this stage. Wastes will be packaged into approved disposal containers prevalent at the time of decommissioning. For spent fuel there is limited information available on the design of a suitable disposal container. It is assumed that currently available routes for disposal of ILW (GDF), LLW (LLWR) and VLLW (permitted landfill sites) will be utilised. Spent fuel is assumed to be disposed of to the GDF in line with UK government policy. Fe-55, Co-60, H-3, Mn-54, Zn-65 Not calculated at this stage. Page 130 of 255

145 Parameter Total Arisings Information Indicative quantities are derived from work carried out to develop the DWMP and reported as a summary. The summary indicates the following quantities; Primary Decommissioning LLW/VLLW LLW/VLLW from deplanting operations, Post Operational Clean Out (POCO), scabbling etc. 17,953 te Secondary Decommissioning LLW/VLLW LLW/VLLW Filters, Used Equipment, Personal Protective Equipment (PPE), Decontamination Resins, used Multipurpose Container (MPC) packages etc. 8,578 te Primary Decommissioning ILW ILW Activated materials, Segmentation of RPV and Reactor Internals (RIN), Control Rods etc. 4,005 m 3 (Packaged Volume) Secondary Decommissioning ILW ILW Decontamination Resins, Cutting Abrasives etc. 237 m 3 (Packaged Volume) Page 131 of 255

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147 6 Monitoring 352. This section provides information about the sampling and monitoring arrangements (including in-process sampling and monitoring) that Horizon proposes to put in place at the Power Station to enable the measurement and assessment of discharges and disposals of radioactive waste. It also presents a description of the proposed environmental monitoring programme that will be deployed during the operational phase of the Power Station. It also gives an overview of the monitoring that has been undertaken to date and the additional monitoring that will be carried out in order to establish a pre-operational baseline at the surrounding enevironment to the Power Station. The section therefore addresses the requirements of parts 5a and 5b of Natural Resources Wales (NRW) application form RSR-B3 [RD16] It is highlighted that, at this stage in the development of the Power Station, decisions have yet to be made regarding the exact specification of the sampling and monitoring equipment, and techniques, which will ultimately be employed. The arrangements described below are therefore reflective of the methods and techniques that are currently available to achieve the required detection limits for the radionuclides which must be monitored, as identified in Section 5. These methods and techniques currently represent Best Available Techniques (BAT), as is demonstrated in Section 4. Their further development will be implemented under Horizon s Forward Work Plan (Section 9), and the development will be documented within the Wylfa Newydd EP-RSR BAT Case (Section 4) [FA RSR-3] The proposals presented below cover routine monitoring only and do not address incident or emergency monitoring 16. This is in line with the requirements of EP-RSR, which covers discharges resulting from routine and reasonably foreseeable events. 6.1 Objectives, Standards and Guidance Objectives of the Sampling and Monitoring Systems 355. The Power Station s sampling arrangements and radiation monitoring systems are designed to: Meet the as low as reasonably achievable (ALARA) principle through BAT, by delivery of optimal plant performance; Provide assurance that the total activity discharged to the environment is lower than the permitted limit, and that the conditions of the EP-RSR are being met; Provide trend data on plant performance and detect abnormal plant operation; Provide robust data to facilitate assessment of the radiological impacts to the public and the environment; and, Ensure maintenance of optimal plant performance through in-process monitoring, thereby minimising radioactive discharges to the environment. 16 Monitoring of incidents and emergency scenarios is covered by the Nuclear Site Licence (NSL) (see Section 1.5.4). Page 133 of 255

148 6.1.2 Standards and Guidance 356. The following standards and guidance are relevant to the development of the Power Station s sampling and monitoring systems. They have been considered throughout its design, and will continue to be used for future design work: EU 2004; BS ISO 2889:2010; ISO 10780:1994; BS EN :2004 and BS EN :2004; MCERTS; BS EN ISO/IEC 17025:2005 and ISO 11929:2010; Radioactive Substances Regulation Environmental Principles; and, Environment Agency (EA)/NRW Technical Guidance Notes. The above standards and guidance documents are each described briefly below Horizon will ensure that the most up-to-date standards and guidance are employed throughout the design process, thus ensuring that techniques and instrumentation will represent BAT EU European Union (EU) 2004 [RD32] [RD33] provides recommendations for reporting on radioactive gaseous and liquid discharges into the environment from nuclear power reactors and reprocessing plants under normal operation. The recommendations address the identification of key radionuclides and the setting of discharge detection limits. This information has been used to determine which radionuclides should be monitored at the Power Station BS ISO 2889: BS ISO 2889:2010 [RD34] is an international standard which has been adopted as a British Standard. It contains sets of criteria and recommendations for sample extraction, sample system design, sample transport, performance criteria and quality control for gaseous discharges. It also contains annexes which provide some options for the collection and analysis of selected analytes. This standard will be used to ensure BAT is being applied to the gaseous sampling and monitoring design of the Power Station ISO 10780: This standard [RD35] has been used to determine the location of the sampling point as well as the flow measurement point within the R/B stack, ensuring that the sample collected is representative of the discharge as a whole. Page 134 of 255

149 BS EN :2004 and BS EN : The BS EN 60761:2004 series are international standards which have been adopted as a British Standard based on IEC 60761:2002. BS EN focuses on equipment for the continuous monitoring of activity in gaseous effluents. Part 1 [RD36] addresses the general requirements of continuous monitoring, of which there is some overlap with ISO The remainder of the series within this standard goes into detail for specific analytes. The continuous monitoring of noble gases is covered in Part 3 [RD37] MCERTS 362. MCERTS is the Environment Agency s Monitoring Certification Scheme. Its purpose is to promote the production of quality monitoring data and provide the key foundation of an operator s self-monitoring policy. With respect to radioactive discharges, only the analysis of liquid effluent is currently covered by MCERTS [RD38]. There is a requirement for flow measurements (for both gaseous and liquid discharges) to be undertaken to MCERTS standard [RD39][RD40] as these data are required to enable accurate accounting of the discharges as they are released to the environment. Horizon has made a commitment within its Nuclear Security Safety and Environmental Principles (NSSEPs) to adhere to MCERTS requirements and to use appropriately accredited equipment available at the time of procurement. Section provides further information on Horizon s NSSEPs BS EN ISO/IEC 17025:2005 and ISO 11929: BS EN ISO/IEC 17025:2005 [RD41] is an international standard which has been adopted as a British standard. It provides the general requirements for the competence of testing and/or calibrations, including sampling. It comprises five elements, namely Scope, Normative References, Terms and Definitions, Management Requirements and Technical Requirements. This information will be used to ensure the sampling and measurement techniques used are suitable for the Power Station and assist in the application of BAT ISO11929 covers the areas of determination of the detection limits and decision threshold for an ionising radiation measurement. This information will be used by Horizon to ensure that any relevant analysis undertaken from its Environmental Survey Laboratory (ESL), or at an off site laboratory, will be compliant with ISO11929 [RD42] Radioactive Substances Regulation Environmental Principles 365. Sampling and monitoring arrangements will be consistent with industry good practice and will take into account the relevant Radioactive Substances Regulation Environmental Principles (REPs) [RD43]. The REPs considered most relevant to sampling and monitoring, as far as is covered in the scope of this Application, are: RSMDP9 Characterisation; RSMDP13 Monitoring and Assessment; RSMDP14 Record Keeping; ENDP4 Environment Protection Functions and Measures; ENDP10 Quantification of Discharges; ENDP14 Control and Instrumentation - Environment Protection Systems; Page 135 of 255

150 ENDP15 Mechanical Containment Systems for Liquids and Gases; ENDP16 Ventilation Systems; and, SEDP3 Ambient Radioactivity Horizon has developed its own NSSEPs which set down expectations for nuclear safety and both radiological and non-radiological environmental protection. They have been developed using information from a number of sources, including the REPs, and are specifically tailored to Horizon s requirements. A number of principles covered within the NSSEPs relate to sampling and monitoring arrangements. This includes but is not limited to: BP5.3 Environmental Protection Measures; BP10.8 Management and Storage of Waste Pending Ultimate Disposal; and, BP10.7 Environmental Monitoring Programmes Technical Guidance Notes 367. The EA and NRW have produced a number of Technical Guidance Notes (TGN) that are relevant to sampling and monitoring. The main ones pertinent to the sampling and monitoring strategy are M1 [RD44], M11 [RD45], M12 [RD46] and M3 [RD47]. M1 focuses on generic stack monitoring and, in conjunction with ISO 10780, will provide the appropriate guidance for the location of the sampling point within the R/B stack. M11 focuses on the specific requirements for monitoring and sampling gaseous emissions from a nuclear facility. M12 relates to the monitoring of radioactive releases to water from nuclear facilities and will be the main guide for the Power Station s aqueous discharges The Joint Agencies Radiological Monitoring Technical Guidance Note M3 [RD47] applies to all UK nuclear sites with discharges to atmosphere. It identifies objectives to assist operators in showing their compliance with the EP-RSR conditions, including assessing compliance with numerical discharge limits, measuring plant performance and detecting abnormal discharges. It also gives a detailed description of good practice for monitoring systems. 6.2 In-process Monitoring Gaseous Radioactive Effluent In-process Sampling and Monitoring 369. In-process monitoring will be undertaken in the Off Gas (OG) System, the Heating Ventilating and Air Conditioning (HVAC) System and the Turbine Gland Steam (TGS) System to ensure that they are performing as expected Off-Gas System 370. Monitoring will be undertaken at the inlet and outlet of the OG Charcoal Adsorber (see Figure ). At the inlet, a radiation detector will continuously measure the gross gamma radiation level of the pre-treated off-gas (a measure of the amount of the noble gas 17 Figure 6.1 is reflective of arrangements at Unit 1. Unit 2 differs from Unit 1 in not having a HVAC feed from the Radwaste Building (Rw/B). Page 136 of 255

151 transported from the reactor). The system will be linked to an alarm which will be triggered if the measurement is outside a predetermined range 18. The gross gamma radiation level of the treated off-gas will be continuously measured by a second radiation detector located at the outlet of the adsorber. The measured radiation level will be indicated and recorded in the Main Control Room (MCR). The detection system will also be linked to an alarm which will be triggered if the measurement is outside a predetermined range. Horizon will define inlet and outlet alarm trigger points as part of the development of its operational arrangements, along with the appropriate operator response procedures [FA RSR-4]. Figure 6.1 Overview of Monitoring Locations for Gaseous Effluent 371. A grab sampler will be located at the inlet and outlet of the OG Charcoal Adsorber. The samplers will enable advanced laboratory-based analysis of the radionuclide composition of the pre-treated and treated gas. At an appropriate time in the design process Horizon will confirm the techniques which will be used for the grab sampling [FA RSR-3]. The regime for sampling will be confirmed in Horizon s operational arrangements A flow meter is provided at the inlet of the adsorber. The measured flow rate is displayed and recorded in the MCR The performance of the charcoal adsorbers is influenced by temperature. For this reason the temperature will be continuously measured to ensure it is within set parameters. 18 The inlet monitoring can be used for the determination of fuel pin failure, as is explained in Appendix D. Page 137 of 255

152 HVAC System 374. In situ differential pressure monitoring will be carried out across the high efficiency particulate air (HEPA) filters in the HVAC system. The differential pressure is an indicator of filter performance and will inform decisions on when to replace them. Horizon will define the locations of the differential pressure monitors in the HVAC system, and the parameters within which the HEPA filters should operate at an appropriate time in the design process [FA RSR-4] The HVAC systems servicing controlled areas are currently not subject to in-process radiation monitoring. Horizon will review this design aspect at a later stage of the design process [FA RSR-4] The flow rate of the HVAC is continuously measured to monitor the performance of the individual systems Turbine Gland Steam System Off-gas and Mechanical Vacuum Pump Exhaust 377. The TGS system and Mechanical Vacuum Pump (MVP) will both discharge gas to the R/B stack, the former during normal operation and the latter during start-up. Downstream of the junction where the two discharge lines converge, and before they feed into the main stack, a gamma radiation detector will be provided (see Figure 6.1). The detector will continuously measure the gross radiation levels of the TGS off-gas and MVP exhaust with the measured radiation levels being displayed and recorded in the MCR. The detection system will be linked to an alarm which will be triggered if the measurement is outside a predetermined range A flow meter and grab sampler will be provided in the same location to measure the gas flow rate and enable laboratory-based analysis of the radionuclide composition of the gas. The flow rate will be indicated and recorded in the MCR. As is the case for the OG System, detection system alarm trigger point(s) and grab sampling techniques will all be defined at a later stage of the design process [FA RSR-3]. The frequency of sampling and the radionuclides to be analysed will be confirmed in Horizon s operational arrangements Liquid Radioactive Effluent In-process Sampling and Monitoring 379. The components of the Low Chemical Impurities Waste (LCW), High Chemical Impurities Waste (HCW) and Controlled Area Drain (CAD) systems that require in-process monitoring are listed in Table 6.1 along with detail on the monitoring methods that will be applied. The components and methods listed in the table will be developed as part of the detailed design process [FA RSR-4]. Page 138 of 255

153 Table 6.1 LCW, HCW and CAD Sampling and Monitoring Locations Process System Monitor Location of Monitoring Objective Manual sampling Manual sampling and analysis at LCW collection tank. Analysing suspended solids and conductivity to confirm that the properties of collected liquid effulent are suitable for the LCW. LCW Conductivity meter Inlet and outlet of LCW demineraliser. Monitoring conductivity to confirm the processing performance of the demineraliser. Manual sampling Manual sampling and analysis at LCW sample tank. Analysing to confirm that the properties of processed water satisfy the reuse criteria: conductivity, ph, chloride ion (Cl - ), sulphate ion (SO 4 2- ) and total organic carbon (TOC). Conductivity meter Inlet and outlet of HCW demineraliser. Monitoring conductivity to confirm the processing performance of the demineraliser. HCW Manual sampling Manual sampling and analysis at HCW sample tank. Analysing to confirm that the properties of processed water satisfy the reuse criteria (conductivity, ph, chloride ion (Cl - ), sulphate ion (SO 4 2- ) and TOC) or other discharge criteria. CAD Manual sampling Manual sampling and analysis at CAD collection tank. Analysing to confirm that the properties of collected liquid effulent (largely expected to be non-radiological properties) satisfy the discharge criteria HCW Sampling and Monitoring 380. The (HCW) system includes a number of sampling and monitoring points to confirm the performance of the system and to ensure the criteria for reuse are met (see Section for further details). The locations of these monitoring points are shown in brackets in Figure 6.2, while the purpose of the sampling points is described in Table 6.2. Figure 6.2 Outline of HCW and Monitoring/Sampling Points Page 139 of 255

154 Table 6.2 Items and Purpose for Each Monitoring and Sampling Point HCW System Monitoring/ Sampling No Measured Items Purpose of Measurement (1) Conductivity, chloride ion, suspended solid, radioactivity, ph (2) Suspended solid Chloride ion, etc. (Manual sampling and analysis) To confirm the properties of liquid effluent supplied to the evaporator. (Manual sampling and analysis) To manage the concentration of concentrated liquid waste in the evaporator, these items are measured periodically. (3) Conductivity Monitoring of the carryover from the evaporator to confirm whether supplied liquid effulent (distilled water) is suitable for the demineralisation process. (4) Conductivity (inlet/outlet) To confirm the processing performance of the demineraliser. (5) Reuse criteria or discharge criteria (Manual sampling and analysis) To confirm that the properties of processed water satisfy reuse criteria (or discharge criteria). (6) Radiation During the discharge, radiation levels are continuously monitored and if the monitor signal exceeds the pre-set value, discharge automatically stops. Before the discharge, a manual sampling and activity check is performed at the sampling tank. 6.3 Discharge Monitoring 381. Monitoring will be undertaken to ensure that an accurate record is kept of gaseous and liquid effluent discharges to the environment and to demonstrate compliance with the conditions and limits on radionuclides set in the EP-RSR. The sampling locations are downstream of any abatement systems: gaseous effluent will be monitored in the R/B stacks and aqueous effluent in collection tanks prior to its discharge via the cooling water outfall. Further information on the discharge monitoring is given below Gaseous Discharge Monitoring R/B Stack 382. Each of the two R/B stacks has an inner diameter of 3.15 m and discharges to atmosphere at a height of 75 m above ground level. Two input lines are connected at the bottom of the stack, one from the HVAC system and the other from the OG System (prior to entering the stack, the TGS and OG systems combine). Inside the stack, pipework for the emergency gaseous discharge system the Standby Gas Treatment System (SGTS) and the sampling pipework are installed. This arrangement is illustrated in Figure 6.3. The SGTS is only used in accident scenarios (i.e. not invoked for any routine or reasonably foreseeable events) and is therefore not discussed further. Page 140 of 255

155 Figure 6.3 Sketch of the Reactor Stack Sample Collection 383. A sample will be collected at a location within the stack that is at approximately 10 hydraulic diameters (10D) downstream from the last input, namely the upper edge of the HVAC duct, whilst being approximately three hydraulic diameters (3D) upstream from the end of the stack. This arrangement will ensure that the air within the stack is well mixed and therefore any sample collected will be representative of the final discharge. This is in line with the earlier described guidance and standards. The flow will also be measured in the same plane (see below) Samples will be isokinetically collected from the stack using a probe which will be consistent with the requirements within ISO 2889 [RD34], ensuring that there will be no preferential fractionation of the particles within the sample relative to the main emissions. This is standard practice across all industries that have a requirement to sample particulate in gaseous discharges. During the Commissioning Phase the particle concentration profile will be determined (as described within [RD34]) to show that the coefficient of variance is less than 20% across the centre two thirds of the stack. This will assist in determining the number and location of sampling points required. Multi-nozzle probes may be used as applicable The sampling location will be provided with a platform designed to comply with M1 [RD44]. This will give workers safe access for periodic inspection and maintenance. Under M1, the minimum requirement for a stack of the proposed size is a one sided platform with a Page 141 of 255

156 working space of about 4.6 m. However, a full circumference platform with a working area of about 3.1 m is also being considered. The final choice will be dependent on the supplier of the probe and will be made by Horizon at an appropriate time in the design process [FA RSR-3] The collected sample will be transported via a single line to the sampling equipment located within the Sampling Equipment Room. The route will be designed to contain the minimum number of bends, and the curvature ratio of all bends will be a minimum of four to reduce losses (ISO 2889 requires three) [RD34]. The sampling line will be made of stainless steel with a smooth inner wall. This will minimise internal deposition and ensure the longevity of the line. In order to avoid vapour condensation the sample temperature will be kept equal to or above that of the stack flow. Almost all sections of the pipework will be located within the inside of the stack to ensure the sample temperature is equal to the stack flow temperature. Where the pipe is located outside of the stack its temperature will be maintained above the dew point The Sampling Equipment Room will be placed at the nearest possible location to the R/B stack. The location will assist in the sample s representativeness as the sampling pipe length and bends will be minimised, whilst also satisfying other key requirements such as radiation protection. Sample Analysis 388. The design of the sampling system is illustrated in Figure 6.4. It comprises two duplicate trains for the sampling of: Particulate material; Iodine; Noble gases; Tritium; and, Carbon-14. The duplication provides redundancy and will enable one of the systems to be accessed by regulators for the purpose of independent monitoring. During normal operation both sampling trains will be running at all times. However, as part of normal operations it will also be possible for Horizon to isolate one sampling train to conduct maintenance, testing and survillence activities (where appropriate) whilst the second system remains in operation The radionuclides will be collected in an order that ensures the best sample is obtained: Particulates will be collected first to minimise losses through plating out (e.g. loss of particulate material to the sampling infrastructure, such as pipework); Once particulates have been removed the sample will be passed through an iodine adsorber which will typically provide only a short delay; The sample will then be passed into the gas chamber for noble gas analysis. Leaving this analysis until last will remove a lot of other gamma emitters, ensuring greater clarity within the gamma spectrum of the noble gas level. This arrangement complies with BS EN [RD37]; and, Tritium and carbon-14 will be collected on a different line as shown in Figure 6.4. Page 142 of 255

157 Figure 6.4 Stack Sampling System Configuration 390. Particulate material will be collected on filter media. Chemical forms of iodine will be collected using a solid adsorbent material. This is expected to be in the form of a charcoal filter arrangement. Horizon will determine the choice of adsorbent material at an appropriate point in the design process. Following sample collection, samples will be sent for laboratory analysis. As part of the development of its operational arrangements, Horizon will confirm the laboratory analysis to be undertaken including the methods, radionuclides to be analysed, and the required standards that the laboratory should meet [FA RSR-3] After the removal of particulate material and iodine, noble gases will be continuously monitored by the use of a fixed volume calibrated chamber and appropriate detector system(s). The radiation detector assembly will consist of a shielded gas chamber that houses a gamma detector. It is anticipated that the detector will be an NaI (Tl) scintillator. The exact nature of the detector will be determined at a later date [FA RSR-3]. The gas chamber size and geometry will be driven by the detector choice, and these will be optimised to ensure the appropriate detection limits can be achieved. The gas chamber will be purged with ambient air when a background level measurement is made. Page 143 of 255

158 392. A radiation monitoring unit in the MCR will analyse and visually display the measured radiation level in the gas chamber. The system will trigger an alarm if radiation levels exceed predetermined threshold levels. These levels and the associated response procedures will be confirmed by Horizon at an appropriate time in the design process [FA RSR-5] For the determination of H-3 and C-14 it is proposed to use a series of bubblers to collect samples for subsequent analysis. A bubbler comprises traps in series to increase the trapping yield of radioisotopes in sample air. The exact composition and number of the bubblers will be determined by Horizon and BAT will be applied [FA RSR-3] During normal operations, both sampling systems will be running at all times 19. If the sampling period is set at two weeks, then each system will be offset in terms of the sample change, e.g. Side One will sample weeks 1 and 2, and Side Two will sample weeks 2 and 3 and so on. In the unlikely event that both sampling systems stop working, discharges will be calculated based on past data related to the current operational phase. The sample collection time will either be recorded by the instruments, in the case of continuous monitoring, or, in the case of sampling, a data recording system will be developed. This will be an integral part of the data quality arrangements Sample collection allows more material to be accumulated over a longer timeframe, providing a sample for analysis with a higher likelihood of producing a detectable result. Exact timescales for sample collection will depend on the equipment available at the time of purchase, in conjunction with the analytical method employed. Horizon will confirm this at an appropriate time in the design process [FA RSR-3]. Sample Return 396. It is good practice to return any sample downstream of the sample extraction location to prevent either double counting or dilution of the sample [RD45]. However, it is proposed that the return point will be located upstream of the sample extraction point as shown in Figure 6.3. This arrangement can save pipework, which in turn will reduce the amount of potentially radioactively contaminated material that would need to be disposed of at the end of the plant life. It is estimated that this could save over 30 m of pipe The impact of returning the gas upstream of the sampling location has been determined to be negligible. During normal operations, Unit 1 has a flow rate through the R/B stack of 760, 321 m 3 /h, while the maximum sampling flow rate is assumed as 24 m 3 /h. This gives a mixing ratio of approximately 1:30,600 which would have insignificant impacts in terms of both dilution of sample and double counting. It is therefore concluded that it is BAT to have the sample returned upstream of the sampling location as the impact of saving pipe is greater than the impact on the representativeness of the sample. 19 Except during periods of maintenance, testing or surveillance activities, or when independent monitoring is being carried out by the regulators. During these times only one system will be operated. Page 144 of 255

159 Volumetric Flow Rate 398. The flow characteristics of the R/B stack at Unit 1 are summarised in Table 6.3. The main constituent of the flow will be the HVAC exhaust. Therefore the chemical characteristic of the discharged gas can be regarded as air whose temperature and relative humidity will depend on the HVAC conditioning. The total volumetric flow rate will be 760,321 Nm 3 /h (normalised at 0 C and absolute pressure of kpa) with a discharge velocity of approximately 30 m/s. The HVAC will occasionally be stopped during maintenance. However, maintenance will only be carried out on the HVAC system when no radiological discharges are occurring. Table 6.3 Discharge Flow Characteristics of the Unit 1 Reactor Stack System Fluid Filter Flow Rate (m 3 /h) Start-up Power Operation Hot Standby Cold Shutdown Refuelling Outage HVAC Reactor Building Turbine Building Radioactive Waste Building Air HEPA 228,285 Air HEPA 353,143 Air HEPA 175,547 Subtotal Air - 756,975 Off-Gas Air Noble gas Charcoal, HEPA Off-Gas Turbine Gland Steam System (TGS) Mechanical Vacuum Pump Air steam HEPA 0 3,160 3, , ,160 0 Air steam HEPA 0 15, Subtotal ,346 3, , ,346 0 Total Air - 756, , , , , , , ,975 Note: flow rate is normalised to a temperature of 0ºC (273.15K) and ambient pressure of kpa 399. The flow characteristics of the R/B stack at Unit 2 differ from Unit 1 as there is no flow from the Rw/B HVAC. The total volumetric flow rate will be 584,774 Nm 3 /h with a discharge velocity of approximately 23 m/s. Further information on the flow rates are detailed in Table Volumetric flow will be continuously recorded at the sampling point within the R/B stack. The flow meters will generally be situated downstream of any sampling equipment. However, the exact number and locations of the flow meters will be determined during commissioning. It is expected that instruments such as pitot tubes will be used for this purpose. There will also be flow measurements recorded within both the HVAC and OG Page 145 of 255

160 systems (as shown in Figure 6.1). These could be used as supplementary information during any period where the main flow readings might be unavailable To provide back up for the gaseous flow measurement, a secondary identical flow monitoring system will be stored at a location that is not susceptible to damage from a lightning strike that would cause the inline primary system to fail. It is estimated that the time taken to install the spare system would be a maximum of two days. A redundant system will not be permanently installed within the R/B stack. As part of its detailed design process, Horizon will ensure that the monitoring system is weather proofed and able to withstand the effects of airborne sea spray [FA RSR-3] Other Discharge Sources 402. Besides the stacks on each R/B, the Power Station will have additional discharge points (listed in Table 3.8). These are the HVAC exhausts or ventilation exhausts from the following buildings/facilities: Service Building (S/B); Lower Activity Waste Management Facility (LAWMF); ILW Storage Facility; and, Spent Fuel Storage Facility (SFSF) The sampling and monitoring design of these facilities will be developed in Horizon s detailed design process. Monitoring arrangements will also be confirmed at an appropriate time in the design process when more information comes available regarding the precise plant and equipment which will be housed in the facilities [FA RSR-6] Aqueous Discharge Monitoring Flow 404. Samples will be collected from the final discharge line exiting the discharge tank using a flow proportional sampler. At the sample location the flow of the discharge will also be measured. Redundancy for discharge sampling has been provided in the form of duplicate flow measurement apparatus along with a second flow proportional sampler Contamination Sampling 405. Each of the subsystems which discharge treated liquid effluent to the environment will have two sample tanks (see Figure 6.5). Tank contents will be sampled prior to discharge in order to confirm that they meet required discharge criteria with respect to radioactive and chemical contamination (to be defined by Horizon as part of the development of its operational arrangements). Each subsystem will be operated independently and will not discharge simultaneously. Page 146 of 255

161 Figure 6.5 Aqueous Effluent Monitoring Locations 406. Once the volume of a tank has reached a predetermined level, the liquid will be agitated by pumping it through a recirculation line. Following this, an in-process sample will be collected from the line and analysed. If the measured activity meets the criteria the liquid will be discharged. Samples will also be collected from the final discharge line, downstream of the discharge tank, using a flow proportional sampler. This will give an accurate record of what is discharged to the environment Interlocks will be in place to prevent the pump for the recirculation line and the discharge line being operational if the inlet valve to a tank is open. This will greatly decrease the likelihood of the possibility of simultaneous discharge and filling of the tank. The techniques and approach for sampling for aqueous discharges will all be defined at a later stage of the design process and in Horizon s operational arrangements [FA RSR-3] Radiation Detection 408. In addition to the sample collection, a continuous radiation monitor will be provided in the aqueous discharge line. If the radiation level exceeds a predetermined level (to be confirmed by Horizon as part of the development of its operational arrangements), the monitor will activate an alarm and will close an isolation valve to stop the discharge to the environment. Page 147 of 255

162 Instrumentation 409. Specific instrumentation has not been specified at this stage. However, measurements and samples of aqueous discharges will be taken using MCERTS accredited instrumentation. This will give confidence in the information collected from the devices in terms of its accuracy, precision and tolerances All samples will be analysed using ISO [RD41], and MCERTS accredited methods [RD38], [RD39] where available. It is expected that methods and techniques will develop over time and that the most suitable technique will be determined by Horizon, using the application of BAT and in consideration of the MCERTS available services in the analytical market. As part of the development of its operational arrangements, Horizon will confirm the laboratory analysis to be undertaken including the methods, target radionuclides, and the analytical standards required [FA RSR-3]. 6.4 Solid Radioactive Waste Monitoring and Characterisation 411. Solid radioactive waste will be monitored and/or sampled at various stages in order to ensure its traceability can be maintained and to measure the performance of the Solid Radioactive Waste Management system (see for further details). Prior to the final disposal, waste will be monitored or sampled to ensure it meets the waste acceptance criteria (WAC) of the facility to which it will be sent for treatment or disposal Monitoring and sampling will be completed for the purpose of characterisation. Characterisation of radioactive waste involves determining its physical, chemical, biological and radiological properties. It may be carried out as a standalone process or during other processing such as during segregation. By applying robust characterisation requirements from the outset, Horizon will be able to optimise the segregation of wastes and application of the waste hierarchy, thereby reducing wastes to the lowest achievable category and minimising the volumes of waste sent for disposal. Section 4 provides further information on the characterisation that will be undertaken at the Power Station At this stage in the development of the Power Station it is not appropriate to provide details of solid radioactive waste monitoring and sampling arrangements that will be adopted. Technology and methods will continue to develop and therefore BAT will be applied to the sampling and monitoring equipment selection and procurement closer to the time of construction [FA RSR-3]. 6.5 Independent Sampling 414. An independent system will be available for the collection of gaseous and aqueous effluent samples by the regulators. There will be no independent continuous monitoring system provided. However, all data and quality control information will be made available to the regulator or their representative for inspection Gaseous Sampling 415. It is proposed that NRW will use one of the gaseous sampling systems when they wish to conduct independent sampling. The system will be provided with tamperproof seals on filters, cartridges, valves, etc. There is no requirement for NRW to have independent access to the noble gas monitoring system. Page 148 of 255

163 416. Both systems will be run at all times, even when NRW does not require access, and will be maintained and calibrated by the operating personnel. Use of one sampling system by NRW will not impact on the use of the other by Horizon. The total sampling flow rate will be maintained constant in order to keep the isokinetic sampling condition when sampling lines are isolated In addition to access to the sampling equipment located within the sampling rooms, a sampling port on the R/B stack will also be provided for independent flow measurement, as close to the sample extraction point as is feasibly possible. The design of the access ports will be consistent with the requirements laid out in M1 [RD44] Aqueous Sampling 418. The Power Station s aqueous discharge sampling design, illustrated in Figure 6.5, has built-in redundancy as there are two flow proportional samplers, both of which will be MCERTS-accredited. During normal operation only one sampler will be in active use. However, both systems will be maintained and calibrated by the operating personnel in line with the manufacturer s requirements. They will also be tested during commissioning to ensure they are both collecting a representative sample When NRW wishes to collect an independent sample, it will be given exclusive access to one of the systems. The samplers can be capped with tamperproof seals for the duration of the sampling period. The system will have the ability to vary the amount of sample collected (NRW can specify this before sampling commences) because their laboratory requirements may be different from those of Horizon. The collected sample will allow for all analytes to be determined. When NRW is using one of the sampling systems, there will be no impact on the operating personnel s ability to collect their samples via the second system. 6.6 Environmental Monitoring 420. Guidance documents relating to EP-RSR applications, and standard conditions of issued EP-RSRs, require operators to undertake site-specific environmental monitoring programmes, the main aims of which are to: Provide public/stakeholder reassurance; Understand how radionuclides behave in the environment and assess long-term trends in their behaviour and distribution; Assess the impact on the most exposed group of individuals (the Representative Persons ) by determining the radioactivity they may be exposed to in the environment in which they spend their time, and in the local products that they consume; and, Detect radioactivity levels in the environment which might require reporting to NRW and further investigation In order to design the environmental monitoring programme and systems for the Power Station, national and international guidance will be taken into account along with information from the current environmental monitoring programme at the Existing Power Station and information on the exposure pathways from the Power Station [FA RSR-8]. Page 149 of 255

164 422. The sections below summarise: Key guidance on the design of environmental monitoring programmes on UK nuclear licenced sites; Pre-construction radiological monitoring that has been undertaken by Horizon which, when considered together with the Existing Power Station monitoring results, provides a pre-construction baseline for the site; The current environmental monitoring programme at the Existing Power Station; and, Monitoring undertaken by the environment agencies and the Food Standards Agency (FSA) around the Existing Power Station.Finally, initial proposals for the future radiological environmental programme at the Power Station are outlined Guidance on the Design of Environmental Monitoring Programmes 423. Guidance has been produced by a number of regulatory bodies and advisory boards to promote consistent, best practice approaches to environmental monitoring. The key guidance, described below, will be taken into account in the design of the environmental monitoring programme and systems for the Power Station Guidance from the environment agencies on EPR16, the Radioactive Substances Regulations (RSR), and radiological monitoring provides a useful framework on which to base decisions. For example, the REPs [RD43] outline how radioactive substances activities will be regulated under the EPR16. Several of the REPs are relevant to the monitoring of radioactive gaseous and aqueous discharges, including RSMDP13, RSMDP14, and ENDP The EA and the Scottish Environment Protection Agency (SEPA) (the Joint Agencies ) have developed a Radiological Monitoring Technical Guidance Note 1 [RD48] which provides information to nuclear operators on best practices for assessing discharges. This guidance is aimed at helping nuclear operators comply with the requirement to use BAT, as specified in the EP-RSR. The guidance is not mandatory, but the Joint Agencies would expect it to be followed (subject to an agreed reasonable timetable) for new facilities The Joint Agencies Radiological Monitoring Technical Guidance Note 2 [RD49] provides guidance on planning and implementing routine environmental radiological monitoring programmes, on their underpinning objectives and principles, and the process for their definition, including stakeholder engagement, where appropriate. The guidance is aimed primarily at designing new, or reviewing existing environmental radiological monitoring programmes around nuclear licensed sites for the purpose of monitoring the environmental effects of authorised discharges. It considers programme design (what to monitor, where and how often) the monitoring and sampling techniques to be employed, and the underlying approach to creating a monitoring programme The EA s Monitoring Technical Guidance Note M5: Routine Measurement of Gamma Ray Air Kerma Rate in the Environment (currently under revision) includes protocols for the measurement, interpretation and reporting of environmental gamma ray air kerma rates. The EA s MCERTS, described earlier in Section , provides for the product certification of monitoring systems, the competency certification of personnel and the accreditation of laboratories and organisations involved in sampling and monitoring There are a number of guidance documents available on how to demonstrate BAT, including guidance on the principles of optimisation in the management and disposal of Page 150 of 255

165 radioactive waste [RD50]. Additional guidance is given in the UK Nuclear industry Code of Practice on BAT [RD51] for the management of the generation and disposal of radioactive wastes, which has been prepared by a Technical Working Group from across the UK nuclear industry Baseline Radiological Monitoring 429. The aim of a baseline radiological monitoring programme is to characterise and establish the existing environmental setting of the development site and its surroundings. The resulting environmental baseline forms a basis against which future environmental monitoring results can be compared, allowing the likely significant effects to be identified and assessed. In line with best practice, the baseline for assessment should represent the conditions that will exist in the absence of the project at the time that the project is likely to be implemented, or when the effects being assessed are likely to arise A pre-construction radiological environment baseline for the site has been established for the purposes of the Development Consent Order (DCO) and Nuclear Site Licence (NSL). Full details of the monitoring undertaken as part of this pre-construction baseline are presented in [RD52] For the purposes of compliance with the EP-RSR, a pre-operational radiological environmental baseline for the site and surroundings of the Power Station will be established prior to the Commissioning Phase [FA RSR-7]. The baseline is likely to be established using data from the existing operator programme, the Joint Agencies monitoring, and additional monitoring undertaken by Horizon Monitoring Undertaken by the Operators of the Existing Power Station Magnox Terrestrial Monitoring Programme 432. The Existing Power Station terrestrial monitoring programme that is undertaken by Magnox Limited (hereafter referred to as Magnox) is designed to monitor radioactivity in agricultural samples and radiation dose rates over land or pasture. Measurements taken and media monitored include: Radiation dose rates over pasture; Radiation dose rates, at the site perimeter fence and at set distances from the station; Passive shade deposition collectors; Grass samples; Soil cores; and, Milk samples. Page 151 of 255

166 Magnox Marine Monitoring Programme 433. The marine monitoring programme carried out by Magnox includes the following measurements and media sampling: Radiation dose rates over shoreline or sediment; Contamination of fishing equipment and the beach; Seaweed samples; Sediment samples; Fish samples; Crustacean samples; and, Mollusc samples Radioactivity in Food and the Environment (RIFE) 434. Routine monitoring of the local environment around the Existing Power Station has been carried out for many years by a number of bodies, for example the current and former operators of the Existing Power Station and NRW. Consequently a substantial body of data on radioactivity in various environmental media exists The environment agencies (EA, NRW, the Northern Ireland Environment Agency (NIEA), SEPA and the food standards agencies (FSA, Food Standards Scotland (FSS)) together publish an annual summary of the independent radioactivity monitoring programmes in the UK, known as Radioactivity in Food and the Environment (RIFE). This includes data from around the Existing Power Station. Horizon has also been undertaking an environmental measurement programme with the intention of supplementing the existing baseline data The primary purpose of the routine monitoring programmes is to check on levels of radioactivity in food and the environment. Nuclear sites are the prime focus of the programme as they are responsible for the largest individual discharges of radioactive waste. Sampling and direct monitoring is conducted close to each of the sites. The monitoring programmes conducted by the agencies are independent of, and also used as a check on, site operators programmes. The results are used to demonstrate that radioactivity in food is well within anticipated levels and that exposure to members of the public from authorised discharges and direct radiation around nuclear sites in the UK remains within legal limits The frequency and type of measurement and the materials sampled vary between sites and are chosen to be representative of existing exposure pathways. Knowledge of such pathways is gained from surveys of local people s diet and way of life. As a result, the programme varies from site to site and may also vary from year to year. The routine programme is supplemented by additional monitoring when necessary, for example, in response to incidents or reports of unusual or high discharges of radioactivity with the potential to get into the food chain or the environment In the report, an assessment is made of doses to the public near nuclear licensed sites using the results of monitoring of radioactivity in food and the environment, supplemented by modelling where appropriate. The assessments use radionuclide concentrations, dose rates and information on the habits of people living near the sites. Changes in the doses received by people can occur from year to year and are mostly caused by variations in radionuclide concentrations and external dose rates. However, in some years doses are Page 152 of 255

167 affected by changes in people s habits, in particular the foods they eat, which is reported in habits surveys. The dose quantity presented in the report summary is known as the total dose and is made up of contributions from all sources of radioactivity from man-made processes Environmental Monitoring for the Power Station 439. This section describes how the environmental monitoring programme for the Power Station will be defined. The development of the programme will be informed by the results of the radiological effects modelling of exposure pathways (see Section 7) along with the results from the monitoring programme for the Existing Power Station and the results presented in the RIFE report. Options to coordinate with and complement the Existing Power Station monitoring programme will be explored with Magnox. The principles outlined in the Joint Agencies Radiological Monitoring Technical Guidance Note 2 [RD49] will be utilised and the programme will be optimised in line with legislation and best practice guidance in place at the time to ensure that BAT is applied [FA RSR-8] Throughout the design of the monitoring programme Horizon will engage with NRW to ensure that the programme meets their expectations. The programme will be subjected to a periodic review and improvement cycle, the timescales and logistics of which will be defined at a later date. Periodic reviews will capture any changes required to the programme as a result of radiological impact (i.e. dose) assessments and Horizon s understanding of the distribution of radioactivity in the environment The programme development will be carried out in the following stages: Collate information and assess site impact; Establish the objectives of the environmental monitoring programme; Determine what to monitor, where, how and how often; Determine analysis and reporting requirements; and, Communicate results and maintain records. These stages are described below Collate Site Information and Assess Site Impact 442. The Power Station s environmental monitoring programme will take into account the principal exposure pathways for radioactivity discharged from the site. This will enable an assessment to be made of those members of the public that will be the most highly exposed to radioactivity discharged from the Power Station (the representative group). In order to do this, information is required on: The environment around the site including land use, water body types and properties; Existing monitoring programmes indicating where radionuclides might accumulate; Habit surveys indicating where and how people spend their time and what food they consume; and, Modelling data to enable monitoring to target areas of highest activity concentration. Page 153 of 255

168 443. The level of impact associated with the site will be established to provide focus to the monitoring: the Joint Agencies guidance states that the magnitude of effort in designing and carrying out the monitoring programme should be commensurate with the level of impact. The guidance defines higher impact sites as those where: Dose to the representative group is greater than 0.02 msv/yr; There is the potential for abnormal releases; The environment is complex and difficult to characterise; or, There is high public concern Much of the information required for this first step has been collated for the radiological effects assessment modelling described in Section 7. This includes information on the habits of people around the site which was obtained from reports produced by the Centre for Environment, Fisheries and Aquaculture Science (CEFAS) Initial assessments, presented in Section 7, indicate that the over 90% of the exposure due to radioactive gaseous discharges is due to the consumption of terrestrial foodstuffs, with over 35% of the exposure due to consumption of milk containing C-14 and over 15% to consumption of root vegetables containing C-14. Over 75% of the exposure due to radioactive aqueous discharges is through consumption of sea fish, crustaceans and molluscs with the doses dominated by contributions from Co-60 and H Therefore grass, milk and root vegetables sampling will be undertaken as part of the terrestrial environmental monitoring programme, and sea fish, crustaceans and molluscs, sediment, seaweed and seawater sampling will be undertaken as part of the marine monitoring programme Establish Objectives 447. The Joint Agencies Radiological Monitoring Technical Guidance Note 2 [RD49] outlines the objectives of environmental monitoring programmes around nuclear facilities. Horizon will therefore take these into account when designing its environmental monitoring programme, as shown in Table 6.4. Page 154 of 255

169 Table 6.4 Objectives of the Environmental Monitoring Programme for the Power Station Objective Guidance notes Power Station programme Objective A - Assess total Representative Person dose. Objective B - Assess dose as an operator s performance measure. Objective C - Assess total impact on wildlife (e.g. dose). Objective D - Assess impact on wildlife as an operator s performance measure (e.g. dose). Objective E - Provide public and stakeholder reassurance. Objective F - Check/complementary monitoring. Objective G - Assess background (very far field). Regulators are responsible for ensuring that dose limits from authorised practices are not breached. Assessing dose as a performance measure is only likely to be informative to an operator where the dose assessed is >0.02 msv/y for discharges under EPR16. Objective should be assigned to monitoring of exposure pathways which contribute doses > msv/y to the total dose. Regulators are responsible for ensuring that reference dose values for exposure of wildlife are not exceeded. Assessing the impact on wildlife as a performance measure is only likely to be informative to an operator where the dose rate calculated is >10 μgy/h* (annual average) for discharges at EPR16 limits. Objective should be assigned to monitoring of exposure pathways which contribute dose rates >1 μgy/h (annual average) to the total dose. Regulators to provide reassurance to public. The need for this objective will be dependent upon ongoing and emerging concerns. This objective should generally not be assigned to a particular exposure pathway where the dose is msv/y. Normally, elements of the regulator s programme provide a check on the operator s programme. However, for foodstuffs in England and Wales, the FSA s monitoring programme may be considered to be the main programme and elements of the operator s programme provide a check on the FSA programme. Check monitoring should provide about a 10% check on the main programme, taking account of the number of locations and frequency of monitoring in the main programme. This objective is generally satisfied by national monitoring programmes required by the European Commission. Measurements from environmental monitoring programme will be used to assess total Representative Person dose. Dose from the Power Station may be > 0.02 msv/y (see Section 7). Exposure pathways that are major contributors to dose (> msv/y) are: consumption of milk, green vegetables and root vegetables, inhalation of the plume. Measurements from environmental monitoring programme will be used to assess doses to wildlife. Will be confirmed during the design of the programme [FARSR-8]. Monitoring will focus on exposure pathways that are major contributors to dose (> msv/y). Will be confirmed during the design of the programme [FARSR-8]. Not included. Page 155 of 255

170 Objective Guidance notes Power Station programme Objective H - Assess long term trends (Indicator). Objective I - Comply with international obligations. Objective J - Detect abnormal, fugitive and unauthorised releases (Indicator). Objective K - Understand/monitor behaviour of radionuclides in the environment. The need for this objective depends upon the observed or potential rate of change of environmental concentrations, for example due to variable discharge profile. The objective should be assigned to monitoring parts of the environment which accumulate or integrate radionuclides (e.g. seaweed). Objectives assigned to monitoring programmes or elements of programmes which are required to comply with international agreements. Objective required where there is the potential for abnormal or fugitive releases. Instrumentation to detect abnormal releases on, for example, stacks may provide better early warning/information and preclude the need for this objective. This objective is likely to be assigned to sites with the largest environmental impact to ensure that the main source pathway receptor routes have been identified, the scientific basis of the programme remains acceptable and any constraints on the data are understood. The objective should only be assigned to elements of the monitoring programme where there are detectable activity concentrations (i.e. greater than limit of detection). Will be included as part of Forward Work Plan [FARSR-8]. To be included for reporting to the European Community (EC). Horizon s in-process and discharge monitoring will enable early detection of abnormal releases. Will be included as part of Forward Work Plan [FARSR-8]. *Note the guidance from NRW to assess impact on wildlife as an operator s performance measure (in excess >10 μgy/h), differs from the criterion stated in the NRW guidance on the basis for establishing radionuclide groups for discharge limit setting (radiological impact on non-human species: exceeds >40 μgy/h). See Table 5.18 for further information Determine What to Monitor, Where, How and How Often 448. Guidance on what to monitor or sample, where and how often, to meet different monitoring objectives is provided in the Joint Agencies Radiological Monitoring Technical Guidance Note 2 [RD49]. Indicative numbers of samples of monitoring activities (monitoring locations at different times) are given for high impact sites Horizon will work closely with the regulators and with Magnox to optimise the environmental monitoring programme, taking into account existing monitoring by the regulators and Magnox. It is anticipated that the programme currently undertaken by Magnox will be scaled back during decommissioning of the Existing Power Station. This will be taken into account in the planning of the Horizon monitoring programme The reactor technology in place at the Existing Power Station differs from that to be operated by Horizon at Wylfa Newydd. As a result, the radionuclides discharged to the environment will differ. Horizon will ensure that its monitoring programme will include sampling and analysis of radionuclides discharged to the environment from the Power Station as part of its demonstration of BAT [FARSR-8]. Page 156 of 255

171 451. The frequency or timing of sampling will take into account a number of considerations including: Detecting abnormal releases will require higher frequency monitoring; Monitoring of short-lived radionuclides may require higher frequency monitoring; and, Monitoring may be timed to coincide with particular food growing seasons, seasonal activities or activities of members of the public (e.g. beach occupancy) The geographical locations at which samples and radiation dose rate measurements are taken will, where possible, be evenly located around the Power Station and be at appropriate distances. This will help to ensure that the survey programme provides representative data about the levels of radioactivity in the local area and ensure that locations where higher results might be found are sampled. The locations of the sampling points could also be informed by modelling All sampling locations will be identified by National Grid references to ensure that the sample is always taken from the same place such that representative trends can be compiled as far as practicable. The use of existing sampling locations will be utilised where practicable and appropriate. In addition, Horizon will consider the areas with the potential for maximum deposition to ensure that these areas are targeted as part of its environmental monitoring programme Determine Analysis and Reporting Requirements 454. The environmental monitoring programme will be undertaken in line with arrangements set out in the Horizon Management System (HMS) see Section 8. Sampling and analysis will be carried out in accordance with relevant standards by competent personnel and will be demonstrated as being BAT. An audit trail of all samples will be maintained from the point of collection to final analysis. Samples will be transported to and stored in the laboratory in a secure manner under storage conditions that minimise or eliminate loss or change of the principal constituents under investigation. Samples will be retained to enable future analysis for a period to be agreed with NRW Methods for analysis of samples will meet the requirements of relevant British standards, MCERTs standards or other nationally recognised standards. The analytical methods will be adequately validated and controlled such that they are or could be accredited by the United Kingdom Accreditation Scheme (UKAS) (or equivalent) under BS EN ISO/IEC 17025:2005 [RD41]. Full details of the relevant standards that will be met by the environmental monitoring programme will be set out as part of FARSR Continued assessment of the competence of personnel carrying out the sampling and monitoring will be undertaken by internal audit and formal suitably qualified and experienced personnel reviews where appropriate. Training will be undertaken and documented with evidence of competence Communicate Results and Maintain Records 457. Reports of verified environmental monitoring results will be produced in a timely manner and communicated to the regulators in accordance with the EP-RSR conditions, limtations and requirements. Reports will reference the methods used and the quality assurance processes employed. The results will be presented with information on units, uncertainties and detection limits. It will be clearly stated whether the results are decay corrected to the Page 157 of 255

172 date of sampling. Horizon will also present the annual results of the environmental monitoring that takes place on site and around Anglesey on its website Records will be kept for a defined period of time (in accordance with stakeholder requirements), the information will be traceable and retrievable, taking account of changing storage technology In the event that the abnormal results (i.e. any significant adverse environmental effects that could reasonably be seen to result from the operation of the facility) are identified through its environmental monitoring, Horizon will contact NRW immediately. Page 158 of 255

173 7 Radiological Assessment 461. This section presents a prospective dose assessment for humans and an assessment of the impact on non-human biota of radioactive discharges from the Power Station at the discharge limits specified in Section 5. Impacts (doses) due to direct radiation are also calculated for humans. The dose assessment includes estimates of the following: Annual dose to the most exposed members of the public 20 from planned gaseous, and aqueous radioactive discharges from the Power Station; Potential short-term doses from the maximum anticipated short-term discharges from the Power Station (resulting from fuel pin failure); Annual dose to the most exposed members of the public from direct radiation from the Power Station; Collective dose to the UK, European and World populations; The build-up of radionuclides in the local environment at the end of the Power Station s life, and an assessment of whether this has the potential to prejudice legitimate users or uses of the land or sea; Annual dose to the most exposed members of the public from combined discharges from both the Power Station and the Existing (Magnox) Power Station at Wylfa; Annual dose due to historical discharges and sources of radiation from the Existing Power Station; and, Total dose, i.e. the prospective combined dose from operation at the Power Station and the Existing Power Station, and doses due to historical discharges There is no intention either to carry out incineration at the Power Station site or to dispose of aqueous radioactive waste into the ground. Doses due to these activities are therefore not relevant to the radiological assessment and are not further discussed With regard to the non-human dose assessment, the dispersion and build-up of radionuclides has been estimated in three specific habitat types terrestrial, marine and fresh water for radioactive discharges from the Power Station, the Existing Power Station, and other permitted sites. Dose rates have been estimated for the most exposed organisms and compared with a threshold screening value For both the public and non-human assessments an evaluation has been made of the uncertainty associated with the analysis The information presented in this section meets the requirements of parts 6a and 6b of Natural Resources Wales (NRW) application form RSR-B3 [RD16]. 20 Referred to as the Representative Person which, is an individual receiving a dose that is representative of the more highly exposed individuals in the population [RD54], [RD59]. In order to identify the Representative Person for the Power Station, firstly assessments need to be performed for a number of candidate Representative Persons/groups comprising a suitable age range [RD57]. Page 159 of 255

174 7.1 Human Dose Assessment Guidance 466. The human dose assessments undertaken for the purposes of this Application have been informed by the guidance published in Principles for the Assessment of Prospective Public Doses arising from Authorised Discharges of Radioactive Waste to the Environment [RD57]. The guidance document, which has been prepared by the UK environment agencies, outlines the regulatory framework that currently applies to radioactive discharges into the environment and describes 13 underlying principles: Prospective dose assessment methods, data and results should be transparent and made publicly available; Workers, who are exposed to discharges of radioactive waste, but do not work directly with ionising radiation and are therefore not normally exposed to ionising radiation, should be treated as if they are members of the public for the purpose of determining discharge permits or authorisations; When determining discharge permits or authorisations, the dose to the Representative Person should be assessed; Doses to the most affected age group should be assessed for the purpose of determining discharge permits or authorisations. Assessment of doses to a one year old, 10 year old and adult (and foetus when appropriate) is adequate age group coverage; The dose to the Representative Person which is assessed for comparison with the source constraint and, if appropriate, the site constraint, should include all reasonably foreseeable and relevant future exposure pathways; Significant additional doses to the Representative Person from historical discharges from the source being considered and doses from historical and future discharges and direct radiation from other relevant sources subject to control should be assessed and the total dose compared with the public dose limit of 1000 µsv/y; Where a cautious estimate of the dose to the Representative Person exceeds 20 µsv/y, the assessments should be refined and, where appropriate, more realistic assumptions made. However, sufficient caution should be retained in assessments to provide confidence that actual doses received by the Representative Person will be below the public dose limit; The assessment of dose to the Representative Person should take account of accumulation of radionuclides in the environment from future discharges; The realistic habits adopted for the Representative Person should be those which have actually been observed at the site, within a period of about 5 years. Changes to habits which are reasonably likely to occur should be taken into account; Land use and infrastructure should have sufficient capacity to support the habits of the Representative Person. Any changes to land use and infrastructure should be reasonably likely to occur over a period of about 5 years and be sustainable year on year for them to be considered; The dose assessed for operational short term release at proposed notification levels or limits should be compared with the source constraint (maximum of 300 µsv/y) and the Page 160 of 255

175 public dose limit (1000 µsv/y), taking into account remaining planned discharges during the remainder of the year and contributions from other relevant sources under control; For permitting or authorisation purposes, collective doses to the populations of UK, Europe and the World, truncated at 500 years, should be estimated; and, Where the assessed mean dose to the Representative Person exceeds 20 µsv/y, the uncertainty and variability in the key assumptions used for the dose assessment should be reviewed Dose Limit and Dose Constraints 467. The assessment of prospective human dose 21 has been made with respect to the following criteria referred to in [RD57]: The sum of the doses arising from exposure to an individual member of the public shall not exceed the public dose limit of 1000 µsv a year; The dose received by an individual member of the public from any single site shall not exceed 500 µsv a year (the site constraint ); and, The dose received by an individual member of the public from any new discharge source since 13 th May 2000 shall not exceed 300 µsv a year (the source constraint ) Equivalent dose limits exist for particular organs (skin and the lens of the eye), but assessment of these doses are not required for the purposes of this Application. Specific dose limits for occupational exposure to radiation also exist but these are separately regulated and are therefore not discussed further within this document Public Health England (PHE) has advised the Government (including the Devolved Administrations) that a lower dose constraint not exceeding 150 µsv/y should be applied at the design stage of new nuclear power stations and waste disposal facilities [RD60]. However, it is noted that this advice has not been formally accepted The International Atomic Energy Agency (IAEA) has presented dose criteria which are considered sufficiently low that doses arising from sources or practices that meet these criteria may be exempted from regulatory control [RD61], [RD62]. One of the criteria is that the dose should be less than about 10 μsv/y per practice The statutory guidance to the Environment Agency (EA) states that where the prospective dose to the most exposed group of members of the public resulting from discharges from a site at its current discharge limits is below 10 μsv/y [RD63], [RD99] the EA should not seek to reduce further the discharge limits that are in place, provided that the holder of the permit or authorisation applies and continues to apply Best Available Techniques (BAT) for limiting discharges to the environment and hence reducing the resulting radiological impact. This criterion is applicable to England and Wales and applies to sites that are already operating under an environmental permit. 21 The term dose refers to the calculated sum of the committed effective dose from internal exposure (from radionuclides entering the body through the intake of food and liquids and by inhalation) and external dose from radionuclides remaining outside the body. The quantity is for one year s exposure to external dose and committed dose (over 50 years for adults and to age 70 years for children) from one year s exposure to each radionuclide under consideration. Page 161 of 255

176 472. Taking the internationally accepted assumption (for the purpose of radiation protection) that any dose, no matter how small, has the potential to cause harm, an annual dose of 10 to 20 μsv/y can be broadly equated to an annual risk of death of about one in a million per year. In [RD58] it is stated that 0.01 msv/y and 0.02 msv/y can be considered to be broadly equivalent for the purposes of this principle and so 0.02 msv/y has been retained to ensure consistency of this guidance across the UK and with the approach adopted previously. However, it is worth noting that there is no lower dose level where the need to demonstrate BAT (and specifically the Principle of Optimisation to demonstrate ALARA) can be discounted The EA has stated that discharges giving rise to a per caput (per person) collective dose (the dose calculated for a human population group and different to that of the total effective dose calculated for the Representative Person) of less than a few nanosieverts per year of discharge can be regarded as insignificant [RD57] Assessment of Doses due to Planned Continuous Discharge 474. An assessment was made of doses to representative members of the public to enable identification of the Representative Person in the vicinity of the Power Station resulting from the following radioactive discharges, with all assumed to occur continuously: Gaseous discharges from the Unit 1 and Unit 2 R/B stacks at the annual discharge limits given in Section 5.4; and, Aqueous discharges from the main cooling water outfall, also at the annual discharge limits given in Section Gaseous discharges are anticipated from other building outlets. However, the amount of radioactivity released from each of these sources is anticipated to be very low (see Section 3.3) and they have therefore been discounted from the dose assessment The following annual doses were determined: Doses to the Farming Family (adult, child and infant) due to exposure to gaseous discharges only; Doses to the Fishing Family (adult, child and infant) due to exposure to aqueous discharges only; Doses to the Farming Family (adult, child and infant) due to exposure to both gaseous and aqueous discharges; Doses to the Fishing Family (adult, child and infant) due to exposure to both gaseous and aqueous discharges; Doses to the Magnox worker due to exposure to gaseous discharges; and, Doses to the foetus and breast-fed infants The results of the dose assessment established the most significant radionuclides in terms of radiological impact upon people and the principal exposure route. The results were then combined with those from the assessment of direct radiation doses (Section 7.1.5) to confirm the Representative Person and evaluate compliance with the source dose criteria of 300 µsv a year It is highlighted that this dose assessment, and the others presented later in this section, are very precautionary in nature in order to lend confidence to the conclusions of the Page 162 of 255

177 analysis. The conservative assumptions are identified throughout the text. A discussion of the uncertainty and variability of the public dose assessment is presented in Appendix G Methodology Modelling 479. The gaseous and aqueous discharges from the Power Station are assumed to disperse into the atmosphere and the local marine compartment respectively, and contribute to a radiation dose via ingestion, inhalation and external exposure. The transfer of radionuclides from the Power Station s discharges into the environment and the subsequent dose to the public were assessed using the computer code PC-CREAM 08 (version v /2.0.0) PC-CREAM 08 is a well-documented and tested software package developed for the assessment of continuous radioactive discharges from nuclear facilities in the EU and is the assessment tool used by the environmental regulators in their independent review of RSR permit applications. It enables the assessment of individual and collective dose due to gaseous and aqueous discharges, including internal and external exposure, doses due to deposition and accumulation in the environment, inhalation of re-suspended material including sea spray and transfer of contamination into foodstuffs PC-CREAM 08 comprises the following models: PLUME, the atmospheric dispersion model which predicts the air activity concentrations, deposition rates and external gamma dose rates from radionuclides in the cloud per unit discharge rate; RESUS, which estimates activity concentrations in air arising from the resuspension of previously deposited radionuclides per unit deposition rate; GRANIS, which models the external gamma dose from radionuclides deposited on the soil per unit deposition rate; FARMLAND, which predicts the transfer of radionuclides into terrestrial foods following deposition on the ground. Activity concentrations are calculated for a unit deposition rate; DORIS, which models the dispersion of radioactive discharges to the marine environment and the concentration of radionuclides in seawater, sediment and seafood; and, River models. PC-CREAM 08 also contains two models for calculating radionuclide dispersion within rivers. The first is a simple dilution model that assumes equilibrium conditions and is used for screening purposes. The second is a dynamic model that can be used for studies requiring greater detail. However, as there are no discharges to rivers from the Power Station these models were not used The models calculate explicitly the activity of radionuclides discharged into the environment, and (for most models) then make assumptions with regards to the activity of radioactive progeny. Details of the treatment of progeny are provided in the PC-CREAM 08 documentation [RD64]. PC-CREAM 08 also contains an assessment module, ASSESSOR, which calculates individual and collective doses from discharges to the atmosphere and sea, and individual doses from discharges into rivers. Page 163 of 255

178 Dose Calculation 483. The annual dose to the receptors for atmospheric discharges were assessed using the PLUME, FARMLAND, RESUS, GRANIS and ASSESSOR models of PC-CREAM 08. Once activity concentrations and gamma dose rates in environmental media were calculated, ASSESSOR scaled the derived activity concentrations and dose rate per unit activity by the actual discharge rates, site specific data, habit data and dose coefficients in order to calculate annual effective doses for various exposure pathways The annual doses to the receptors for aqueous discharges were assessed using the DORIS and ASSESSOR modules of the PC-CREAM 08 suite. A local compartment, representing the local marine environment around the nuclear power plant into which the discharge occurs, was defined within DORIS. The model then calculated the transfer of radioactivity around, and its removal from, the marine environment. The activity concentration in various marine environmental media and seafood was calculated for a unit release rate. ASSESSOR was then used to determine the annual effective dose from the marine pathways by scaling for the actual annual discharge rate and appropriate occupancy and habit data. Exposure Pathways 485. Members of the public can be exposed to aqueous and gaseous radioactive discharges via a range of exposure pathways. In the case of ingestion, this is via the consumption of locally produced terrestrial foods and consumption of locally caught seafood. For inhalation, exposure occurs due to inhalation of the plume and also inhalation of sea spray and re-suspended soils. External irradiation occurs when a person is directly exposed to the plume or via ground shine from radionuclides deposited on the ground. It may also occur from the handling of items or exposure to sediments that have associated radioactivity resulting from discharges made to the marine environment The pathways considered for releases of gaseous discharges were: Internal irradiation from inhalation of radionuclides in the plume and radionuclides resuspended following deposition; External irradiation from radionuclides in the plume; External irradiation from radionuclides deposited on the ground; and, Internal irradiation from consumption of contaminated terrestrial foodstuffs grown or reared locally following deposition on the ground (only for radionuclides with a half-life greater than three hours) The pathways considered for releases of aqueous discharges to the marine environment were: External exposure to beach sediments; External irradiation from handling fishing gear; Inhalation of sea-spray when on the coast; and, Consumption of sea fish, crustaceans and molluscs caught locally. Page 164 of 255

179 Identification of the Representative Person 488. As it is not practicable to assess doses to each individual member of the public, the most exposed individual and Representative Person approach was used where: The most exposed individual is the person receiving the highest dose from a single discharge pathway, for example an individual who receives a dose from aqueous discharges only or an individual who receives a dose from gaseous discharges only; and, The Representative Person is an individual receiving a dose that is representative of the more highly exposed individuals in the population due to both aqueous and gaseous discharges and also direct radiation [RD57] The dose to the Representative Person can then be compared with the public dose limit of 1000 µsv/y (future emissions and direct radiation from both the Power Station and Existing Power Station and any dose contribution from other controlled sources plus historical discharges), the source dose constraint of 300 µsv/y (future emissions and direct radiation from the Power Station), and the site dose constraint of 500 µsv/y (for future emissions from both the Power Station and Existing Power Station, but not direct radiation) In order to estimate doses to the public for this and subsequent assessments, candidates for the Representative Person (other than the Magnox worker) were selected with reference to EA guidance [RD57]. This guidance suggests that candidates are chosen following consideration of realistic combinations of habits and a full range of exposure pathways, where the habits survey reports form the basis for such determination The candidates for the Representative Person identified from a review of the three most recent CEFAS habit surveys [RD68], [RD69], [RD70] and used for this dose assessment study were: Farming family (adult, 10-year old child and 1-year old infant); Fishing family (adult, 10-year old child and 1-year old infant); and, Worker at the Existing Power Station (Magnox worker - adult only). Full details of how the habits profiles were constructed are provided in Appendix E The habits surveys did not identify any other potential exposure pathways other than those included within the two candidates for the Representative Person and Magnox worker assessed within this work. The habits surveys did identify activities taking place in or on water. However, these pathways are generally considered to be minor in comparison with other exposure pathways such as the ingestion of locally produced foods and also activities taking place on nearby beaches in the vicinity of the Existing Power Station. In- and/or onwater activities were therefore not considered further within the assessment The Magnox worker is included on the basis of the second of the Principles for the Assessment of Prospective Public Doses, which provides that Workers, who are exposed to discharges of radioactive waste, but do not work directly with ionising radiation and are therefore not normally exposed to ionising radiation, should be treated as if they are members of the public for the purpose of determining discharge permits or authorisations. It is considered that, as the decommissioning of the neighbouring Existing Power Station progresses, workers from non-nuclear sectors (e.g. construction and demolition firms) could be contracted to support the effort and could potentially be exposed to radioactivity from the proposed Power Station. Page 165 of 255

180 Input Data 494. Dose assessments were performed on the basis of proposed annual discharge limits for both aqueous and gaseous discharges from the Power Station, as well as a range of other input data. A summary of this data is presented below and the full details (including the rationale for choice of parameter values) are provided in Appendix E and F. Discharges 495. Data on gaseous radioactive discharges was taken from Table The discharge characteristics of each stack diameter, height, volumetric flow rate are presented in Table 3.7. A conservative effective discharge height of 15 m was used within PC-CREAM 08 to account for the wake effects of nearby buildings on the dispersion of gaseous releases to the atmosphere. The rationale behind this approach is described in Appendix E Data on aqueous discharges was taken from Table Other relevant parameters used in the PC-CREAM 08 marine modelling are presented in Table A.7 in Appendix E. It is highlighted that flow rate and release temperature are not required when modelling discharges to the marine environment using PC-CREAM 08. This is because the discharges are considered to be made into a well-mixed local compartment in which radionuclides will reach equilibrium within the assessment period (typically occurring within a few decades). Other Model Parameters 497. Other data inputs relating to the radiological assessment are presented in Appendix E. These include: Site-specific meteorological data. 10 years of site-specific weather data based on the Numerical Weather Prediction (NWP) model were procured from the UK Meteorological Office. The weather data files were formatted into PC-CREAM 08 compatible files, and used for the atmospheric dispersion model for gaseous discharges; Surface roughness. A surface roughness of 0.3 was used in the radiological assessment for gaseous releases. The factor of 0.3 is representative of agricultural land, which is the dominant land use of the areas surrounding the Power Station site; Ground deposition and resuspension rates modelled by PC-CREAM 08. Default PC-CREAM 08 ground deposition factors and resuspension rates were used in the assessments; and, Parameters defining the marine environment. A variety of parameters were used to specify the marine environment. Some, such as the local compartment coastline length, sediment density and diffusion rate, were PC-CREAM 08 default values. Others, such as the local compartment volume, depth, and volumetric exchange rate, were userspecified based upon values recommended by the EA [RD67] Habits Data and Assumptions 498. A number of assumptions were made in order to ensure that predicted doses are bounding. These are described below. Page 166 of 255

181 Most Exposed Members of the Public for Gaseous Discharges 499. The most exposed members of the public for gaseous discharges were assumed to be a farming family who live at a nearby dwelling and consume 100% locally produced terrestrial food. The dwelling is located at the receptor point which corresponds to the most restrictive residential location (i.e. the location with the highest deposition rate). Adults, children and infants are assumed to spend some time outdoors Adult ingestion rates are based on the top two approach (see Appendix J), with root vegetables and milk being consumed at 97.5 th percentile rates and other foods being consumed at mean rates. Consumption rates for children and infants were derived from the adult values using CEFAS scaling factors in Annex 4 of [RD68] CEFAS habit data extends to a radius of 5 km from the Existing Power Station. Following a review of the data from the last three habits surveys [RD68], [RD69], [RD70], consumption of milk products made from locally produced milk was not identified and so was not considered in this assessment. Grain was also not considered as only one of the three CEFAS surveys identified a farmer who sold barley nationally for human consumption. Grain produced for human consumption is normally mixed with other grain obtained over a wide geographical area before processing and distribution, and therefore it is unlikely that an individual or group of individuals would consume grain produced by a single producer Products from pigs and poultry were not included due to these livestock normally being housed inside and supplied with feed from a number of sources, most of which will be located at some distance from the site of interest [RD73]. Cow milk, cattle meat and sheep meat were all assumed to be produced at a local farm location whilst green vegetables, root vegetables and fruit were assumed to be grown in a garden or allotment adjacent to the residential location Occupancy was taken as the highest occupancy of residency recorded in the last three habit surveys. The fraction of time spent outdoors was taken to be 50% for adults, 20% for children and 10% for infants. Breathing rates for the Farming Family were derived from data published by the National Radiological Protection Board (NRPB) see Appendix I The habit data and other parameters used to estimate the radiological dose to the Farming Family are summarised in Table 7.1. Page 167 of 255

182 Table 7.1 Assessment Parameters for the Farming Family (Gaseous Discharges) Parameter Adult Child Infant Cow s milk* (l/y) Root vegetables* (kg/y) Green vegetables (kg/y) Fruit and wild food** (kg/y) Cattle Meat (kg/y) Sheep meat (kg/y) Occupancy (h/y) 8,656 8,656 8,656 Fraction of time indoors Cloudshine factor Groundshine factor Breathing rates (m 3 /h) Notes * Top two foodstuffs, 97.5 th percentile consumption rates. ** The values presented for fruit are the sum of the consumption rates of domestic fruit and wild/free foods, which is comprised of blackberries, sloes, crab apples and damsons. Most Exposed Members of the Public for Aqueous Discharges 505. The most exposed members of the public for aqueous discharges were assumed to be a fishing family who consume seafood at critical rates. The adults go fishing near to the coast and the children and infants spend time playing on the beach. All activities are assumed to take place within the local marine compartment. All ingestion rates and occupancies were assumed to be at the 97.5 th percentile, using the median value of the 97.5 th percentile values from the three CEFAS reports [RD68], [RD69], [RD70] see Appendix I Individuals fishing for sea bass off the rocks adjacent to the Existing Power Station have been identified within the CEFAS surveys. However, it is not clear whether catches made in this small area are consumed by the fishing families or sold. For marine foodstuffs, 100% of crustaceans and molluscs were assumed to be caught in the local marine waters adjacent to the Power Station site, whereas 50% of the fish consumed were assumed to be caught in the local compartment with the remaining 50% caught from regional waters further offshore from the coast of north Wales. Consumption of seaweed was not considered as no observations of seaweed consumption were reported in [RD68] The activities that were assumed to be carried out in inter-tidal areas are boat maintenance, dog walking, beach warden and nature reserve duties, walking and angling. For the purposes of the assessment, the combination of these activities undertaken on sand, sand and stone, and mud and sand, was utilised to establish the number of hours spent per year in the inter-tidal zone and the median of the 97.5 th percentile values was used. As the activities are located in the inter-tidal zone, they have been conservatively assumed to occur at a distance of 1 m from the sea and so lead to exposure to sea spray. Adult Page 168 of 255

183 members of the fishing family (and to a far lesser extent, children) were considered to handle fishing gear in the intertidal area (1 m from the sea) in addition to the time spent on the beach The habit data used to estimate the radiological dose to the fishing family are summarised in Table 7.2. Once again, the derivation of breathing rates is described in Appendix I. Table 7.2 Habit Data for the Fishing Family (Aqueous Discharges) Parameter Adult Child Infant Fish (kg/y) Crustaceans (kg/y) Molluscs (kg/y) Inter-tidal activities (h/y) 1, Handling fishing gear, catch and sediment (h/y) 1, Breathing rate (m 3 /h) Candidates for the Representative Person from Exposure to Gaseous and Aqueous Discharges 509. In order to determine the exposure of the candidates for the Representative Person (CRP) from both aqueous and gaseous discharges, two cases were considered: the most exposed member of the public for gaseous discharges (the farming family) consumes locally sourced seafood at mean rates, and spends time at a local beach; and, the most exposed member of the public for aqueous discharges (the fishing family) consumes locally produced terrestrial foodstuffs at mean rates, and lives in close proximity to the site. Farming Family 510. In the case of the farming family the following combined exposure pathways were assessed: External exposure to beach sediments; Inhalation of sea spray when on the coast; Consumption of sea fish, crustaceans and molluscs caught locally; Internal irradiation from inhalation of radionuclides in the plume and re-suspended following deposition; External irradiation from radionuclides in the plume; External irradiation from radionuclides deposited on the ground; and, Internal irradiation from consumption of contaminated terrestrial foodstuffs following deposition on the ground (only for radionuclides with a half-life greater than three hours). Page 169 of 255

184 511. The consumption rates of seafood were assumed to be at mean rates, using the median of the mean values from the last three CEFAS reports. It was, however, assumed that the farming family will spend the same amount of time in a year in the inter-tidal zone as the fishing family, and so the median of the 97.5 th percentile values have each been used for intertidal occupancy. The habit data used to estimate the radiological impact to the CRP are provided in Table 7.3. Table 7.3 Farming Family Habit Data (Gaseous and Aqueous Discharges) Parameter Adult Child Infant Terrestrial Exposure Cow s milk* (l/y) Root vegetables* (kg/y) Green vegetables (kg/y) Fruit and wild food** (kg/y) Cattle Meat (kg/y) Sheep meat (kg/y) Occupancy (h/y) 8,656 8,656 8,656 Fraction of time indoors Cloudshine factor Groundshine factor Breathing rates (m 3 /h) Marine Exposure Fish (kg/y) Crustaceans (kg/y) Molluscs (kg/y) Inter-tidal activities (h/y) 1, Fishing Family 512. In the case of the fishing family the following combined exposure pathways were assessed: External exposure to beach sediments; External irradiation from handling fishing gear; Inhalation of sea spray when on the coast; Consumption of sea fish, crustaceans and molluscs caught locally; Internal irradiation from inhalation of radionuclides in the plume and re-suspended following deposition; External irradiation from radionuclides in the plume; External irradiation from radionuclides deposited on the ground; and, Internal irradiation from consumption of contaminated terrestrial foodstuffs following deposition on the ground (only for radionuclides with a half-life greater than three hours). Page 170 of 255

185 513. This family was now assumed to live in close proximity to the site, and be exposed to the same pathways as the farming family, i.e. the plume (inhalation and external exposure) and also deposited radionuclides (including the consumption of terrestrial foodstuffs) from gaseous discharges. However, the consumption rates of terrestrial foodstuffs were assumed to be at mean rates, using the median of the mean values from the three latest CEFAS reports The occupancy rates for the fishing family at its residence were not captured in the CEFAS habits survey. Therefore, a conservative assumption was made that the family has 100% occupancy at the residence when not at work (fishing at sea) or spending time at the beach (including handling fishing gear). The fraction of time spent outdoors was assumed to be 50% for adults, 20% for children and 10% for infants. The habit data used to estimate the radiological impact to this CRP are provided in Table 7.4. Table 7.4 Marine Exposure Fishing Family Habit Data (Aqueous and Gaseous Discharges) Parameter Adult Child Infant Fish (kg/y) Crustaceans (kg/y) Molluscs (kg/y) Inter-tidal activities (h/y) 1, Handling fishing gear, catch and sediment (h/y) 1, Breathing rate (m 3 /h) Terrestrial Exposure Cow s milk (l/y) Root vegetables (kg/y) Green vegetables (kg/y) Fruit (kg/y) Cattle Meat (kg/y) Sheep meat (kg/y) Time at resident location (h) 5,874 8,136 8,724 Fraction of time indoors Cloudshine factor Groundshine factor Breathing rates (m 3 /h) Page 171 of 255

186 Magnox Workers 515. An additional CRP for gaseous discharges is a worker at the Existing Power Station. This individual is assumed to live outside of the local area and only be exposed to gaseous discharges during the working day As decommissioning of the neighbouring Magnox site progresses, workers from nonnuclear sectors (e.g. construction and demolition firms) could be contracted to support the effort and could potentially be exposed to radioactivity from the proposed Power Station. These contractors are generally more likely to live away from the immediate local area. It was also considered that food pathways would be excluded from the assessment, i.e. that the Magnox worker would only be exposed to direct radiation and the various pathways relating to gaseous discharges during their hours of work on the site It has been assumed that the Magnox worker is always outdoors and works at ground level, at a receptor location that is 480 m and 34⁰ from the Power Station s reference stack. The Magnox worker dose assessment is for an adult only. Habit data for the Magnox worker are presented in Table 7.5. Table 7.5 Habit Data for the Magnox Worker (Gaseous Discharges) Parameter Adult Time at location (h) 2,000 Fraction of time indoors 0 Breathing rate (m 3 /h) 1.12 Assessment of Doses to the Foetus and Breast-fed Infants 518. Guidance from PHE [RD74] suggests that doses to the foetus need only be considered for four radionuclides (i.e. P-32, P-33, Ca-45 and Sr-89) in assessments where these radionuclides form a significant part of any release to the environment. Only Sr-89 has been identified as a constituent of the radioactive inventory to be discharged from the Power Station, and it is expected that this radionuclide will form only a very small part (very much less than 0.1%) of the release to the environment Assessment results show that foetal exposure is lower than that to one year old infants. Exposure to offspring will therefore not be explicitly reported. The exposure to offspring is reported in full in Appendix L. Page 172 of 255

187 Inhalation Gamma from Plume Beta from Plume Gamma from Ground Beta from Ground Resuspension Foodstuffs Total Terrestrial Dose Wylfa Newydd Project - Radioactive Substances Regulation Environmental Permit Application Results 520. The results from the dose assessments are presented below. Exposure pathways and dose estimates are set out along with the identification of the main radionuclides contributing to the dose. Doses to the Most Exposed Members of the Public from Exposure to Gaseous Discharges 521. The doses to the farming family due to gaseous discharges are summarised in Table 7.6. A full breakdown of doses by radionuclide and pathway for each age category is presented in Table A.67 to A.69 of Appendix O. For each age group, food consumption is the dominant pathway for dose, with milk being the main contributor (40%, 53% and 78% for adult, child and infant respectively). Cattle meat and root vegetables are the next greatest contributors. The ingestion and inhalation doses are largely due to C-14 (> 90% for all age groups). External dose from the plume is mainly associated with Ar-41, while external dose from the ground along with resuspension dose are mainly due to I-131. Table 7.6 Dose to the Farming Family by Exposure Pathway (Gaseous Only) Age Group Annual Dose by Exposure Pathway (µsv) Adult 1.39E E E E E E E E+01 Child 1.11E E E E E E E E+01 Infant 8.92E E E E E E E E+01 Doses to the Most Exposed Members of the Public from Exposure to Aqueous Discharges 522. The dose to the fishing family due to aqueous discharges is summarised in Table 7.7. A full breakdown of doses by radionuclide and pathway for each age category is presented in Table A.71 to A.73 of Appendix O. For each age group food consumption was the dominant pathway for dose, with crustaceans being the main contributor (38%, 43% and 49% for adult, child and infant respectively). The remainder of the dose was associated with gamma radiation from beach sediment (14%, 14% and 2% for adult, child and infant respectively). In terms of radionuclides, the main contributors to dose were Co-60 (26%, 37%, 37%, through external radiation) and H-3 (41%, 24% and 27%, through marine foodstuff ingestion). Page 173 of 255

188 Total Terrestrial Dose Crustacea Fish Mollusca Beta (sediment) Gamma (sediment) Sea Spray Combined Total Crustacea Fish Mollusca Beta (sediment) Beta (fishing) Gamma (sediment) Gamma (fishing) Sea Spray Total Marine Dose Wylfa Newydd Project - Radioactive Substances Regulation Environmental Permit Application Table 7.7 Dose to the Fishing Family by Exposure Pathway (Aqueous Only) Annual Dose by Exposure Pathway (µsv) Age Group Adult 2.58E E E E E E E E E-05 Child 1.46E E E E E E E E E-05 Infant 6.86E E E E E E E E E-05 Doses to the Candidate Representative Person from Exposure to Gaseous and Aqueous Discharges Farming Family 523. The dose to the farming family due to combined gaseous and aqueous discharges is summarised in Table 7.8. The figures for Total Terrestrial Dose are taken from Table 7.6. A full breakdown of doses by radionuclide and pathway for each age category is presented in Table A.75 to A.77 of Appendix O. Doses to the farming family from the combined discharges are dominated by doses due to the gaseous pathway. Doses due to aqueous discharges are mainly associated with the consumption of seafood, but the total contribution to dose from aqueous discharges is negligible compared to that from the terrestrial pathways. The radionuclide which contributes the most to the overall dose from discharges is C-14 (greater than 90% for all age groups). Table 7.8 Dose to the Farming Family by Exposure Pathway (Aqueous and Gaseous) Annual Dose to the Farming Family (Gaseous and Aqueous Discharges) by Exposure Pathway (µsv) Age Group Adult 1.97E E E E E E E E+01 Child 2.06E E E E E E E E+01 Infant 3.77E E E E E E E E+01 Fishing Family 524. The dose to the fishing family due to combined aqueous and gaseous discharges is presented in Table 7.9. The figures for Total Marine Dose are taken from Table 7.7. A full breakdown of doses by radionuclide and pathway for each age category is presented in Table A.78 to A.80 of Appendix O. Doses to the fishing family from both aqueous and gaseous discharges are also dominated by doses due to the gaseous pathway. Consumption of cow s milk is the main contributor to the total dose (35%, 46% and 74% Page 174 of 255

189 Inhalation Gamma from Plume Beta from Plume Gamma from Ground Beta from Ground Resuspension Total Total Marine Dose Inhalation Gamma from Plume Beta from Plume Gamma from Ground Beta from Ground Re-suspension Foodstuffs Combined Total Wylfa Newydd Project - Radioactive Substances Regulation Environmental Permit Application for adult, child and infant respectively). The radionuclide which contributes the most to the overall dose from discharges is C-14 (greater than 91% for all age groups). Table 7.9 Dose to the Fishing Family by Exposure Pathway (Aqueous and Gaseous) Annual Dose to the Fishing Family (Aqueous and Gaseous Discharges) by Exposure Pathway (µsv) Age Group Adult 6.81E E E E E E E E E+01 Child 3.38E E E E E E E E E+01 Infant 1.39E E E E E E E E E+01 Dose to the Magnox Worker 525. The dose to the Magnox worker due to gaseous discharges is summarised in Table 7.10 and a full breakdown by radionuclide and pathway is presented in Table A.81 in Appendix O. For the Magnox worker, 91% of the dose comes from inhalation of the plume, and 9% is due to external radiation by gamma emitters in the plume. The inhalation dose is mainly associated with C-14 (93%) and the external dose with Ar-41 (>99%). The small contribution to the total dose from external radiation from the ground and resuspension is largely due to I-131 (85-98%). Table 7.10 Dose to the Magnox Worker (Gaseous Discharges) by Exposure Pathway Age Group Annual Dose by Exposure Pathway (µsv) Adult 2.03E E E E E E E+00 Dose to Foetus and Breast-fed Infants 526. Exposure of the foetus can occur via transfer of radioactivity across the placenta and by means of photon irradiation from radioactive deposits in the tissues of the mother and placenta. Exposure can also occur postnatally via the ingestion of contaminated milk during breast feeding. It is possible to derive the dose to offspring by calculating the ratio of dose coefficients between the offspring and the mother, and applying the ratio to the mother s estimated dose (see Appendix L). In this context, the mother s dose originates from intake by inhalation and ingestion only (no external radiation pathways). Page 175 of 255

190 527. Assuming a 40 week pregnancy and exposure to mother s milk up to the age of six months, it is estimated (Appendix L) that the dose to offspring will be: Representative marine group (fishing family) offspring: 31.1 µsv Representative terrestrial group (farming family) offspring: 36.8 µsv 528. It is thus concluded that the exposure of offspring due to discharges into the environment at the proposed annual discharge limits are bounded by the dose to infant farming family (37.7 µsv see Table 7.8) Assessment of Doses due to Planned Short-term Discharges 529. An assessment was made of doses to the public in the vicinity of the Power Station, and the Magnox worker at the site boundary resulting from planned short-term gaseous discharges The dose to a member of the public residing throughout the discharge at the residential location is based on the 95 th percentile air concentrations resulting from short term discharges from the R/B 2 stack during the winter months over the ten year period The dose to the Magnox worker was based on the 95 th percentile air concentrations resulting from short term discharge from the R/B 2 stack over the whole ten year period R/B 2 discharges were selected as these corresponded to the highest short term ground level air concentrations. The different scenarios for the member of the public at the local residence and the Magnox worker have been selected to reflect the respective worst case at each of the two receptor locations Fuel pin failure has been identified as the principal Expected Event that could result in elevated levels of gaseous radionuclide discharges to atmosphere over the short term (see Section 5). It is noted that the elevated activity associated with a plant shutdown would also be discharged to the atmosphere (irrespective of a fuel pin failure), but fuel pin failure has been used as the bounding case for the purpose of this radiological assessment Methodology Modelling 532. Atmospheric Dispersion Modelling Software (ADMS) version 5.2 [RD76] was used to model the discharge of four representative radionuclides with different characteristics in the vicinity of the Power Station: H-3 (as HTO), Kr-85, Cs-137 and I-131. ADMS is a new generation Gaussian plume air dispersion model which uses the boundary layer depth and the Monin-Obukhov length to characterise the atmospheric boundary layer properties rather than using the single parameter Pasquill-Gifford class. It is continually validated against available measured pollutant concentration data obtained from real world situations, field campaigns and wind tunnel experiments by the model developers, CERC, and is shown to have good agreement with observed values. A list of recent model validation studies for buildings, complex terrain and flat terrain can be found on the CERC website [RD75]. 22 No planned short-term release events have been identified for aqueous discharges as liquid radioactive discharges are monitored in the final discharge tank prior to discharge, and discharge is only permitted when monitoring data demonstrates that it is at or below permitted limits. This will be managed through operational arrangements. Page 176 of 255

191 533. The outputs from ADMS were 24-hour rolling averages of the concentration in air, dry deposition rate, wet deposition rate and total deposition rate for the four representative radionuclides. From these results, the 95 th percentiles were calculated for 10 years of meteorological data (see input data below) and for selected periods: Summer (01 June 31 August) and Winter (01 December 28 February) The above parameters were determined for three receptor points located at distances of 555 m, 1069 m and 480 m at bearings of 95⁰, 106⁰ and 34⁰ respectively from the reference stack (Main Stack 1). The three locations correspond to the food production location, residential location, and the Magnox worker. Dose Calculation 535. The radionuclides that are released from the Expected Event are noble gases so the only exposure pathways that need to be considered are those due to submersion in the plume (cloud dose). This is because these radionuclides do not deposit and so do not accumulate in the environment or food chain. Also, they are not absorbed through inhalation into the body and therefore would not contribute to an internal exposure via that pathway Cloudshine doses are calculated for the period of the passage of the plume using standard methods. External doses from immersion in air containing radioactive noble gases are based on cloud gamma and cloud beta skin dose coefficients taken from DCFPAK v 3.2 [RD77] and PC-CREAM 08 [RD64] respectively External dose from submersion in the plume was calculated using the following formula: Dose AirC T DC ( LF O LF O i cloud / n cloud / cloud i cloudo o ) Eq. 7.1 Where: - AirC is the ground level air concentration (Bq/m 3 ) = discharge rate (Bq/s) x dispersion coefficient (s/m 3 ); - T is the exposure time (s); - DC cloudβ/γ is the gamma or beta external dose coefficient for submersion for radionuclide n (Sv/s per Bq/m) taking daughter radionuclides into account; - LF cloud i,o is the cloud gamma location factor for indoors and outdoors occupancy; and, - O i,o is the occupancy factor for indoors and outdoors occupancy Location factors have been taken from [RD78]. The location factor for cloud gamma is 0.2 whilst that for cloud beta is set to 1. The indoor dose reduction factor for inhalation is set to 1 [RD78]. Page 177 of 255

192 539. An estimate of total effective dose due to immersion in the noble gas plume was found from: EffectiveDose Dose 0.01Dose Eq. 7.2 cloud cloud Where 0.01 is the tissue weighting factor for skin [RD54] Input Data Source Term 540. The gaseous discharges associated with a fuel pin failure are presented in Table 5.8. For the purposes of this short-term assessment it is conservatively assumed that the total discharge is released uniformly over a period of 24 hours. It is also assumed that, although the proposed Power Station is a twin reactor unit, an expected event will only occur at one reactor at any one time. The likelihood of an expected event happening simultaneously at two reactors is found to be approximately 1:500 over an operating cycle. Therefore, the radiological consequences of a single expected event have been determined. Modelling 541. The discharge characteristics of each R/B stack diameter, height, volumetric flow rate are presented in Table 3.7. Modelling was undertaken with ADMS (as opposed to PC- CREAM 08 ) so an actual stack height of 75 m was used. The ADMS 5.2 model was run using 10 years of hourly sequential data (2007 to 2016) from the Meteorological Office NWP models [RD66] for the latitude and longitude of the proposed development. The locations of the receptors are given in Table Table 7.11 Specified Receptor Locations Name OS Coordinate X (M) Y (M) Z (M) Food Residential Magnox Habits Data 542. In assessing a 24 hour release duration, it was necessary to consider representative group habits as receptors cannot be assumed to be outdoors for the entire duration of the release or breathing continually at heavier rates. For the 24 hour discharge scenario, individuals were assumed to be at the residential location for the whole duration of the release, 24 hours. The indoor and outdoor occupancy was taken to be the same as that defined in the EA s Initial Radiological Assessment methodology [RD79], i.e. the adult, child and infant are indoors 50%, 80% and 90% of the time, respectively. The Magnox worker was Page 178 of 255

193 assumed to be outdoors for 8 hours. After 8 hours the worker leaves the location and their exposure is terminated Results 543. A summary of the effective dose incurred by a member of the public at the residential location due to the short term discharge is given in Table 7.12, and the effective dose incurred by a Magnox worker at the site boundary is given in Table The calculated Kr-85 ground level concentrations resulting from short-term discharges are presented in Table A.82 in Appendix O. Table 7.12 Estimated Doses at the Residential Location due to the Short-term Release Age Group Dose due to Short-term Release (µsv) Cloud Gamma Cloud Beta Total* Adult 1.11E E E-04 Child 6.66E E E-05 Infant 5.18E E E-05 Note. * A tissue weighting factor of 0.01 has been applied to the beta skin dose Table 7.13 Estimated Doses Incurred by a Magnox Worker due to a Short-term Release Age Group Dose due to Short-term Release (µsv) Cloud Gamma Cloud Beta Total* Adult 8.53E E E-05 Note. * A tissue weighting factor of 0.01 has been applied to the beta skin dose 544. The predicted doses due to a discharge of noble gases as a result of the expected event is very low. The maximum predicted exposure due to immersion in the plume is an effective dose of 1.1E-04 µsv for the local resident adult. The corresponding dose incurred by a Magnox worker at the site boundary for eight hours during the release is 8.7E-05 µsv In the unlikely situation of an expected event occurring simultaneously at both reactors then the radiological consequences can be estimated by scaling the estimated total doses given above by the ratio of the ground level air concentrations presented in Table A.82 in Appendix O. It is estimated that an adult member of the public residing at the residential location for the duration of the release would incur a dose of 1.8E-04 µsv 23 and a Magnox worker present in the open air at the site boundary for eight hours would incur a dose of 1.2E-04 µsv 24. These predicted doses are negligible and in any event are significantly below the dose limit for a member of the public of 1000 µsv/y and the source constraint of 300 µsv/y. 23 Calculated by applying the ratio of Table A.69 Winter values for Reactors 1&2/Reactor 2 at the residential location to the Table 7.12 Adult dose. 24 Similarly calculated but applying 10 year values for the Magnox worker to the Table 7.13 Total dose. Page 179 of 255

194 7.1.5 Assessment of Doses due to Direct Radiation 546. An assessment was made of the annual dose to the most exposed members of the public residential receptor (to estimate doses to the Farming Family and Fishing Family), Magnox worker and walkers along three footpaths designated Walkers 1, 2 & 3 (see Appendix E) from direct radiation from the Power Station Methodology Modelling 547. The dose rates due to direct radiation from radioactive sources on the Power Station site were calculated using the software package MCNP [RD80]. MCNP is a general-purpose Monte Carlo N Particle code, developed by Los Alamos National Laboratory in the USA, and is recognised by the UK s nuclear regulators as a standard shielding code. It has the capability to simulate particle interactions involving neutrons, photons, and electrons and can model skyshine, i.e. ionising radiation reaching the earth's surface indirectly through reflection and scattering in the atmosphere. MCNP has been used in this analysis to calculate the dose rates from both photons and neutrons at the receptor locations The direct dose rates were calculated over a range of distances from the site boundary out to the receptor locations. The receptor distance from site was extended to approximately 1,500 m to allow the modelling to cover all receptor positions. The MCNP models do not consider local mounding or terrain, except for modelling explicitly the different levels of the buildings within the site. The surrounding terrain was thus modelled as a flat plane at +26 m AOD (above ordnance datum) Calculations were undertaken to represent both generating units operating at full power and with all buildings filled to their design capacity with due regard to any source term decay that could reasonably be claimed. All dose rates from each building of interest were calculated taking account of the shielding effects of other buildings on-site and taking into account the heights, and different ground levels AOD, of all relevant buildings within the site. Only bulk shielding was assessed in this work. No shielding weaknesses associated with detailed building designs (e.g. penetrations) were included in the calculations. A detailed account of the approach used to model the estimated doses is provided in Appendix N. Receptor Positions 550. The receptor positions considered in the assessment were as follows: Residential Location A representative location for a family who lives near to the site and eats food grown and produced locally; Magnox Worker A worker on the Existing Power Station site which lies due north of the proposed new site; Walkers along three footpaths designated Walkers 1, 2 & 3 ; Tregele A settlement located approximately 600 m from the site; and, Cemaes A settlement located approximately 1.2 km from the site The number of receptor positions used in the analysis and the locations of these receptors is outlined in Figure 7.1 where the (0,0,1700) co-ordinates indicate the position of the centre Page 180 of 255

195 of base of R/B 1 and all other co-ordinates are in units of cm. North is along the Y axis and dose rates were calculated for locations representative of the routes that may be taken by walkers, so the Walker #3 position is the static position for this walker (see Section ). Figure 7.1 Relative Positions of Buildings and Receptors Note. The receptors ATR#2 (Additional Terrestrial Receptor), ATR#3 and Temporary Accommodation have been assessed for purposes unconnected with the Horizon EP-RSR Application and are therefore not reported here. ATR#1 (Terrestrial Habitat) has been used in the non-human biota assessment only. Page 181 of 255

196 Dose Calculation 552. The overall dose at a receptor location was calculated as the dose rate at this location multiplied by the exposure time, taking into account the reduced dose rate while indoors: Dose = D i x (LF i x O i + LF o x O o) Eq. 7.3 Where: D i = external dose rate, µsv/hr Input Data Source Terms LF i = Indoor location factor = 0.1 LF o = Outdoor location factor = 1 O i = Hours per year spent indoors O o = Hours per year spent outdoors 553. The structures containing radioactive sources on the site are shown in Table Dose analysis performed during Generic Design Assessment (GDA) indicated that the contribution of the R/B, Rw/B, C/B, S/B and ILW Storage Facility to the direct radiation dose was orders of magnitude below the contribution from other buildings/facilities on site. These sources were therefore screened out from the current assessment. Table 7.14 Structures Containing Radioactive Sources on Site Main Buildings (1) Turbine Building (T/B) (2) Reactor Building (R/B) (3) Radwaste Building (Rw/B) (4) Control Building (C/B) (5) Service Building (S/B) Auxiliary Facilities (1) Spent Fuel Storage Facility (SFSF) (2) Intermediate Level Waste (ILW) Storage Facility (3) Lower Activity Waste Management Facility (LAWMF) 554. The proposed layout of the Power Station is illustrated in Figure 2.2. The site is assumed to be made up of three levels. The Single Power Island is on ground 16 m ± 2 m AOD. There is a 21 m AOD level where the Backup buildings, LAWMF, SFSF and ILW Storage Facility reside. Outside of the site, the ground is assumed to consist of a single level at 26 m AOD Further detailed information for the source term and model parameters used in the direct radiation assessment are provided in Appendix N. Page 182 of 255

197 Habits Data and Assumptions 556. The modelling for the direct dose rates assumed that both of the proposed units were in place and operating at full power. All waste storage facilities were assumed to contain inventory as expected at 60 years of operation. This scenario is considered to be bounding for all other scenarios in the operational lifetime of the Power Station For each of the receptors (except the walkers) the following are assumed: 2000 hours occupancy per year for a worker on the adjacent Existing Power Station site; For the Residential Location plus the nearest points to the site of the villages of Tregele and Cemaes: - Occupancy of 24 hrs per day for 365 days (i.e hrs) - Fractions of time assumed indoors: Adult 0.5, Child 0.8, Infant The following scenarios are outlined for walkers on the paths around the Power Station (i.e. existing footpaths around the site or paths that have been rerouted) as outlined in Figure 7.2. Walker 1 Someone who parks at the visitors centre and walks their dog every day in a north-westerly direction (location 1 in Figure 7.2); Walker 2 Someone who parks at the visitors centre and walks their dog every day in a north-easterly direction (location 2); and, Walker 3 Assumes that, once per week, Walker 2, spends 20 minutes at the top of the mound It is assumed that each walker walks at a constant speed of 1.3 m/s every day of the year and performs a round trip along the appropriate length of path identified as important for the dose rate assessments In order to model the paths shown in Figure 7.2 explicitly, each path was split into three sections with an associated length and time for walking. The walker was assumed to travel the length of the relevant path two times a day i.e. an outward and a return journey. These individual sections of path were input into the MCNP models to allow the average dose rate over each section to be calculated. Page 183 of 255

198 Figure 7.2 Walker Exposure Locations for Direct Radiation Assessment Results 561. The annual doses for each of the main receptor positions, from all of the sources identified in Figure 7.1 and Table 7.14, are given in Table 7.15 (dose rate data is given in Table A.50 of Appendix N). All doses include exposure to both direct radiation and skyshine (airscattered radiation). The annual doses for the three walkers are presented in Table Annual doses were also determined for the Farming Family and Fishing Family based upon the occupancy data given in Table 7.3 and Table 7.4. The dose rates are presented in Table Page 184 of 255

199 Table 7.15 Annual Doses to Main Receptors Receptor Dose Rate (µsv/y) Magnox Worker Residential Location Tregele Cemaes Adult 4.25E E E E-03 Child E E E-03 Infant E E E For the walkers, the highest dose rates experienced on any individual section of the paths were found to be: 1.71E-02 µsv/h for Walker 1; 9.15E-05 µsv/h for Walker 2; and, 2.36E-04 µsv/h for Walker 3. These dose rates produced the annual doses shown below. Table 7.16 Annual Doses to Walkers Walker No. Walker Dose (µsv/y) E+00* E E-03 Note*: For Walker 1 more than 90% of the total dose comes from the SFSF. Table 7.17 Annual Doses to Farming and Fishing Families Receptor Dose Rate (µsv/y) Farming Family Fishing Family Adult 6.18E E-02 Child 3.14E E-02 Infant 2.13E E-02 Page 185 of 255

200 Gaseous and Aqueous Direct Radiation Combined Total Gaseous and Aqueous Direct radiation Combined Total Gaseous Direct Radiation Combined Total Wylfa Newydd Project - Radioactive Substances Regulation Environmental Permit Application 563. Dose rates were generally found to be low: doses to the public (adult, child and infant) are less than 1 µsv/y in all cases apart from Walker 1 whose dose was dominated by contributions from the SFSF. The turbine building for Unit 1 dominates the dose to the Magnox worker on the Existing Power Station site. The dose to walkers is dominated by a different building to that which dominates the dose to the Magnox worker due to their different locations The SFSF is the main contributor to direct off-site doses at the residential location, being responsible for 80% of the annual dose, followed by T/B 2 at 17% and T/B 1 at 3% For comparison, the range of gamma dose rates measured a few kilometres from the Existing Power Station site ranged from 20 to 60 ngy/h [RD71], which equates to µsv/y (assuming that a person is exposed at exactly the same location for an entire year and to gamma only). This illustrates that predicted impacts from direct radiation from the proposed Power Station would be less significant than the current contributions from the Existing Power Station added to any natural and other sources in the area (including radiation fallout from the atmosphere) Representative Person and Compliance with Source Constraint 566. The dose to the candidates for the Representative Person due to gaseous and aqueous discharges and direct radiation is presented in Table 7.18 for the three age groups within the Farming Family and the Fishing Family. The dose to the Farming Family from gaseous and aqueous discharges is taken from Table 7.8, and the same to the Fishing Family from Table 7.9. Doses from relevant pathways for the Magnox worker are also provided. It can be seen that, out of the candidates for the Representative Person, it is the infant member of the Farming Family who has been found to be the most highly exposed. The Farming Family infant is thus nominated as the Representative Person. Table 7.18 Annual Dose to the Candidates for the Representative Person Age Group Annual Dose for Gaseous and Aqueous Discharges and Direct Radiation (µsv) Farming Family Fishing Family Magnox Worker Adult 1.97E E E E E E E E E+00 Child 2.06E E E E E E+01 N/A N/A N/A Infant 3.77E E E E E E+01 N/A N/A N/A 567. The prospective dose assessment results summarised in Table 7.18 show that future annual doses due to gaseous and aqueous discharges, and direct radiation from the Power Station are predicted to be below the source constraint of 300 µsv/y for all age groups. There is significant margin between the predicted prospective doses and the source constraint. Page 186 of 255

201 568. As noted earlier, some of the assumptions that form the basis of the dose assessment were deliberately conservative to ensure that the predicted doses are overestimated and that the actual dose to members of the public will be below statutory constraints and limits. Examples of conservatism include: An effective discharge height of 15 m (using a 15 m stack height in PC-CREAM 08 is likely to result in lower dispersion and therefore higher concentrations than if releases were modelled from the actual stack height of 75 m); An assumed production of both milk and meat occurring very close to the site boundary, at 555 m from the reference stack; and, The residential location being located at only 1,069 m from the reference stack As such it is expected that use of less conservative assumptions than those above would result in a significantly lower estimate of doses due to gaseous discharges. For example, using the design discharge height of 75 m is likely to result in a reduction of the air concentration, and thus the dose due to ingestion of C-14 in milk, the dominant pathway for the infant Assessment of Collective Dose to Populations 570. An assessment was made of collective doses to the UK, European and World populations Methodology 571. PC-CREAM 08 was used to determine collective doses for the UK, European and world populations for both first-pass and global circulation scenarios, truncated at 500 years, in accordance with the requirements set out in RSR-B3. The first pass refers to the collective dose due to the initial discharge, whereas the global circulation refers to the dose that arises from circulation of mobile, longer-lived radionuclides in the oceans and in the atmosphere, i.e. C-14, H-3 and Kr In addition, per caput doses were calculated based on the population data for UK, EU12, EU25 and world assumed in PC-CREAM 08 to be 59.6 million, 360 million, 456 million and 10 billion respectively [RD64]. For clarity, EU12 is the population representing the 12 European countries that were member states when the European Union (EU) was first established in EU25 is the number of member states (25) that were included in the EU when PC-CREAM 08 was developed (recognising that this number has since risen again). It is also recognised that population figures for each area have changed since the assessment software was last revised. However, it is not possible to alter these data within PC-CREAM 08. Gaseous Discharges 573. In the case of gaseous releases, the following pathways were considered in assessing the collective doses: External irradiation (β,γ) from the cloud; Inhalation of radionuclides in the cloud; External irradiation (β,γ) from deposited activity; Inhalation of re-suspended activity; and, Page 187 of 255

202 Ingestion of contaminated foodstuffs (based on population distribution, production rates of food and the time integrated concentrations within foodstuffs) Grid data around the site and beyond are already defined for a number of nuclear sites that are built in to the PC-CREAM 08 site library. Population and agricultural product distribution within Europe is provided by the in-built database for each site in the PC-CREAM 08 database as is the regional marine compartment that the discharge is released into PC-CREAM 08 divides the area around each discharge point into a number of annular segments, and assumes that the population and agricultural production distributions are uniform. The distributions of individual dose and radionuclide concentrations in the environment are also assumed to be uniform. Individual external and inhalation doses in each annular segment of the polar grid are scaled by the population in that segment to calculate the collective dose As for previous calculations, an effective discharge height of 15 m for gaseous releases was used in PC-CREAM 08. Aqueous Discharges 577. For aqueous discharges, collective doses were calculated using the radionuclide concentrations in each compartment and then summing to obtain the total collective dose. Collective occupancy on beaches, sea food catches and harvest data per year in each compartment were used to produce the collective doses within the regions of interest Habits Data and Assumptions 578. Default, in-built PC-CREAM 08 food consumption and occupancy habits data were used for the collective dose assessments. The models, food production and habits data embedded in the collective dose assessment model are based on the European Commission report RP72 [RD100]. This document makes the assumption that the magnitude of the population of the European Union remains constant over all time, that habits remain the same and that the whole population are adults Results 579. A summary is given in Table 7.19 of the collective dose per year of discharge to UK, EU12, EU25 and World populations due to gaseous discharges truncated to 500 years. Full breakdowns of doses by radionuclide are presented in Table A.89 to A.91 in Appendix O UK first pass collective doses due to gaseous discharges are primarily associated with ingestion of root vegetables (23%) and cow s milk (49%). EU first pass collective doses show a similar distribution with consumption of root vegetables (24%) and cow milk products (50%) being the dominant pathways. For both first pass and global circulation, the principal radionuclide was C-14, which contributed 94% to UK first pass, 97% to EU first pass and 100% to global circulation doses. Page 188 of 255

203 Table 7.19 Collective Dose per Year of Discharge due to Gaseous Discharges Population Collective Dose per Year of Discharge due to Gaseous Discharges First Pass (man Sv) Global Circulation (man Sv) Total (man Sv) Per Caput Dose (Sv) UK 1.95E E E E-09 EU 9.63E EU E E E-09 EU E E E-09 World E E E A summary of collective dose to UK, EU12 and World populations due to aqueous discharges truncated to 500 years is summarised in Table Full breakdowns of doses by radionuclide are presented in Table A.92 to A.94 of Appendix O. The first pass collective dose due to aqueous discharges was estimated by summation of fish, crustacean, mollusc and beach sediment gamma contributions. Table 7.20 Collective Dose per Year of Discharge due to Aqueous Discharges Population Collective Dose per Year of Discharge due to Aqueous Discharges First Pass (man Sv) Global Circulation (man Sv) Total (man Sv) Per Caput Dose (nsv) UK 4.99E E E E-05 EU E E E E-06 World 1.92E E E E The consumption of seafood contributes 62% to the UK collective dose, 42% to the EU12 collective dose and only 4% to the world collective dose. For UK, EU12 and the world the contribution from beach sediment is zero. First pass doses are the most significant for the UK population (62% compared to 38% for global circulation). The EU12 collective dose is more evenly split between first pass and global circulation (42% and 58% respectively), with world collective doses being dominated by global circulation (4% first pass and 96% global circulation). The dominant radionuclide is H-3 for UK (84%), EU12 (89%) and the world (99%) The IAEA considers practices giving rise to collective doses below 1 man Sv per year of operation may be exempted from regulatory control [RD61], [RD62]. As can be seen from the results presented above, the predicted collective dose from the proposed Power Station for aqueous discharges is below 1 man Sv for all population groups. For gaseous discharges the predicted collective doses to EU and World populations is above 1 man Sv. The collective dose for these population groups is dominated by C Individual per caput doses were estimated from the collective dose results using the population data for the UK, EU12, EU25 and World contained in PC-CREAM 08 (59.6 million, 360 million, 456 million and 10 billion respectively). The per caput doses from Page 189 of 255

204 marine discharges from the Power Station thus calculated are 1.35E-05, 8.66E-06 and 5.26E-06 nsv/y for populations of the UK, Europe (assumed to be EU12) and the World, respectively. Comparable per caput doses from atmospheric discharges from the Power Station are equal to 9, 9, 8 and 6 nsv/y (UK, EU12, EU25 and World respectively) With reference to per caput doses, the IAEA states that discharges giving rise to calculated average annual individual doses for a population group in the nanosievert (nsv/y) range or below should be ignored in the decision making process as the associated risks are miniscule [RD62]. The risks presented by the per caput doses presented in this section are therefore negligible Assessment of Build-up of Radionuclides in the Environment 586. An assessment was made of the build-up of radionuclides in the local environment at the end-of-life of the Power Station, and whether this could have the potential to prejudice legitimate users or uses of the land or sea Methodology 587. The assessment was based upon a comparison of the predicted build-up of radionuclides in the terrestrial and marine environments, resulting from 60 years of continuous Power Station operations (i.e at the end of station life), with current regional and local baseline monitoring data Other than for C-14 and H-3, the build-up of radioactivity in soils was determined using the PLUME and FARMLAND modules within the modelling code PC-CREAM 08. The estimate of the build-up in soils of H-3 and C-14 due to gaseous discharges was made using a specific activity (SA) model 25 as advised in [RD64] and [RD103], where the concentration in soil (for C-14) or soil water (for H-3) is assumed to be proportional to the concentration in air, the latter being given by PC CREAM 08 (details are provided in Appendix M) The DORIS module of PC CREAM 08 was used to determine the radioactivity concentration in unfiltered seawater and seabed sediment. The comparator baseline monitoring data was taken from OSPAR and RIFE reports Land Uses and Likely Future Changes 590. A review of Defra s MAGIC interactive tool and Horizon s GIS mapping tool, as well as drawings published in the Wylfa Newydd Scoping and Preliminary Environmental Information reports [RD101] has been undertaken to establish the current land use profile within an area of approximately 2 km radius from the reference emission stack. The Anglesey and Gwynedd Deposit Joint Local Development Plan (JLDP), which will guide spatial development across the County to 2026, has also been reviewed for an indication 25 The transfer of H-3 and C-14 between the atmosphere and the terrestrial environment is more complex than for other radionuclides, primarily because hydrogen and carbon are fundamental to biological systems. This, combined with the fact that both H-3 and C-14 are long lived and highly mobile within the environment, means that a different approach is required to estimate their incorporation within soil. Modelled air concentrations are used as an input to separate calculations to determine the amount of H-3 that combines with hydrogen in soil water and C-14 that combines with stable carbon within soil. The calculations ultimately provide specific activities (activity concentration of H-3 and C-14 per unit mass of soil) for each of the two radionuclides. Page 190 of 255

205 of acceptability of changes to current land use or the likelihood of new developments in close proximity to the Power Station being consented in the future The key points from this review are summarised below: Baseline: Potential developments to the west of the site will be constrained by the Irish Sea, and developments to the north will be constrained by the Existing Power Station site. Under the current NDA programme, the decommissioning of the Existing Power Station site is anticipated to continue over the lifetime of the Power Station. It is therefore considered that there is limited scope for changes in the current land use in respect of the areas to the west (i.e. occupational and recreational activities) and to the north (i.e. nuclear licensed site) of the proposed development area over the lifetime of the Power Station. The land to the south and the east of the Power Station site is predominantly agricultural and is largely used for the production of animal feed and as grazing pasture for livestock. The current site development plan is for the restoration of the areas outside the main Power Station boundary to the original land use. It is therefore considered that these areas will continue to serve as grazing pastures during the operational phase of the Power Station. Agricultural Land Use: Anglesey is predominantly rural and much of the land is agricultural. The Agricultural Land Classification (ALC) scheme 26,27 for Anglesey indicates that the majority of land surveyed within and around the proposed development site comprises grade 3 or lower quality soils [RD101]. These grades of agricultural land generally have moderate to poor yield of a narrow range of crops such as cereals and grasses. This is corroborated by CEFAS [RD68] which reported that crops grown in the area are used for the production of livestock feed and not consumed by humans. Aerial views from online maps 28 indicate that a large proportion of the land within this area comprises pastures with evidence of livestock grazing. Given the low ALC grade of the land, it is considered unlikely that significant changes to the current land use will occur during the lifetime of the Power Station. Recreational and Occupational Use of Coastline: CEFAS [RD68] has reported the use of the local coastline for various recreational activities including dog-walking, birdwatching, picnicking, water sports (swimming, kayaking, angling, boating, etc.) and other beach activities. Occupational activities reported by CEFAS include boat maintenance, bait digging and nature reserve warden duties. Planning Constraints: A review of the Anglesey and Gwynedd Deposit JLDP [RD127] and the associated Proposals Map and Constraints Map, key documents that will guide development across the County to 2026, was carried out to ascertain whether the JLDP s spatial strategies and strategic development objectives for the region (and any consequent changes to land use) could result in significant modification of the habits of the local populace in the future. Some provisions of the JLDP that seek to control the proliferation of new developments in the area include: - Protection of areas outside development boundaries and identified Clusters, i.e. the countryside, from development other than those uses that are essential to a Page 191 of 255

206 rural location and which would not harm its character and appearance (Settlement boundaries, Pg. 53). - Adoption of a restrained approach to development in the Countryside, supporting only selected forms of developments with economic, infrastructural, or social benefits (Settlement Hierarchy, Pg. 6 - Table 1). - Constraining non-essential development in the open countryside (Location of Housing, Pg. 88). - Refusal of proposals for new dwellings or significant modification of existing buildings for residential use within Coastal Change Management Area (CChMA). Refusal of permits for new non-residential permanent buildings not associated with an existing use or building in areas within the CChMA identified as being at risk from coastal change during the first indicative policy epoch up to 2025 (Coastal Change Management Area, Pg. 88). - Refusal of proposals that will lead to the loss of existing open space, including any associated facilities, which has significant recreational, amenity or wildlife value unless they conform to a set of criteria (Safeguarding Existing Space, Pg. 64) On the basis of the information reviewed it is considered reasonable to assume that there will be no (a) proliferation of new developments or (b) significant changes to current land use within the areas close to, and that could be impacted by, discharges from the proposed Power Station. The combination of the low quality of agricultural land and local planning constraints would restrict, to a great extent, significant changes to the current land use and assumptions made about the habits of the Representative Person are likely to remain valid. This is without prejudice to future economic, environmental or social priorities/ prerogatives Results 593. A summary of the potential build-up of radionuclides released in aqueous discharges from the Power Station is given in Table 7.21, and a similar summary for the build-up associated with gaseous discharges is presented in Table Table 7.21 Build-up of Activity in the Marine Environment over a 60-year Period (within the Local Compartment) Radionuclide Discharge (Bq/y) Activity Concentration after 60 Years Continuous Discharge Unfiltered Seawater (Bq/l) Seabed Sediment (Bq/kg) Ag-110m 9.4E E E-11 Am E E E-10 Ba E E E-09 C E E E+00 Ce E E E-07 Ce E E E-05 Cm E E E-10 Cm E E E-12 Page 192 of 255

207 Radionuclide Discharge (Bq/y) Activity Concentration after 60 Years Continuous Discharge Unfiltered Seawater (Bq/l) Seabed Sediment (Bq/kg) Cm E E E-10 Co E E E-07 Co E E E-05 Cr E E E-07 Cs E E E-08 Cs E E E-08 Fe E E E-03 Fe E E E-07 H-3 1.5E E E-02 I E E E-10 I E E E+00 I E E E+00 I E E E+00 I E E E+00 La E E E-08 Mn E E E-05 Nb E E E-06 Ni E E E-03 Np E E E-16 Pu E E E-09 Pu E E E-10 Pu E E E-09 Ru E E E-09 Ru E E E-08 Sb E E E-13 Sb E E E-08 Sb E E E-07 Sr E E E-09 Sr E E E-07 Tc E E E+00 Te E E E-23 Te-123m 8.3E E E-11 Te-125m 0.0E E E-07 U E E E-13 U E E E-17 Page 193 of 255

208 Radionuclide Discharge (Bq/y) Activity Concentration after 60 Years Continuous Discharge Unfiltered Seawater (Bq/l) Seabed Sediment (Bq/kg) U E E E-16 Xe-131m 1.9E E E-10 Xe E E E+00 Xe E E E+00 Zn E E E-06 Zr E E E-06 Table 7.22 Build-up of Activity in the Terrestrial Environment over a 60-year Period (at the Food Production Receptor) Radionuclide Discharge (Bq/y) Activity Concentration in Soil after 60 years Continuous Discharge (Bq/kg) Ag-110m 7.8E E-10 Am E E-13 Ba-137m 1.0E E-13 Ba E E-08 C E E+01 Ce E E-07 Ce E E-06 Cm E E-12 Cm E E-14 Cm E E-12 Co E E-07 Co E E-05 Cr E E-07 Cs E E-07 Cs E E-32 Cs E E-06 Cs E E+00 Fe E E-08 H-3 2.1E E+00 I E E-03 I E E-06 I E E-05 I E E+00 Page 194 of 255

209 Radionuclide Discharge (Bq/y) Activity Concentration in Soil after 60 years Continuous Discharge (Bq/kg) I E E-06 La E E+00 Mn E E-06 Nb E E+00 Np E E+00 Pr E E-12 Pu E E+00 Pu E E+00 Pu E E+00 Rb E E+00 Rb E E-08 Sb E E-10 Sb E E-07 Sb E E-06 Sr E E-07 Sr E E-06 Tc E E-10 Te-125m 2.0E E+00 U E E+00 U E E+00 U E E+00 Y E E-13 Zn E E-07 Zr E E-07 Comparison of Modelled Activity Concentrations Against Published Data Marine 594. The OSPAR Convention is the mechanism by which 15 Governments and the European Union cooperate to protect the marine environment of the North-East Atlantic ( The Radioactive Substances Committee of OSPAR has published baseline data for key radionuclides within a set of monitoring areas [RD104]. The intent is that these data are used to baseline trends in subsequent monitoring data, and thereby establish the effectiveness of discharge reductions on reducing environmental concentrations (the latter being one of the key aims of OSPAR). OSPAR Region III (The Celtic Seas) includes the Irish Sea, which in turn is comprised of three monitoring areas. One of these areas ( Area 6 ) includes Anglesey as shown in Figure 7.3. Page 195 of 255

210 Figure 7.3 Illustration of the OSPAR Monitoring Areas 595. Average baseline activity concentrations published by OSPAR [RD104] for three radionuclides in seawater from OSPAR Monitoring Area 6 are provided in Table The data are limited to three radionuclides. Table 7.23 Measured Radionuclide Activity Concentrations in Seawater from OSPAR Monitoring Area 6 Radionuclide Baseline Average Baseline Lower Bracket* Baseline Upper Bracket* (Bq/l) (Bq/l) (Bq/l) Page 196 of 255 H-3 <1.50E Cs E E E-01 Tc E E E+00 Note. * The baseline bracket represents the baseline value ±1.96 times the standard deviation. Discharge data are assumed to be normally distributed around the mean of the reported values and the bracket is therefore calculated as the interval which should contain 95% of all the values For H-3, the activity in unfiltered seawater predicted by PC-CREAM08 after 60 years of Power Station operation, Bq/l (see Table 7.21), represents just 0.32% of the average baseline activity measured by OSPAR. For Cs-137, the predicted activity in seawater, 4.17E-10 Bq/l, is many orders of magnitude below the average baseline activity measured by OSPAR Provisional environmental monitoring results for 2016 from the Food Standards Agency website ( do not contain any information regarding activity concentrations within marine sediment or seawater samples for the area around the

211 Existing Power Station at Wylfa (information correct as at June 2017). However, data from the marine monitoring programme associated with the Existing Power Station are given in the RIFE reports and the most recent data (for 2014 and 2015) [RD68], are reproduced below. Table 7.24 Measured Sediment and Seawater Activity Concentrations for the Wylfa Area in 2014 and 2015 Sample Type Location Mean Activity Concentration (Bq/kg)* Am-241 Cs-137 H Sediment Cemaes Bay Sediment Cemlyn Bay West Seawater Cemaes Bay <0.26 <0.18 <3.3 Seawater Cemlyn Bay West <0.28 < Sediment Cemaes Bay < Sediment Cemlyn Bay West < Seawater Cemaes Bay <0.30 <0.20 <3.1 Seawater Cemlyn Bay West <0.30 < Note. * The seawater activity concentration unit is Bq/l The data show that the extent of present day radionuclide accumulation in marine sediment and seawater around Wylfa from historical and current authorised releases (from the Existing Power Station and other nuclear licensed installations discharging into the Irish Sea) is low Adding the activity concentrations predicted by PC-CREAM 08 (from Table 7.21) to the relevant data in Table 7.24 makes negligible difference to the existing radionuclide burden. Taking H-3 as an example, the amount of additional H-3 in seawater that could potentially arise over 60 years through operation of the Power Station (0.048 Bq/l) constitutes less than 2% of the existing measured H-3 content of seawater in the area (approximately 3 Bq/l). For Am-241 and Cs-137, the model predicts activity concentrations that are many orders of magnitude below those recorded in marine samples in 2014 and Terrestrial 600. Provisional environmental monitoring results for 2016 from the Food Standards Agency (FSA) website ( do not contain any information on activity concentrations within soils for the area around the Existing Power Station at Wylfa (information correct as at June 2017). Furthermore, published data for terrestrial sampling around the Existing Power Station is limited to that for milk and crops only [RD71]. Comparisons of soil monitoring data from RIFE and activity concentrations predicted by PC-CREAM08 after 60 years of operations of the Power Station are therefore not possible. However, as for the marine environmental matrices described above, additional activity in soil resulting from radioactive gaseous releases from the Power Station would Page 197 of 255

212 be expected to exert little influence on existing radionuclide activity concentrations in the local area The highest build-up in soils is predicted to occur through the release of H-3 and C-14 to air resulting in soil activities of 9 Bq/kg and 11 Bq/kg, respectively (Table 7.22). The activities tend to agree with the levels recorded in soils from site-specific investigations by Horizon within the development area for the Power Station. For example, H-3 has been detected in soils within the range 15 to 36 Bq/kg (although there are relatively high uncertainties associated with the measurements) [RD52]. Considering the measurement uncertainties, the activity predicted by PC-CREAM 08 falls toward the lower end of this range. Site-specific measurements of C-14 in soils have been recorded in the range 3 to 11 Bq/kg by Horizon, i.e. similar to the value predicted by PC-CREAM 08 (although as with H-3, the measured values also have a high level of uncertainty). However, it is noted that the predicted levels (albeit modelled conservatively) tend to reflect typical current UK background levels from nuclear fallout and/or natural sources, and in any case, would not result in unusually high activity concentrations for an area in close proximity to a permitted nuclear installation [RD71] It is thus concluded that the build-up of activity in the environment to the levels predicted by PC-CREAM 08 from the 60-year operation of the Power Station is unlikely to cause any restriction in the future use of land or sea Assessment of Total Dose 603. The EP-RSR application guidance note requires a comparison of the potential impact of the proposed facilities against the statutory site constraint of 500 µsv/y and the annual dose limit of 1000 µsv/y. In order to facilitate comparison to the site constraint, the predicted dose to the Representative Person was aggregated with the predicted dose associated with future discharges from the adjacent Existing Power Station. This aggregate dose was then combined with the retrospective dose to representative individuals residing close to the Power Station (attributable to historical discharges) and compared to the annual dose limit. A summary of the overall methodology and assumptions, as well as the estimated site dose and total dose to a Representative Member of the public is provided below Methodology Dose from Future Discharges (i.e. Combined Discharges from the Power Station and the Existing Power Station) 604. An estimate of future dose predicted to arise due to discharges from both the Power Station and the Existing Power Station was derived by combining the dose to the Representative Person (presented in Section 7.1.6) and the dose to the most exposed individual reported in the Wylfa Magnox EIADR (Nuclear Reactors (Environmental Impact Assessment for Decommissioning) Regulations) [RD72]. The Existing Power Station is the only facility sufficiently close to the Power Station to warrant explicit consideration of its contribution to the overall dose impacts from the site It is noted that whilst the specific habits data used to estimate the doses reported in the EIADR submission for the Existing Power Station are slightly different to those used in the Page 198 of 255

213 current assessment, the overall methodology and key assumptions used in the EIADR submission are consistent with those adopted for the current assessment: For gaseous discharges, the EIADR assumes that people could be exposed through inhaling or ingesting radionuclides, or through direct irradiation from radionuclides in the air and radionuclides deposited on the ground. The candidates for the Representative Person are assumed to live at the nearest habitation to the Existing Power Station, to grow their own green vegetables, root vegetables and fruit in their garden and drink milk from a local farm. They are also assumed to spend some time near the site fence; and, For aqueous discharges to the sea, people are assumed to be exposed through the ingestion of fish, crustaceans and molluscs, the handling of fishing gear, direct exposure to radiation from activity deposited on beaches and in coastal sediments and inhalation of sea spray. Thus, the source dose reported in the EIADR submission has been adopted for this assessment without any revision The EIADR only provides a predicted dose for the adult, and so doses to child and infant have been estimated by applying the ratio of child:adult and infant:adult habit data using the farming family data from Table 7.1 and the fishing family data from Table 7.2 to calculate the doses due to gaseous and aqueous discharges, respectively. Dose from Historical Discharges 607. The dose from historical discharges was derived from the review of annual RIFE reports for the period from 2007 to 2015 [RD71]. No detailed calculations were performed for this assessment The candidates for the Representative Person for historical dose are defined within the RIFE reports as people who eat large quantities of locally harvested food (high-rate consumers) or who spend long periods of time in areas where radiation sources may exist. The methodology for determining this person is captured in Figure 7.4. Page 199 of 255

214 Figure 7.4 RIFE Total Historical Dose Methodology Total Dose to the Representative Person 609. The total dose was calculated as the sum of the predicted future dose to the Representative Person from the Power Station, and the estimated dose to the Representative Person from historical discharges, including ongoing discharges from the Existing Power Station (as published in RIFE reports) Results Dose from Future Discharges 610. Based upon the assumptions in the EIADR for the Existing Power Station [RD72], it is predicted that the highest annual dose to the candidates for the Representative Person from gaseous discharges during Care and Maintenance Preparations for the Existing Power Station would be 1.07 μsv (to an infant). Likewise, the predicted highest annual dose to the candidates for the Representative Person from aqueous discharges from the Existing Power Station is μsv (to an adult) The combined doses due to the Existing Power Station and the Power Station are summarised in Table The data on prospective dose from the new Power Station is taken from Table Page 200 of 255

215 Gaseous Aqueous Direct** Gaseous and Aqueous Direct Radiation Combined Total Gaseous and Aqueous Only Combined Total Wylfa Newydd Project - Radioactive Substances Regulation Environmental Permit Application Table 7.25 Predicted Combined Dose Age Group Prospective Annual Dose from the Existing Power Station Total Prospective Annual Dose from the New Power Station Total Combined Annual Dose (µsv) (µsv) (µsv) Adult 5.70E E E E E E E+01 Child 5.93E-01* 1.70E-02* E E E E E+01 Infant 1.07E+00* 7.00E-03* E E E E E+01 Note * Estimated based on the adult dose. ** No detailed assessment has been reported in the EIADR. However, it is stated that direct radiation will not exceed 20 µsv/y The combined total cumulative dose presented in Table 7.25 may be compared with the annual dose limit of 1,000 µsv/y and the total cumulative dose due to gaseous and liquid discharges only may be compared with the site constraint of 500 µsv/y. The predicted total cumulative doses for the Representative Person (the infant member of the farming family) are significantly below both the annual public dose limit and the site constraint. Dose from Historical Discharges 613. The dose contributions to the RIFE [RD71] Representative Person from historical discharges and sources of radiation due to operations at the Existing Power Station are summarised in Table The contributors are defined as the pathways and radionuclides which contribute to more than 10% of the total historical dose. Table 7.26 Dose Contributions to the RIFE Representative Person Year Dose (µsv/y) < Main contributor Direct radiation Direct radiation Direct radiation, milk Fish, gamma dose rate over sediment, Cs-137 Fish, gamma dose rate over sediment, Cs-137 Fish, gamma dose rate over sediment Gamma dose rate over sediment Gamma dose rate over sediment Direct radiation 614. The annual dose to local residents due to historical sources recorded in the RIFE reports from 2007 to 2015 ranged from <5 µsv/y to 13 µsv/y. The principal exposure pathway varied from one year to another, although exposure is dominated by external pathways (direct radiation and gamma dose rates above sediment). Page 201 of 255

216 Total Dose 615. The total dose, which aggregates the predicted dose to the Representative Person from the Power Station with the estimated future and historical doses described above, is presented in Table Table 7.27 Summary of the Total Doses for the Power Station Criteria Limit Annual Dose (µsv) Sum Source of Radiation for the Proposed Power Station Other Sources of Radiation from the Existing Power Station Historical Discharges Future Discharges Future Direct Radiation Historical Discharges Future Discharges Future Direct Radiation Dose limit (effective dose) 1000 µsv/y 3.77E+01 (b) 6.18E-02(c) 1.30E+01(d) 1.10E E+01 (e) 7.19E+01 Site constraint (effective dose) 500 µsv/y 3.77E+01 (b) (a) 1.10E+00 (a) 3.88E+01 Source constraint (effective dose) 300 µsv/y (max) 3.77E+01 (b) 6.18E-02 (c) 3.78E+01 Note (a) The derivation of this UK specific constraint strictly excludes consideration of future direct radiation. (b) This corresponds to the infant age group (see Table 7.18). (c) This relates to exposure of the adult (see Table 7.18). (d) The value shown is the highest total dose reported in RIFE over the past decade. (e) No detailed assessment has been reported in the EIADR. However it is stated that direct radiation will not exceed 20 µsv/y [RD72] Uncertainty and Variability within the Public Dose Assessment 616. Guidance [RD57] indicates that a semi-quantitative evaluation of uncertainty and variability should be performed to demonstrate that an appropriate level of caution has been applied to the assessment. The review [of uncertainty and variability] will ensure that sufficient caution has been retained such that the dose limit is unlikely to be exceeded on the basis of a retrospective assessment.. but balancing this against ensuring that there has not been an undue level of caution applied in the assessment Undertaking any assessment of prospective doses necessarily involves the application of models and the making of assumptions regarding future transfers, activities and human behaviour. This results in varying degrees of uncertainty. The key uncertainties associated with the dose assessments presented in this Application are: The estimate of the radioactive discharge to the environment; The dispersion of radioactivity following aqueous and gaseous discharges to the environment; The transfer of radioactivity in the environment; Assumed habits; and, Dose coefficients for the inhalation or ingestion of radioactive species. Page 202 of 255

217 618. The following are the major assumptions in the calculation of direct dose: No decay has been assumed in either the neutron or gamma ray sources for the 10 year decayed fuel in any of the 236 SFSF casks (this being the inventory of casks in the SFSF at 60 years of operation of the Power Station); Zircaloy components of the fuel assemblies were not considered in the dose modelling; Radioactive decay of Co-60 has been assumed for the sources in the HLW; No shielding is provided by the SFSF; The modelling of the LLW assumed that the sources inside it consisted of two surface sources covering the surface areas of the drums and ISO containers. These surface sources were modelled as pure Co-60 emitters. The LLW source surface strengths were normalised to the reported dose rates at the nominal site boundary (100 m) from the outer surfaces of the two sources modelled; and, The MCNP models do not consider local mounding, earthworks or terrain The degree of uncertainty of each of the above areas, and others viewed as being of lesser significance, is discussed in detail in Appendix G Conclusions on the Human Dose Assessment 620. A series of assessments have been undertaken to determine the radiological impact on members of the public resulting from the operation of two UK ABWRs at the proposed Power Station site An assessment was undertaken to determine the radiological consequences of routine gaseous and aqueous discharges on candidates for the Representative Person living close to the Power Station site. Doses were calculated for the candidates for the terrestrial Representative Person the farming family for the candidates for the marine Representative Person the fishing family and the Magnox worker. A top two consumption rate scenario was used, based upon site-specific consumption rates The predicted annual doses to a farming family due to gaseous and aqueous discharges and direct radiation during normal operations are 1.98E+01 µsv/y, 2.06E+01 µsv/y and 3.77E+01 µsv/y for an adult, child and infant, respectively. The predicted annual dose to the fishing family due to aqueous and gaseous discharges and direct radiation arising through normal operations are 1.64E+01 µsv/y, 1.70E+01 µsv/y and 2.90E+01 µsv/y for an adult, child and infant, respectively As the highest estimated annual dose due to normal operations at the Power Station has been predicted for the farming family infant, the farming family infant is therefore identified as the Representative Person The results from the prospective dose assessment are below the source constraint of 300 µsv/y for all age groups. The Representative Person is predicted to receive just 13% of the source constraint The total dose from all historic and future sources (due to the Existing Power Station and the Power Station) is 7.19E+01 µsv/y. This is 7% of the public dose limit of 1000 µsv/y. Page 203 of 255

218 626. The prospective doses due to future discharges from the Power Station and the Existing Power Station is 3.88E+01 µsv/y which is 8% of the site constraint of 500 µsv/y The prospective dose due to discharges and direct radiation from the Power Station is 3.78E+01 µsv/y which is 13% of the source constraint of 300 µsv/y As noted previously, some of the assumptions that form the basis of the dose assessment mean that the dose assessment is deliberately conservative. As such it is likely that a more realistic assessment would result in a lower estimate of doses due to gaseous discharges in particular. For example, using the design discharge height of 75 m is likely to result in a reduction of the air concentration (and thus the dose due to ingestion of C-14). Therefore there is a strong degree of confidence that sufficient caution has been retained within the calculations such that actual doses received by the Representative Person are more likely to be lower than the doses calculated within this assessment. This would be demonstrated through future retrospective (measurement based) dose assessments The predicted collective dose from the proposed Power Station is within the 1 man Sv threshold for aqueous discharges for all population groups (the threshold is set for the purpose of regulatory control rather than being a health limit ). For gaseous discharges the 1 man Sv threshold is predicted to be exceeded for EU and World populations. The collective dose for these population groups is dominated by C-14. However, per caput doses for both gaseous and aqueous discharges are in the nanosievert range which can be ignored in the decision making process as the associated risks are considered to be negligible For a single short-term release from Unit 2 due to the expected event, it is estimated that a member of the public residing at the residential location for the duration of the release will incur a dose of 1.14E-04 µsv and a Magnox worker present in the open air at the site boundary for eight hours will incur a dose of 8.69E-05 µsv. These predicted doses are negligible and in any event are significantly below the dose limit for a member of the public of 1000 µsv/y Using OPEX, the only short-term release for present consideration is that to air and arising from a fuel pin failure (as a bounding case). The resulting doses are very low as the source term is assumed to be limited to the release of noble gases only. The validity of this assumption will be underpinned by future operational experience which will also identify if other scenarios will require consideration and assessment Based upon comparison of modelling results with measured data, the build-up of activity in the environment to the levels predicted by PC-CREAM 08 from the 60 year operation of the Power Station is unlikely to cause any restrictions in the future use of land or sea. Page 204 of 255

219 7.2 Non-human Dose Assessment Guidance 633. Current recommendations from the International Commission on Radiological Protection (ICRP) [RD54] specify that during planned, existing and emergency situations, radiological impact to all of the environment needs to be considered, including areas where people are absent. The aims of environmental radiation protection are focused on preventing or reducing the frequency of radiation effects to a level where they would have a negligible impact on the maintenance of biological diversity, the conservation of species, or the health and status of natural habitats, communities and ecosystems The EA and NRW are required to ensure that they do not give any consent or permission for a plan or project which is likely to result in a significant effect on a European Site without first having undertaken an appropriate assessment of the implications for that site and found that plan or project will not adversely affect the integrity of the European Site. This duty is implemented as part of the Radioactive Substances Regulations (RSR) permitting process under the EPR16, and as part of the consenting process under the Planning Act 2008 (as amended) The EA and NRW therefore require an assessment of the likely combined impact of radioactive discharges from all relevant existing and prospective sites on non-human biota as part of the application for an EP-RSR [RD16]. The results of such assessments may be compared to a guideline value of 40 μgy/h, the threshold below which the regulators consider there will be no adverse effect on non-human biota [RD16] or the integrity of protected sites (such as Special Areas of Conservation (SACs) and Special Protection Areas (SPAs)) [RD55], [RD56] In assessing radiological impacts of discharges and on-site disposals to non-human biota NRW guidance [RD16] requires that: Worst-case dose rates should be calculated by assuming the presence of the reference organisms for the relevant ecosystem at the position of maximum environmental concentration due to discharges (usually close to the site boundary for the terrestrial ecosystem, and close to the point of discharge for aquatic ecosystems); Information be provided on the model used to calculate the dose rates and why it is appropriate; anfd, All the data and assumptions (with reasoning) used as input into the model be set out where not already covered by the human dose assessment Methodology and Data Inputs Assessment Approach 637. The assessment of radiological impacts due to discharges from the Power Station and the Existing Power Station on non-human biota was based on the ERICA Integrated Approach, which comprises the ERICA tool and the associated FREDERICA database [RD82] [RD83]. The ERICA approach does not facilitate the assessment of radiological impacts arising from releases of noble gases and it was therefore supplemented with the EA s R&D128 methodology [RD56] which facilitates such assessment. ERICA and R&D128 are described below. Page 205 of 255

220 ERICA 638. ERICA (Environmental Risks from Ionising Contaminants: assessment and management) is a software tool that is used to assess the radiological risk to terrestrial, freshwater and marine biota. It has a proven history of use in previous permit applications for new nuclear build, including new nuclear power plant and other facilities in the UK. It has also been used to support other applications, e.g. for submissions made under the EIADR [RD63] [RD84]. The tool and supporting documentation are available on the Centre for Ecology and Hydrology website ( For this assessment the latest version of ERICA was used (v1.2.1, released in February 2016) The ERICA tool [RD84] calculates dose rates to organisms by applying dose conversion coefficients 29 to the concentrations of radionuclides in environmental media or in biota. A range of dose conversion coefficients for internal and external exposures has been calculated for reference organisms-radionuclides-radiation type combinations and are stored in databases embedded within the ERICA tool. The reference organisms used within ERICA complement the Reference Animals and Plants (RAPs) proposed by ICRP [RD85]. It uses some of the plant and animal geometries currently outlined within the RAPs, but is broader in range For the purpose of the current assessments, the more advanced dispersion models within PC-CREAM 08 were used to model environmental dispersion of radioactive releases and calculate activity concentrations in the assessed habitats. However, PC-CREAM 08 does not contain a model for assessing radionuclide dispersion in lakes. Therefore the IAEA Safety Report Series (SRS) 19 model [RD88] incorporated within the ERICA tool was used for this purpose. The SRS 19 models are suitable for undertaking simple, conservative screening calculations of environmental activity concentrations and can enable estimates of the dispersion of radionuclides in small lakes (<400 km 2 ) and large lakes (>400 km 2 ) The ERICA assessment tool consists of three tiers: Tier 1: an initial screening assessment by means of which, if the pass criteria are met (dose rate screening value), the user can exit the assessment process; Tier 2: where more site-specific parameters can be used; and, Tier 3: which consists of a probabilistic risk assessment used when the screening dose criteria are exceeded at Tier 1 and Tier For Tier 1, the predefined screening dose rate is back-calculated to yield Environmental Media Concentration Limits (EMCLs) for all reference organism/radionuclide combinations. The tool compares the input media concentrations with the most restrictive EMCL for each radionuclide and determines a Risk Quotient (RQ) The RQ is the ratio of the predicted environmental dose rate to a benchmark dose rate that is assumed to be environmentally safe [RD83]. Within ERICA, the benchmark dose rate is set as 10 µgy/h. This value is therefore typically adopted as the ERICA screening value [RD83]. If the RQ is less than one, then the tool suggests that the user should exit the assessment process. If the RQ is greater than one, the user is advised to continue with the assessment. Although the RQ approach might be deemed overly conservative, it is 29 Dose conversion coefficients are defined as the absorbed dose rate (μgy/h) per unit activity concentration in an organism (Bq/kg fresh weight) or environmental media (Bq/kg or Bq/l media fresh weight). Details of the derivation of the coefficients and the calculation of internal and external dose rates can be found in [RD83], [RD86] and [RD87]. Page 206 of 255

221 reasonably consistent with other assessment approaches currently available, e.g, [RD56], and reflects the uncertainty associated with the severe lack of data for some radionuclidereference organism combinations Tier 2 allows the user to be more interactive, i.e. to change the default parameters, and facilitates the addition of non-default radionuclides and new reference organisms. The evaluation is performed directly against the screening dose rate, with the dose rate and RQs generated for each reference organism selected for assessment. Tier 2 also facilitates the inclusion of Uncertainty Factors (UF) to account for uncertainties in the dose rate calculations and estimates the probability of exceeding the screening dose rates. The UF is multiplied by the RQ to obtain a conservative RQ value The methodology used to assess impact is based on a set of reference organisms to which most species can be assigned. Input from field ecological surveys and investigations is used to identify any species in the area surrounding the Power Station site that cannot be assigned to the reference organisms included in ERICA. If such species are identified then the ERICA tool can be used to define representations of those species to enable an assessment of the impact to be made The current assessment was undertaken at Tier 2 using a user-defined screening dose rate of 40 μgy/h corresponding to the guideline dose rate at which NRW will make its determination of the assessment [RD16] and a UF of 5 (corresponding to the 99.0 th percentile values) [RD89], and default radiation weighting factors for alpha of 10.0, beta/gamma of 1.0, and low energy beta of 3.0. This setup permits the conservative quantification of the radiological impact on each of the reference organisms. R&D The ERICA tool does not currently facilitate the assessment of dose impacts from noble gases. The radiological impact due to the discharge of these radionuclides has therefore been calculated using the methodology described in R&D Publication 128 Impact Assessment of Ionising Radiation on Wildlife version 2 [RD90] R&D 128 is used by EA regulators and its methodology is recognised within the EA s guidance [RD55] on applying the Conservation of Habitats and Species Regulations 2010 to radioactive substances activities regulated under EPR16. It describes the behaviour and transport of radionuclides in terrestrial, freshwater and coastal habitats much like ERICA and the associated spreadsheets calculate dose per unit concentration for a range of radionuclides and species The dose to non-human biota from the noble gases Ar-41, Kr-85, Kr-88, Xe-131m and Xe- 133 was modelled using R&D 128 spreadsheet version 2.0 for the terrestrial environment [RD90]. This version has been updated to include a larger number of noble gas radionuclides and the representative animals and plants considered in ERICA. The approach for assessment of noble gases used for this work is consistent with that published in [RD91]. Page 207 of 255

222 Site Layout and Surrounding Habitats 650. The habitat types and species of European Designated Sites in the vicinity of the Power Station are identified in Table 7.28 along with their marine, terrestrial and freshwater qualifying features. Within the table feature organisms have also been mapped against the generic reference habitats and species used within the ERICA model. Please note that it is not the intention within Table 7.28 to list all of the European Designated Sites that potentially may be affected by radioactive discharges, but rather to capture the range of marine, terrestrial and freshwater environments for which the ERICA modelling has been undertaken 651. The most notable of the sites subject to ecological conservation designations includes Cemlyn Bay, which forms part of the Morwenoliaid Ynys Môn / Anglesey Terns SPA, the Cemlyn Bay SAC to the west of the proposed Power Station site, and the Gogledd Mon Forol / North Anglesey Marine csac. This latter site is a candidate SAC for harbour porpoise. It includes an area of approximately 325,000 hectares around the northern half of Anglesey [RD81], and its boundary includes water that is within and adjacent to the Wylfa Newydd Development Area and partially within the Power Station site 652. Three distinct habitat types, generally representative of the designated sites (European sites or otherwise) have been identified as being potentially sensitive to gaseous and aqueous radioactive effluent released from the Power Station on account of their ecological significance and their proximity. These are: A terrestrial habitat, which is assumed to lie on the site boundary immediately to the east of the site, just outside the Power Station fence. This habitat is the area where deposited activity from gaseous releases to the atmosphere is predicted to be greatest; A marine habitat, in the coastal waters to the north and west of the Power Station site. This habitat is analogous to the 100 km 2 local (Wylfa) marine compartment within the DORIS marine dispersion module of PC-CREAM 08 and and represents the Anglesey Terns SPA, and the Gogledd Mon Forol / North Anglesey Marine csac; and, A freshwater habitat, assumed to consist of a small lake which represents Tre r Gof SSSI to the north of the Power Station. Tre r Gof SSSI is situated in the predominant wind direction and its catchment includes the area receiving the highest deposition rates for gaseous radionuclides released from the Power Station. The SSSI is therefore considered to be more limiting than the other important freshwater habitat Cae Gwyn, which is situated to the south of the Power Station site. For the purposes of this assessment the location of the freshwater habitat is based on the food production area used in the human dose assessment Llanbadrig Dinas Gynfor has been designated as a SSSI on account of its geological significance. As it is not a European site it is not relevant to the assessment of radiation impact on non-human biota A fourth important habitat, a brackish lagoon designated as the Cemlyn Bay SAC was also considered. However, current assessment methodologies do not facilitate direct assessment of radiological impacts to brackish habitats. In addition, given that the Cemlyn Bay SAC is fed by seawater (from the marine habitat identified above) and freshwater from local watercourses, it is considered that activity concentration in the brackish water and sediment within this habitat is less than and bounded by the activity concentrations in the marine habitat. Page 208 of 255

223 Table 7.28 Local European Designated Sites with Marine, Terrestrial and Freshwater Qualifying Features European Site Designation Distance from WNDA Boundary Qualifying Features Proxy Habitat and Reference Species within the ERICA Modelling Environment to which the proxy habitat and species are assigned in the ERICA modelling (km) Gogledd Môn Forol / North Anglesey Marine csac Within Harbour porpoise (Phocoena phocoena) Mammal Marine Morwenoliaid Ynys Môn / Anglesey Terns SPA Within Arctic tern (Sterna paradisaea) Bird Marine and terrestrial Sandwich tern (Sterna sandvicensis) Roseate tern (Sterna dougallii) Common tern (Sterna hirundo) Bae Cemlyn / Cemlyn Bay SAC 0.1 Coastal lagoon (this is a priority habitat feature) Perennial vegetation of stony banks Various species characteristic of marine habitat Vascular plant Shrub/grasses and herbs Marine Marine and terrestrial Glannau Ynys Gybi/Holy Island Coast SAC 13.1 Vegetated sea cliffs of the Atlantic and Baltic coasts Shrub/grasses and herbs Terrestrial European dry heaths Northern Atlantic wet heaths with Erica tetralix Corsydd Môn/Anglesey Fens SAC 14 Northern Atlantic wet heaths with Erica tetralix Molinia meadows on calcareous, peaty or clayey-silt-laden soils Calcareous fens with Cladium mariscus and species of the Caricion davallianae Shrub/grasses and herbs/tree Terrestrial Page 209 of 255

224 European Site Designation Distance from WNDA Boundary Qualifying Features Proxy Habitat and Reference Species within the ERICA Modelling Environment to which the proxy habitat and species are assigned in the ERICA modelling (km) Alkaline fens Hard oligo-mesotrophic waters with benthic vegetation Southern damselfly (Coenagrion mercuriale) Arthropod - detritivorous/flying insect/insect larvae Terrestrial and freshwater Marsh fritillary butterfly (Eurodryas, Hypodryas) Arthropod - detritivorous/flying insects Terrestrial Geyer`s whorl snail (Vertigo geyeri) Gastropod Llyn Dinam SAC 14.3 Natural eutrophic lakes with Magnopotamion or Hydrocharitiontype vegetation Various species characteristic of freshwater habitat Freshwater Page 210 of 255

225 Comparison of ERICA Default Organisms with Feature Species and Other Nonhuman Biota within the Area 655. The range of plant and animal organisms inhabiting the environments around the Power Station site was identified following a series of ecological surveys undertaken in support of the various permit and planning applications [RD93]. These organisms were then compared to the ERICA default reference organisms which were found to be broadly representative of the local species The ecological surveys confirmed the presence of a number of bat species at several locations around the Power Station site. Bats are of conservation significance and the potential impact of discharges from the site on these organisms needs to be considered. Bats are not currently included as default organisms within the ERICA tool. The bat was therefore added to the database as a new organism, and modelled using default radioecology parameters for a small burrowing animal In common with the modelling of other flying animals (including birds) within ERICA, exposure is assumed to occur through contact with contaminated soil rather than contaminated air. For the purposes of this assessment, the organism bat is assumed to wholly reside on soil, thus representing a cautious assumption for organisms that spend most of their time flying or roosting at height above ground. However, the approach is in keeping with that adopted for birds Marine organisms (sponges, hydroids, bryozoans, sea squirts, etc.) have been identified by marine ecologists as being integral to the rocky reef habitats surrounding the cooling water outfall. These organisms are not present on the ERICA reference organisms list. It was considered though that the identified organisms are closely related to some of the default ERICA organisms (such as the sea anemones and corals, and polychaete worms) and that given the limitations of the ERICA tool, the use of the default organisms would be adequate A number of coastal areas around Wales are currently being considered for possible designation as new SACs and SPAs including the coastal waters around the Power Station site which are currently under consideration as a possible SAC for the harbour porpoise [RD94]. Porpoises are analogous to the ERICA default marine organism mammal and are therefore included in the assessment of non-human biota Calculation of Radionuclide Concentration in the Environment 660. In order to carry out an assessment of impacts on non-human biota, it was first necessary to determine the activity concentration of discharged radionuclides in water and soil/sediment. Page 211 of 255

226 Aqueous and Gaseous Releases from the Power Station 661. Discharges of aqueous radionuclides were assumed to be made into the local compartment of the Irish Sea via outfall structures currently used by the Existing Power Station as this is the setup already defined within PC-CREAM 08 for the Existing Power Station site. Releases of gaseous radionuclides into the atmosphere will be via the two 75 m high R/B stacks. However, for the purposes of this assessment the modelling has assumed that all gaseous discharges occur from a two stacks each with an effective height of 15 m. The proposed annual discharges to the environment from the Power Station are presented in Section 5 and are the same as those used for the human dose assessment The modelling of activity concentrations in environmental media using PC-CREAM 08 is described earlier in Section The calculated radionuclide concentrations in environmental media are presented in Table A.99 to Table A.101 of Appendix P. Aqueous and Gaseous Releases from the Existing Power Station 663. Gaseous and aqueous discharges are still made from the Existing Power Station despite it having ceased power generation at the end of December The discharges are not expected to increase throughout Care and Maintenance Preparations, a period of about ten years following cessation of power generation [RD72]. Therefore, for the purposes of this assessment it has been conservatively assumed that the discharges to the environment are based on the current authorised limits [RD95] (see Table 7.29). Table 7.29 Authorised Limits for the Existing Power Station Radionuclide Authorised Limit (Bq/y) Aqueous Tritium Other radionuclides 1.50E E+11 Gaseous Particulate beta Tritium (H-3) Carbon-14 Sulphur-35 Argon E E E E E For the purpose of this assessment Ru-106 was allocated as the representative radionuclide for other radionuclides that make up part of the aqueous discharge. Caesium-137 was allocated as the surrogate radionuclide for beta particulate which forms part of the gaseous discharge. These radionuclides were chosen as they have the largest dose per unit release value for beta/gamma emitters included in the Initial Radiological Assessment Tool (IRAT) spreadsheet tools [RD96]. Page 212 of 255

227 Calculation of Radionuclide Dispersion and Activity Concentration in the Environment Marine Habitat 665. The dispersion and build-up of radionuclides in the local marine environment from radionuclides entrained in aqueous discharges from the Power Station were modelled using the DORIS module of PC-CREAM 08 [RD64]. DORIS calculates the timedependent activity concentration of radionuclides in the local and regional marine compartments. The local marine compartment (the Wylfa local compartment) is modelled as a single well-mixed body of water and associated sediment processes, extending a few kilometres (typically 10 km) along the shoreline and outward into the sea. The local compartment is contained within the larger regional compartment (Irish Sea west) with which it exchanges water and suspended sediment [RD64]. Discharges from the Existing Power Station are assumed to be released into the same marine environment (Wylfa local compartment) The DORIS parameter values used to model the dispersion of radionuclides in aqueous effluent discharged from the Power Station into the marine environment (Wylfa local compartment) are presented in Section Terrestrial Habitat 667. The dispersion and concentration in air of radionuclides originating from gaseous effluent discharged from the Power Station and the subsequent deposition and build-up in soils were modelled with the PLUME and FARMLAND modules of PC-CREAM 08 [RD64] using an effective release height of 15 m. Details of the modelling of atmospheric dispersion are given in Section The determination of build-up of radionuclides in the terrestrial and marine environment due to gaseous and aqueous discharges are described in Section Freshwater Habitat 668. The accumulation of radionuclides in Tre r Gof SSSI (a freshwater habitat) from deposition of gaseous releases was calculated using the IAEA SRS-19 model for a small lake [RD88]. The SRS-19 model takes account of both direct deposition of radionuclides into the lake and indirect contribution due to runoff and washout of radionuclides deposited within the lake catchment. The model assumes that the catchment is 100 times the lake surface area, and that 2% of radionuclides deposited on to the catchment reaches the waterbody [RD85]. The SSSI has an area of 10.1 hectares [RD97] so the default modelled catchment area of around 1000 hectares is likely to be conservative The deposition rates derived for the food production location were conservatively adopted for assessing the radiological impacts to the generic freshwater lake habitat [RD65]. The parameter values used to model the concentration of radionuclides in the Tre r Gof freshwater habitat are presented in Table Page 213 of 255

228 Table 7.30 SRS-19 Parameter Values for Small Lake Parameter Value [RD98] Catchment area 0.92 km 2 Lake surface area 9,200 m 2 Flow rate m 3 /s Lake depth 0.3 m Lake volume 2,760 m 3 Discharge duration 60 years 670. Some parameters used for the assessment of non-human biota are not precisely known. In particular the lake depth and flow rates are subject to some degree of uncertainty. In order to determine the sensitivity of the assessment to these two parameters a sensitivity study was undertaken. Three additional cases were assessed: A lake depth of 0.3 m and a zero flow rate; A reduced lake depth of 1 cm and a flow rate of m 3 /s; and, A reduced lake depth of 1 cm and a zero flow rate Representative Organisms Assessed 671. The default ERICA organisms were adopted for each of the three habitats assessed as being representative of European sites relevant to the Power Station (Table 7.31). The bat was added to the list of terrestrial organisms assessed, as previously described. Further details regarding the selection of occupancy factors and concentration ratios (CR) for each of the organisms assessed are given in Appendix P. Table 7.31 Reference Organisms Modelled for each Proxy Habitat Habitat / Organism Terrestrial Marine Freshwater Amphibian Benthic fish Amphibian Annelid Bird Benthic fish Arthropod detritivorous Crustacean Bird Bird Macroalgae Crustacean Flying insects Mammal Insect larvae Grasses & Herbs Mollusc bivalve Mammal Lichen & Bryophytes Pelagic fish Mollusc bivalve Mammal large Phytoplankton Mollusc gastropod Mammal small burrowing Polychaete worm Pelagic fish Mollusc gastropod Reptile Phytoplankton Reptile Sea anemones & True coral Reptile Shrub Vascular Plant Vascular plant Page 214 of 255

229 Habitat / Organism Terrestrial Marine Freshwater Tree Zooplankon Zooplankton Bat* Note. * Bats are not included in the standard set of reference organisms and plants within ERICA Results Terrestrial Habitat 672. The risk quotients for terrestrial organisms at the terrestrial habitat are presented in Table 7.32 for each of the default ERICA organisms based on a user-defined screening value of 40 µgy/h. The contribution from noble gases, as determined using R&D128 [RD90], is presented in Table Detailed results are presented in Appendix Q Table A.102. It should be noted that R&D 128 does not permit the inclusion of user-defined models. Therefore the dose to bats cannot be determined using R&D 128. However, if bats are assumed to be represented by rodents the estimated dose is 9.46E-05 µgy/h. The total dose rates to terrestrial organisms from combining the ERICA and R&D128 results are presented in Table Page 215 of 255

230 Table 7.32 Terrestrial Habitat ERICA Results (Dose Rate and Risk Quotient) Organism Total Dose Rate per Organism Screening Value Risk Quotient (No UF applied) Risk Quotient (UF applied) (µgy/h) (µgy/h) (unitless) (unitless) Amphibian 5.96E E E-02 Annelid 2.01E E E-02 Arthropod detritivorous 2.02E E E-02 Bird 6.17E E E-02 Flying insects 2.00E E E-02 Grasses & Herbs 4.01E E E-02 Lichen & Bryophytes 4.05E E E-02 Mammal large 6.17E E E-02 Mammal small burrowing 6.17E E E-02 Mollusc gastropod 2.01E E E-02 Reptile 6.17E E E-02 Shrub 4.01E E E-02 Tree 5.99E E E-02 Bat 6.10E E E-02 Table 7.33 Terrestrial Habitat R&D128 Results (Dose Rate and Risk Quotient) Organism Total Dose Rate per Organism Screening Value Risk Quotient (No UF applied) Risk Quotient (UF applied) (µgy/h) (µgy/h) (unitless) (unitless) Amphibian 2.30E E E-05 Annelid 5.91E E E-08 Arthropod detritivorous 2.61E E E-05 Bird 2.14E E E-05 Flying insects 6.29E E E-09 Grasses & Herbs 2.53E E E-05 Lichen & Bryophytes 4.91E E E-05 Mammal large 2.76E E E-05 Mammal small burrowing 1.22E E E-05 Mollusc gastropod 5.28E E E-09 Reptile 2.45E E E-05 Shrub 2.30E E E-05 Tree 3.14E E E-05 Bat 9.46E E E-05 Page 216 of 255

231 Table 7.34 Combined (Total) Terrestrial Habitat Results (Dose Rate and Risk Quotient) Organism Total Dose Rate per Organism Screening Value Risk Quotient (No UF applied) Risk Quotient (UF applied) (µgy/h) (µgy/h) (unitless) (unitless) Amphibian 5.97E E E-02 Annelid 2.01E E E-02 Arthropod detritivorous 2.02E E E-02 Bird 6.17E E E-02 Flying insects 2.00E E E-02 Grasses & Herbs 4.01E E E-02 Lichen & Bryophytes 4.06E E E-02 Mammal large 6.17E E E-02 Mammal small burrowing 6.17E E E-02 Mollusc gastropod 2.01E E E-02 Reptile 6.17E E E-02 Shrub 4.01E E E-02 Tree 5.99E E E-02 Bat 6.10E E E It can be seen from Table 7.34 that the RQ values for all of the organisms, both with and without the UF applied, are at worst an order of magnitude below 1.0. The dose rates are therefore considered to be of no significance. The organisms that are predicted to incur the highest dose rate and therefore have the highest RQ are birds, large mammals, small burrowing mammals and reptiles. The radionuclides that make the greatest contribution to the dose rate are C-14 (55%) and H-3 (42%) Marine Habitat 674. The predicted dose rates and risk quotients for organisms within the marine habitat are presented in Table 7.35 for each of the default ERICA organisms based on a user-defined screening value of 40 µgy/h. Detailed results are presented in Appendix Q Table A.103. Page 217 of 255

232 Table 7.35 Marine Habitat ERICA Results (Dose Rate and Risk Quotient) Organism Total Dose Rate per Organism Screening Value Risk Quotient (No UF applied) Risk Quotient (UF applied) (µgy/h) (µgy/h) (unitless) (unitless) Benthic fish 4.96E E E-08 Bird 5.39E E E-08 Crustacean 9.70E E E-07 Macroalgae 4.62E E E-08 Mammal 5.01E E E-06 Mollusc bivalve 1.22E E E-07 Pelagic fish 4.77E E E-08 Phytoplankton 9.82E E E-07 Polychaete worm 6.29E E E-08 Reptile 5.76E E E-08 Sea anemones & True coral 7.51E E E-08 Vascular Plant 5.10E E E-08 Zooplankton 1.45E E E The RQ values for all of the organisms, both with and without the UF applied, are many orders of magnitude below 1.0. The dose rates are therefore considered to be of no significance. The organism that is predicted to incur the highest dose rate and therefore have the highest RQ is the marine mammal. The radionuclides that make the greatest contribution to the dose rate incurred by the marine mammal are Fe-55 (90%), Fe-59 (9%) and H-3 (1%) Freshwater Habitat 676. The calculated dose rates and risk quotients for the initial assessment for organisms in the freshwater habitat for a lake depth of 0.3 m and a flow rate of m 3 /s are presented in Table 7.36 for each of the default ERICA organisms based on a user-defined screening value of 40 µgy/h. Detailed results are presented in Appendix Q Table A.104. Page 218 of 255

233 Table 7.36 Freshwater Habitat ERICA Results (Dose Rate and Risk Quotient) Organism Total Dose Rate per Organism Screening Value Risk Quotient (No UF applied) Risk Quotient (UF applied) (µgy/h) (µgy/h) (unitless) (unitless) Amphibian 1.69E E E-05 Benthic fish 1.51E E E-03 Bird 1.35E E E-05 Crustacean 2.02E E E-03 Insect larvae 3.90E E E-03 Mammal 1.83E E E-05 Mollusc bivalve 1.67E E E-03 Mollusc gastropod 1.75E E E-03 Pelagic fish 1.75E E E-05 Phytoplankton 1.30E E E-05 Reptile 1.51E E E-03 Vascular plant 1.99E E E-03 Zooplankton 1.33E E E Once again, the RQ values for all of the organisms, both with and without the UF applied, are orders of magnitude below 1.0. The dose rates are therefore considered to be of no significance. The organism that is predicted to incur the highest dose rate and therefore have the highest RQ is insect larvae. The radionuclides that make the greatest contribution to the dose rate incurred by the insect larvae are I-131 (70%), I-133 (13%) and I-132 (10%) Combined Impacts 678. Combined impacts to non-human biota were assessed in two regards: Doses arising from the combined discharges from both the Power Station and the Existing Power Station; and, Doses arising from the combined discharges from the Power Station and all other relevant permitted nuclear and non-nuclear sites within the UK. This assessment is applicable for the Habitats Regulation Assessment (HRA) screening see Appendix R. These assessments are described below. Combined Impacts from the Power Station and the Existing Power Station 679. The assessment of the radiological impact on non-human biota of discharges from the Existing Power Station was made using the IRAT tool version 2 [RD79] using the current authorised annual discharges (see Table 7.29) given in the EIADR for the Existing Power Station site [RD102], and the following assumptions: For a release into an estuary/coastal environment a coastal exchange rate of 995 m 3 /s was defined. This is the value given for Cemaes coast in the IRAT spreadsheet tool; and, Page 219 of 255

234 For a release to air, a ground-level release was defined. This will result in a conservative estimate of the consequences to terrestrial wildlife The assessment of dose to freshwater organisms from the Existing Power Station was also made using the IRAT tool for terrestrial organisms albeit excluding C-14 and Ar-41 as these are non-depositing radionuclides. The results for terrestrial and freshwater habitats are presented in Table 7.32 and Table 7.36 respectively. The RQ for terrestrial organisms is at least an order of magnitude higher than that for the freshwater habitat so this approach would give an upper estimate of the dose to freshwater organisms The estimated dose rates to the worst affected wildlife groups due to discharges at the current annual limits from the Existing Power Station were: Terrestrial wildlife Marine wildlife Freshwater wildlife 1.20 µgy/h; 0.59 µgy/h; and, 0.15 µgy/h Adding these doses to those received by the most exposed organisms from discharges from the proposed Power Station gives the following combined dose rates for impacts from both stations: Terrestrial birds, mammals and reptiles Marine mammals Freshwater insect larvae 1.82 µgy/h; 0.59 µgy/h; and, 0.19 µgy/h The proposed Power Station s contribution to the predicted combined dose rates provided above are lower compared to those from the Existing Power Station, and when added together, are far below the guideline threshold of 40 µgy/h. Combined Impacts from the Power Station and Other Sites with Permitted Discharges of Radioactive Substances 684. Unlike dose assessments for members of the public there is no site constraint applied to estimations of non-human biota doses. However, it is worth conducting a wider site assessment by considering the two adjacent stations and comparing the findings with results generated for discharges from a larger number of permitted installations The total dose rate for radioactive substances within each of the European sites included in Table 7.37 was determined using the EA s own habitats assessment for radioactive substances [RD55]. The EA s assessments published in [RD55] were used to calculate dose rates to reference organisms in coastal, freshwater and terrestrial environments, taking into account the combined radiological impact of existing discharges from multiple permitted releases around the UK and cautiously assuming that those discharges occurred at the permitted limits in force at the time. All releases with the potential to have an impact on the Natura 2000 sites were included in the assessment, including those from the Existing Power Station. The calculated total dose rates were compared to the guideline threshold of 40 µgy/h. Page 220 of 255

235 686. The EA s method for determining the total dose rate at each Natura 2000 site is by summing the highest predicted terrestrial dose rate (for the worst affected organism) and the maximum water dose rate (i.e. the maximum of the freshwater dose rate and the coastal water dose rate), again for the worst affected organism The highest calculated dose rates arising for the most exposed reference organisms in the terrestrial and water habitats due to future discharges from the proposed Power Station were selected from the results provided in Table 7.34, Table 7.35 and Table These values were then added to the total dose rate reported by the EA in [RD55] for each European site to give the final estimated combined impact from the Power Station with other installations having an EP-RSR The in-combination total dose rates due to discharges to air and water at proposed limits for the Power Station and at permitted limits for all other relevant sources are provided in Table It can be seen that the 40 μgy/h guideline dose rate is not anticipated to be exceeded for any of the European sites relevant to this Application. Page 221 of 255

236 Table 7.37 In-combination Total Dose Rates for European Designated Sites in the Vicinity of the Power Station European site Designation Total Dose Rate (µgy/h) [RD55] [A] Dose Rate for Most Exposed Organism in Terrestrial Environment (Proposed Power Station) [B] (3) Dose Rate for Most Exposed Organism in Water Environment (Proposed Power Station) [C] (4) In-combination Total Dose Rate [A+B+C] Bae Cemlyn / Cemlyn Bay SAC 8.8E E E E-01 Corsydd Môn/Anglesey Fens SAC 6.7E E E E+00 Glannau Ynys Gybi/Holy Island Coast SAC 1.1E E E E-01 Llyn Dinam SAC 3.4E E E E-01 Gogledd Môn Forol / North Anglesey Marine 2 psac 8.8E E E E-01 Morwenoliaid Ynys Môn / Anglesey Terns 1 pspa 8.8E E E E-01 Notes. 1. As this is a recently fully designated SPA, this site has not been assessed by the EA [RD55]. Therefore the data for Bae Cemlyn / Cemlyn Bay SAC has been used as a proxy in this case given its location and the fact that this site has the highest coastal dose rate of those European sites included in this assessment (i.e. the approach is conservative). 2. As this is a csac, this site has not yet been assessed by the EA [RD55]. Therefore the data for Bae Cemlyn / Cemlyn Bay SAC has been used as a proxy in this case as this site has the highest coastal dose rate of those European sites included in this assessment (i.e. the approach is conservative). 3. Worst case results are for bird, large mammal, small mammal (burrowing), reptile. 4. The worst case result from a fresh/marine water habitat applies to insect larvae in the freshwater environment. Page 222 of 255

237 7.2.4 Results and Conclusions 689. The reference terrestrial wildlife organisms that incur the highest dose rate due to gaseous discharges at the proposed limits from the Power Station are birds, mammals (large), mammals (small, burrowing) and reptiles. Each are predicted to incur a dose rate of 0.62 µgy/h. The expected Risk Quotient (RQ) using a threshold dose rate of 40 µgy/h is while the corresponding conservative RQ, using an Uncertainty Factor (UF) of five is No reptiles or terrestrial mammals are identified as qualifying features in the European sites assessed for the Power Station. The only terrestrial bird species of interest is the red billed chough. However, the dose rate to the chough (from gaseous discharges from the Power Station only) is predicted to be less than 1 µgy/h at the assessed distance of just 550 m from the Power Station (in an area predicted by PC-CREAM 08 to receive the maximum ground level air concentration of radionuclides released in gaseous effluents). The dose rate is far below the 10 µgy/h screening threshold within ERICA (the value used to determine if further assessment is required) If gaseous discharges from the Existing Power Station are made at the current permitted limit then the worst affected terrestrial wildlife is predicted to incur a dose rate of 1.20 µgy/h from a normal operational continuous release. The corresponding expected and conservative RQs are 0.03 and 0.15 respectively The combined effect of discharges from the Power Station and the Existing Power Station occurring at the relevant limits is an overall exposure of 1.82 µgy/h. The corresponding expected and conservative RQs are and 0.23 respectively. A dose rate of this low magnitude would again not be expected to be detrimental to populations of red billed chough European sites identified as predominantly terrestrial in nature are Corsydd Môn/Anglesey Fens SAC and Glannau Ynys Gybi/Holy Island Coast SAC. Worst case terrestrial and freshwater dose rates calculated by a combination of ERICA and R&D 128, when applied to the EA habitats assessment methodology, give rise to total dose rates of between 0.66 to 7.4 µgy/h due to gaseous discharges from all sites previously assessed by the EA as well as contributions from the operation of the Power Station. The range in dose rates constitutes around 2 to 18% of the 40 µgy/h guideline suggesting there is no significant risk of harm from radioactive discharges at terrestrial habitats within European sites The radionuclide that makes the largest contribution to the dose rate incurred by terrestrial organisms is C-14, which contributes about 98% of the dose to the most exposed organisms due to the discharges from the Power Station. C-14 and Ar-41 gaseous discharges from the Existing Power Station contribute about 46% and 45% of the dose rate to terrestrial organisms respectively The marine mammal is the reference organism predicted to receive the highest dose rate from aquatic discharges from the Power Station. However, the harbour porpoise is estimated to receive a dose rate of just 5.0E-05 µgy/h which is radiologically insignificant. The expected RQ using a threshold dose rate of 40 µgy/h is 1.3E-06. The corresponding conservative RQ, using a UF of five is 6.3E If aqueous discharges from the Existing Power Station are made at the current authorised limit then the worst affected marine wildlife incur a dose rate of 0.59 µgy/h. The corresponding expected and conservative RQs are and respectively. Page 223 of 255

238 697. A combined exposure of 0.59 µgy/h is therefore predicted if discharges from the Power Station and the Existing Power Station occur at the authorised limits. The corresponding expected and conservative RQs are and respectively and the dose rate to the marine mammal is of little significance even if the assessment has been performed for the local marine compartment, i.e. the compartment that contains the psac at Gogledd Môn Forol / North Anglesey Marine csac The radionuclides that make the largest contribution to the dose rate incurred by marine mammals due to liquid discharges from the Power Station are Fe-55 and Fe-59. These two radionuclides contribute about 90% and 9% of the dose rate due to liquid discharges from the Power Station respectively. The dose rate due to liquid discharges from the Existing Power Station are due almost entirely to other radionuclides (modelled as Ru-106) For the freshwater habitat the organism that incurs the highest dose rate due to gaseous discharges from the Power Station is insect larvae. This organism incurs a dose rate of µgy/h. The expected RQ for a threshold dose rate of 40 µgy/h is The corresponding conservative RQ, using a UF of five is The IRAT spreadsheet tool includes a freshwater model, although this is used for modelling discharges to a river rather than a lake. However, it is considered that the terrestrial model will give an indication of the likely dose to freshwater organisms if only depositing radionuclides are considered (i.e. if C-14 and Ar-41 discharges are set to zero in the spreadsheet tool). On this basis the gaseous discharges from the Existing Power Station made at the current authorised limit result in the worst affected freshwater wildlife incurring a dose rate of 0.15 µgy/h. The corresponding expected and conservative RQs are and 0.019, respectively The combined effect is an overall exposure to freshwater organisms of 0.19 µgy/h if discharges from the Power Station and the Existing Power Station occur at the authorised limits. The corresponding expected and conservative RQs are and respectively The in-combination impact of sites already assessed by the EA plus dose rates arising from discharges to air at the Power Station produces total dose rates for freshwater dominated European sites (Llyn Dinam and Corsydd Môn/Anglesey Fens) (and also the fen-type system at Tre r Gof SSSI) of 0.7 to 7.4 µgy/h. These values amount to 2 to 18% of the guideline dose rate The freshwater assessment for the Power Station was based on a site location for the lake that was assumed to be 555 m from the Reference Stack (i.e. at the point of maximum deposition. In reality Tre r Gof is located at a distance of 1130 m from the reference stack). The ERICA assessment used various other conservative assumptions, as did the methodology giving rise to the freshwater habitat total dose rates published by the EA. Therefore, it is highly likely that any freshwater habitats named above, including the qualifying species within them that are further away than 555 m, would receive an even smaller proportion of the guideline dose rate than has been estimated in this study The radionuclide discharges from the Power Station that make the largest contribution to the dose rate incurred by freshwater organisms are the radioiodines. I-131 contributes about 57% and the total radioiodines contribute about 99% of the dose rate incurred by the gaseous discharges from the Power Station. H-3 and particulate Beta gaseous discharges from the Existing Power Station contribute about 71% and 25% of the dose rate to freshwater wildlife respectively. A sensitivity study has been undertaken of the influence Page 224 of 255

239 of lake depth and flow rate on freshwater non-human biota and it is found that the expected RQs remain below one over the range of parameters studied In all habitats the conservative RQs due to discharges at the proposed limits from the Power Station are less than 0.1, with the exception of the conservative RQ for the extreme freshwater case which considered a lake depth of 1 cm and a zero flow rate, where all freshwater biota are predicted to have a conservative RQ of over 1.0. This is considered to be a very unlikely scenario due to the conservatisms used, and therefore not of concern. The combined impacts from discharges from the Power Station and the Existing Power Station are all less than 1.0. As the RQs are all less than one this indicates that the discharges from the Power Station will not pose a risk to non-human biota. Similarly, the combined impacts of discharges from the Power Station, the Existing Power Station and all other discharging sites will not pose a risk to non-human biota Uncertainty and Variability in the Non-human Biota Dose Assessment 706. There are a number of uncertainties associated with the assessment of the impact of radioactive discharges on non-human biota. These are: The estimate of the radioactive discharge to the environment; The dispersion of radioactivity following aqueous and gaseous discharges to the environment; The transfer of radioactivity in the environment; and, Uncertainties related to the ERICA and R&D 128 assessment methodologies. These are discussed in Appendix H. Page 225 of 255

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241 8 Management Arrangements 707. As a precondition to the application for an EP-RSR, there is a requirement that an applicant must have in place an effective, written management system which will enable it to achieve compliance with the conditions of the issued EP-RSR. In observance of this precondition Natural Resources Wales (NRW) requires, under Q12 of its application form RSR-A, the provision of a Management Prospectus (MP). In fulfilment of this requirement a copy of Horizon's MP is presented in Appendix S and a summary is given immediately below The MP does not itself set out the detail of permit compliance, but instead references Horizon s RSR Environmental Permit Compliance Matrix [RD105] where this aspect is addressed. The matrix identifies, for each Project Lifecycle Phase from Development through to Decommissioning (as described in Section 2.5) the management arrangements which are or will be put in place to deliver compliance with the standard conditions which Horizon anticipates to receive in its EP-RSR. The phased proportionate approach to the development of these arrangements is the principal topic of this section of this Application. 8.1 Management Prospectus and Company Manual 709. The MP sets out Horizon's organisational and management arrangements in order to demonstrate that the company is capable of holding and maintaining both the EP-RSR and the Nuclear Site Licence (NSL) for the Power Station. It provides an overview of how Horizon will manage the regulated activities that could affect the nuclear safety, security, radiological environmental protection and quality at the Power Station. It explains that Horizon has an adequate organisational and management structure, a developing integrated management system (IMS) and sufficient human and financial resources to discharge the legal and business obligations associated with the company's activities The MP is supported by the Company Manual which outlines Horizon's corporate governance structure including the composition and accountabilities of the Horizon Board of Directors and Leadership Team. The Company Manual is presented in Appendix T. Both documents are maintained and will be kept under review and updated at appropriate points in Horizon's lifecycle. 8.2 EP-RSR Compliance Compliance Requirements 711. Compliance requirements for the EP-RSR fall into two categories: Standard conditions which Horizon anticipates it will receive in the EP-RSR, as set out in the guidance document How to comply with your environmental permit for radioactive substances on a nuclear licensed site [RD106]; and, Specific conditions which will be set by NRW as part of the EP-RSR The standard conditions are set out in Horizon s RSR Environmental Permit Compliance Matrix [RD105]. Against each condition the matrix identifies the condition owner (or Responsible Person) thereby assigning individual responsibility for compliance within the business. It also lists the associated top tier documentation that is currently in place, or will be produced for future Lifecycle Phases. Page 227 of 255

242 713. The Compliance Matrix is managed in accordance with the Licence and Permit Compliance Management Procedure [RD107]. It is a maintained document and will be updated as Horizon develops, in particular following EP-RSR Grant, to ensure any specific conditions are captured and tracked Additional Guidance and Arrangements 714. Additional detail on NRW s expectations for RSR compliance are provided in the following permitting guidance documents: Radioactive Substances Regulation: Management Arrangements at Nuclear Sites (MANS) [RD108]; and, Radioactive Substances Regulation: Environmental Principles (REPS) [RD43] Horizon has also produced its own set of Nuclear, Safety, Security and Environmental Principles (NSSEPs) [RD110] which include expectations for both non-radiological and radiological environmental protection. The NSSEPs are aimed at ensuring the adequacy of Horizon s management arrangements and the Power Station design. They are specifically tailored to Horizon s requirements and objectives, and are written from the operator s point of view. They have been developed using information from a number of sources, including the ONR and the environment agencies The above guidance and principles, along with the standard EP-RSR conditions, are listed and tracked in Horizon s Consolidated Compliance Requirements (CCR) list [RD111]. Against each requirement the CCR includes a compliance statement, lists relevant HMS documents and process groups (e.g. Power Station Development; Environment, Health and Safety; Design Integrity; etc), compliance timeframe requirements, the process owner, and the current status (low level activity; arrangement published but not fully implemented; detailed arrangement fully implemented). 8.3 Horizon s Approach to Developing Compliance 717. Horizon is developing its compliance arrangements in a phased manner according to the Project s Lifecycle Phases: Development; Construction; Commissioning; Generation; and, Decommissioning. Information on these phases is given in Section 2.5. Page 228 of 255

243 718. The development of compliance involves the following: Identification of the Horizon Licence/Permit Condition Owner for each compliance requirement; Identification by the Horizon Licence/Permit Condition Owner of the management arrangements necessary to achieve compliance. The owner is supported in this activity by the Licence and Permit Condition (LC/PC) Compliance Team; Against each compliance requirement, listing of the associated management arrangements in the RSR Environmental Permit Compliance Matrix and the CCR by Project Phase. Each compliance requirement can have multiple associated management arrangements; Allocation of the delivery date for each management arrangement consistent with Project Phase and current programme, and registry within the HMS. This allows the progress of each arrangement s development to be tracked (using a traffic light system). The HMS is also the site through which the resulting documentation can be accessed using the portal illustrated in Figure 5.2 of the MP; Development of the management arrangements by the various process owners; Verification and other QA processes; and, Issue of the management arrangements The phased approach means that the arrangements will be developed and implemented at the most appropriate time in the Project s development, and will be proportionate to the significance of the risks from activities that are being or are due to be undertaken. In contrast, producing and maintaining a range of management arrangements for activities not due to be undertaken for some considerable time would be unnecessary, and would not proportionately contribute to environmental protection. It is noted that the phased approach also means that at the point of EP-RSR Application, a comprehensive set of arrangements to meet the standard EP-RSR conditions will not be in place At the end of each Lifecycle Phase, and prior to the commencement of the next, Horizon will complete a readiness review [FA RSR-9]. Readiness reviews are the mechanism by which the company demonstrates that it is in a suitable state to proceed with activities where there is a significant uplift in risk. A set of relevant criteria are identified for the review and a panel makes a recommendation to proceed based upon the organisation s demonstrable capability and maturity The readiness reviews are intended to provide reassurance of Horizon s suitability to undertake the activities defined in the next Project Phase. Successful completion of a review is a prerequisite for the release of the associated Hold Point, thereby allowing transition to the next Phase. Following completion of a readiness review the resulting report will be provided to NRW An outline of some of the compliance issues associated with the current Development Phase is given below. Thereafter a flavour is given of those associated with subsequent Phases. Page 229 of 255

244 8.3.1 Development Phase Horizon has been developing management arrangements suitable for its current activities at the Development Phase. During the period between EP-RSR Application and EP-RSR Grant, the company will further grow and develop the arrangements so that they are fit for purpose and fully capable of ensuring compliance with the EP-RSR conditions and limitations. Specific examples of the management arrangements are as follows: Arrangements to manage the Early Contractor Engagement (ECE) and Engineering, Procurement and Construction (EPC) contracts are in the process of being developed. These include the demonstration of Best Available Techniques (BAT) for the detailed design ahead of the procurement of Long Lead Items (LLI). BAT assessments undertaken as part of this activity will be discussed with NRW and can be provided as part of the on-going regulatory engagement process; Management arrangements for the shadow Corporate Radioactive Waste Adviser (RWA) body have been developed and these arrangements are currently in operation. Horizon is expecting to make its application for Corporate RWA Body status by the end of 2017 (see Section 8.4.5); Before EP-RSR Grant Horizon will have appropriate arrangements in place to act as a permit holder (operator): Page 230 of Unless otherwise agreed, conditions for routine reporting and notification under Section 4 of the EP-RSR will have immediate effect. Therefore, arrangements to manage those activities will be in place in advance of EP-RSR Grant. - Arrangements to ensure appropriate access to and display of the EP-RSR will also be in place, reviewed and updated as necessary. The Generic Design Assessment (GDA) process is due to be concluded during determination of the EP-RSR Application. Following completion of the GDA and the granting of a Statement of Design Acceptability (SoDA) by the Environment Agency (EA) (in conjunction with NRW), Horizon will adopt the generic design to form the baseline design for the Power Station. This will require the development of arrangements to manage the design, including a comprehensive configuration control process for the Safety and BAT Cases; As part of the GDA process, a number of draft Assessment Findings have been generated which Horizon must resolve during the detailed design, procurement, construction or commissioning phase of the Power Station. It is expected that these Assessment Findings will be formalised when NRW/EA provide their decision of whether or not to issue a SoDA for the UK ABWR in December Horizon will develop a programme for the management of the Assessment Findings and will provide NRW with a copy of the programme; During the Development Phase physical work is likely to be performed to clear the Wylfa Newydd site (including demolition and removal of existing buildings/structures) and to install the perimeter fence and controlled access points. Whilst the risk has been identified as being low, there is potential during site clearance work for the discovery of radioactive objects or radioactive contamination (originating from the Existing Power Station). Arrangements will thus be put in place to ensure that, should such a situation occur, an appropriate solution for the management of the waste can be identified and implemented. The arrangements (which aren't a requirement of the EP-RSR Application) will include procedures on reassurance monitoring and responding to the detection of on-site contamination, and local rules.

245 8.3.2 Construction Phase 723. During the Construction Phase arrangements will be developed to manage the design, procurement, manufacturing, transport, storage, installation and testing of Structures, Systems and Components (SSC) and to ensure they meet the expectations of the BAT Case. These arrangements will be completed prior to the commencement of the EPC contract A number of the SSCs will be procured as LLIs, in advance of the EPC contract being let. A rigorous specification and auditing programme, and procedures for independent assurance and third party inspection will be put in place in order to ensure that the Power Station is delivered to the expected quality and specification Technical Specifications will be developed for the operation of the SSCs. These will specify requirements for Examination, Maintenance, Inspection, and Testing (EMIT), and incorporate the procedures to operate the Power Station Prior to the Commissioning Phase, a programme for environmental monitoring will be established [FA RSR-8], as described in Section 6. Procedures for the performance of environmental monitoring, including the specification of monitoring standards and suitable equipment, will be developed in advance of the commencement of the programme As part of the Construction Phase, the detailed design and construction of facilities that generate, process, store and dispose of solid, aqueous and gaseous radioactive wastes, will commence. This includes the Lower Activity Waste Management Facility (LAWMF). A BAT Case for the detailed design and intended operation of LAWMF, and other waste facilities (i.e. the ILW Storage Facility and Spent Fuel Storage Facility (SFSF)) will be provided to NRW in advance of the procurement of equipment to be installed in these facilities [FA RSR-10] Commissioning Phase 728. As part of the Commissioning Phase, Start-up Testing will be undertaken to demonstrate that the reactor is capable of commencing commercial operations. In advance of placing fuel in the reactor, Horizon will demonstrate that it is capable of complying fully with all of the operational requirements of the EP-RSR. The Control of the Operations Process System Description and underpinning Technical Specifications will be tested, reviewed, and updated by Horizon as part of the commissioning process [FA RSR-11] Arrangements will be in place for the management of radioactive waste, including disposal in line with the requirements of the EP-RSR. These will be developed, implemented and tested in advance of the commencement of regulated activities Generation Phase 730. During the Generation Phase Horizon will operate the Power Station in accordance with its Technical Specifications. These will be subject to configuration (change) control, and will be reviewed and updated on an on-going basis. Comprehensive arrangements for the management of EP-RSR compliance requirements will be maintained and updated in line with operating experience throughout this Phase. Routine activities such as reporting (which will have been performed since receipt of the EP-RSR) will continue as information is generated from the Power Station s operations. Performance will be monitored and further optimisation of discharges and disposals of radioactive waste would be anticipated. Page 231 of 255

246 8.3.5 Decommissioning Phase 731. Arrangements for management and disposal of radioactive wastes will be reviewed, updated or, where appropriate, replaced to enable the management of bulk radioactive wastes during the Decommissioning Phase. The company organisation is likely to be restructured to support these activities The activities which will be undertaken in the Decommissioning Phase will differ from those carried out in the Commissioning and Generation Phases of the Power Station s lifecycle. Therefore, there will be a requirement to vary the EP-RSR. Horizon will engage with NRW / the relevant authority at the time, over the necessary variation prior to the commencement of decommissioning activities. Arrangements for the performance of environmental surveys in advance of releasing the site from radioactive substances regulation will be developed and agreed. 8.4 Development of Key Compliance Areas 733. This section summarises Horizon s management approach to the development of compliance arrangements for a number of key areas Compliance Management 734. The Construction Oversight Director is responsible for obtaining and maintaining the EP- RSR. Assuring compliance with the EP-RSR will be the responsibility of the Safety and Licensing Director The Licence and Permit Condition (LC/PC) Compliance Team, working with Licence/Permit Condition Owners, is responsible for the identification of requirements and expectations with respect to the EP-RSR. Oversight of both the adequacy and the level of compliance is achieved through provisions including management self-assessment, audit, assurance and senior management review Compliance Management will continue over the whole life of the Project. Once the arrangements are developed and implemented, Horizon will remain subject to NRW scrutiny until the site has been fully decommissioned and the EP-RSR surrendered. This scrutiny will include regular interventions and inspections with the submission of requested documentation and records Organisational Readiness The Nuclear Baseline 737. The organisational capability requirements for foreseeable EP-RSR activities are detailed by the Horizon Nuclear Baseline against a defined resource plan. The Nuclear Baseline [RD112] is key in demonstrating Horizon s organisational readiness to proceed through the Project Lifecycles. It quantifies the current capability and capacity to deliver activities and demonstrates that the organisation understands what resources are required to deliver its planned activities as the nature and extent of the activities change A number of key performance indicators are being developed for the Nuclear Baseline. These are designed with the intention of readily portraying the status of Horizon s Nuclear Baseline Organisation in terms of: Completeness of Defined Nuclear Baseline Roles demonstration of progress of resourcing Horizon s Nuclear Baseline Roles against its recruitment plan; Page 232 of 255

247 Competency Status an assessment of Horizon s overall competency of its Nuclear Baseline Organisation; and Adequacy aims to provide an indicative overview of the suitability of Horizon s Nuclear Baseline Organisation to deliver its activities by monitoring the number of changes to the Nuclear Baseline Organisation. Further information on the Nuclear Baseline key performance indicators are detailed in [RD112] Key to the successful ongoing maintenance and management of the Nuclear Baseline are robust management of change arrangements. Accordingly, Horizon has put in place the Management of Nuclear Baseline Change Procedure [RD113] to ensure that any implementation of a change to its Nuclear Baseline Organisation is assessed and managed The competency of Horizon s Nuclear Baseline Organisation is overseen by the Organisational Effectiveness Committee. In order to assure itself that its activities are being carried out adequately, Horizon has processes in place to define competency requirements of the Nuclear Baseline Role holders [RD114], and assess and record the competency of the Horizon Organisation [RD115] The prioritisation of training and competency assessment is driven by the need to ensure that the required levels of organisational competency are in place in time to undertake planned activities. It also maintains the required competency within the Nuclear Baseline Learning Experience 742. Horizon is a learning organisation committed to continuous improvement in all aspects of its business, but particularly in the delivery of nuclear safety, security and radiological environmental protection. Horizon also seeks to be a high reliability organisation as outlined in Fundamental Principle #1 one of its NSSEPs. This will be achieved through the company s approach to learning and improvement, a positive safety and environment culture, the drive for continuous improvement, and the avoidance of complacency The management and organisational principles in the NSSEPs describe how the actions, attitudes and expectations of all, from the Board downwards, contribute to Horizon s development and learning. Horizon will continue to develop its organisational learning and Knowledge Management arrangements such that these remain proportionate to each Phase of the Project Further information on Horizon as an Learning Organisation is presented in the MP. Page 233 of 255

248 8.4.4 Records 745. The Head of Information Technology (IT) within Horizon is responsible for specifying requirements for Horizon s records management system, and ensuring that management arrangements are developed to meet these requirements. This is achieved via the Horizon Management of Records Process System Description [RD116] Horizon is currently developing a comprehensive system for the management of records that will cover the lifetime of the Power Station: Electronic records are accessible from Horizon s Sunrise House headquarters and site office; Paper records are stored in the secure archive at Sunrise House. It is anticipated that secure off-site storage will be used in future as deemed appropriate; Following EP-RSR Grant information will be maintained in relation to compliance. Horizon s Record Retention Schedule and Record Registers will be developed and maintained to ensure that Horizon is compliant with its legal requirements; Records will be maintained to enable Horizon to demonstrate that the design of the Power Station represents BAT; Records from the procurement, manufacturing, transport, storage, installation and testing of SSC performing environmental protection requirements will be retained to substantiate that the as-built design meets Horizon s expectations and remains demonstrably BAT; Records will be produced throughout the Power Station s life time to demonstrate that wastes stored, pending disposal to the Geological Disposal Facility (GDF), meet the relevant acceptance criteria, and are maintained in accordance with the Joint Regulatory Guidance for the management of high activity radioactive waste [RD117]; and, Records of waste disposals under the EP-RSR will be maintained by Horizon until otherwise notified by NRW. This responsibility will be clarified as part of the EP-RSR surrender process Radioactive Waste Adviser 747. Horizon is currently operating a shadow Corporate RWA body but is expecting to make its application for Corporate RWA Body status to the RWA Approval Board by the end of If approved, Horizon will be granted Corporate RWA status The Corporate RWA body shares the duty of advising on EP-RSR compliance amongst more than one individual. The Safety and Licensing Director is accountable for appointing the Corporate RWA Body and its effective operation, in the latter case by ensuring the appointment of a sufficient number of individuals of the required level of competency and experience to provide the necessary advice. Members of the Corporate RWA body have/will have been provided with appropriate training and competency-assessed for their roles prior to appointment The Head of the Corporate RWA body is responsible for its management and for jointly chairing the Corporate RWA & Radiation Protection Adviser (RPA) Committee meetings with the Head of the Corporate RPA Body. Regular Corporate RWA & RPA Committee meetings will be held, in accordance with the Corporate RWA & RPA Body Committee Page 234 of 255

249 Terms of Reference, in order to review performance of the Corporate RWA Body and ensure its on-going effectiveness Procedures [RD119] [RD120] are in use for the operation of the Corporate RWA body, as well as the appointment of individuals to the body, and assessing resilience against future needs. Evidence of advice provided by Members of the Corporate RWA body to Horizon (including contractors working on behalf of Horizon) is routinely recorded in the Corporate RWA Body Advice Log. To determine how that advice has been acted upon, Horizon will undertake spot checks as part of the self-assessment and internal audit processes A RWA & RPA Lifecycle Matrix will define the level and scope of advice that is anticipated from the Corporate RWA Body during each Lifecycle Phase. This will inform the resource requirements for the Corporate RWA Body, and the number of individuals appointed at any one point in time Control of Operations 752. The development of detailed management arrangements for the control of operations will be carried out in a phased manner throughout the Development and Construction Phases, as described earlier. The procedures developed will be tested, implemented and where required updated as part of the Commissioning Phase. The Operations Team will operate the Power Station in strict compliance with the procedures Technical Specifications (see Section 8.3.2) and Environmental Specifications will ensure that SSC performing functions in the management, monitoring, discharge or disposal of radioactive waste meet acceptance criteria, are kept in good repair through appropriate maintenance, and are operated in a manner that is BAT and complies with EP-RSR conditions and limitations During the Commissioning Phase approved criteria will be used for the acceptance testing of equipment which delivers the requirement to avoid and reduce the generation of radioactive waste. This will be undertaken using the Technical and Environmental Specifications It is anticipated that updated and additional procedures will be developed in support of the Decommissioning Phase, although some Technical and Environmental Specifications will be retained during the initial decommissioning works, for example to enable the removal and packaging of spent fuel and drainage of the reactor system and spent fuel pools Management of Design 756. Horizon has established a Design Authority which has responsibility for managing all aspects of the overall facility design that have an effect on nuclear safety, security and radiological environmental protection. This includes regulated activities under the EP- RSR During the Development Phase, the Design Authority is responsible for: Specifying and ensuring design compliance with nuclear safety, security and environmental requirements. The scope of this activity covers the entire facility (and subsequent Lifecycle Phases); Accepting and owning the totality of the design via a rigorous design review process. This includes acting as the Intelligent Customer (IC), as required; Page 235 of 255

250 Managing design changes, including concessions and deviations, where there is a potential impact on nuclear safety, security or the environment; and, Ensuring the production and maintenance of the Safety Case and the BAT Case (see Section 8.3.1) The Design Authority is also responsible for configuration control. All design solutions from suppliers will be managed via a review and acceptance process. The point at which Horizon becomes responsible for the design of or changes to a SSC varies for: Generic Design scope: the responsibility transfers from the reactor vendor (Hitachi-GE) to Horizon at the end of the GDA process; LLIs: Horizon will be responsible for change control management after its requirements have been defined and it has approved for use the specification and design of each LLI; and, Owners scope (including site-specific items not covered by the above): Horizon will be responsible for change control management after having specified the requirements and approved the design Arrangements to support LLI design (via reviews) and procurement, and BAT Case management [RD118] are in place at the point of EP-RSR Application. A mature design process addressing adoption of EPC and other vendor designs will be developed prior to EP-RSR Grant It is anticipated that changes/modifications to the Power Station will be required in future and this will be managed via the Design Change Control Process [RD121]. Configuration management [RD122] will be used to record the information upon which the EP-RSR Application is based, and design changes will be assessed against this. Procedures for the management of design changes during Construction and Commissioning will be developed and implemented ahead of those Phases. Support by the EPC contractor (and other contractors) will take place under the oversight of the Design Authority Periodic review and improvement arrangements will be developed and implemented ahead of the Generation Phase. Requirements for decommissioning are considered as part of the on-going design work within the Development Phase. In addition to these, it is expected that facilities and their design will require modification/modernisation towards the end of the commercial operations to enable decommissioning of the Power Station Managing the EPC Contract 762. Horizon will develop and manage the EPC Contract. This contract will cover the construction of the Power Station, including the detailed design of SSC performing environmental protection and the procurement of LLIs. The terms of the contract, including the the technical requirements for the design of the Power Station and the outline schedule for its construction, are currently being negotiated The focus of the EPC contract is on the Construction and Commissioning Phases of the Project. During these Phases the site will be managed by the appointed EPC contractor using systems, processes and procedures, developed for the Project, that incorporate applicable industry requirements and integrate with Horizon s own processes and requirements. As part of its IC responsibilities, Horizon will maintain oversight of activities with the level of scrutiny being dependent upon the significance of each activity. Page 236 of 255

251 764. The majority of arrangements for the management of the EPC contract will be developed during the Development Phase, in a prioritised manner, in readiness for activities being undertaken under the ECE and EPC contracts Following handover of the site to Horizon for the final Commissioning Phase (Start-up Testing) and movement into the Generation Phase, the EPC contract will be replaced with Long Term Service Agreements which will provide services to support Horizon s proposed arrangements Maintenance 766. A formal system for ensuring the adequacy of maintenance of SSC will be produced, tested and reviewed ahead of the commissioning of that plant. The system will be developed and implemented as Horizon progresses the detailed design of SSC and in doing so identifies the maintenance requirements. Processes for identifying the environmental protection functions of SSC are being developed to support the BAT Case. Performance indicators will be developed to demonstrate that maintenance is effective and the performance of the plant is not deteriorating Horizon has commenced work looking at the maintenance governance documentation which will be required for the operation of the Power Station. The production of these documents will start in the Construction Phase following benchmarking and interacting with Horizon s operating partners (Exelon and The Japan Atomic Power Company) A process will be developed to manage the requirements of the EP-RSR and NSL for EMIT to enable Horizon to support the detailed design of LLIs. This will address the maintenance activities necessary to ensure the requirements of the BAT and Safety Case on LLIs are maintained through manufacturing, transport, storage and installation. This activity will primarily be performed by the EPC contractor. However, Horizon will provide oversight to ensure that the activities are being undertaken appropriately Horizon is currently in discussions with the EPC contractor regarding a maintenance request for proposal (RFP). The RFP will cover various interactions with the EPC contractor including an embedded period within Horizon to promote familiarity with maintenance procedures Environmental monitoring SSC (i.e. plant and equipment to take samples and conduct measurements, tests, surveys, analyses and calculations in compliance with the EP-RSR) will be installed prior to the Commissioning Phase. Following commissioning and operation of the equipment, maintenance activities will be undertaken on a routine basis. During the latter part of the Construction Phase (i.e. during Construction Testing and Pre-operational Testing), a complete maintenance regime will be confirmed to ensure compliance with the EP-RSR. This will include performance metrics and dashboards, maintenance scheduling, and competency assessments for maintenance engineering and operational functions A complete set of maintenance arrangements will be in place once Horizon is ready to undertake Start-up Testing. These arrangements will have been written, tested, reviewed and accepted into service ahead of nuclear fuel being brought to site. They will be in place throughout the Power Station s commercial operations and, as required, beyond. The arrangements will be updated if necessary to reflect best practice and changing requirements throughout the Generation and Decommissioning Phases. Page 237 of 255

252 Waste Management 772. The Environment and Waste Management (E&WM) Function is responsible for the management of radioactive waste together with spent fuel, conventional waste and environmental protection. During the Development Phase, the E&WM is providing specialist technical input in support of activities relating to environment and waste management, including optioneering and BAT justifications which will inform design development. E&WM has specified Horizon s Radioactive Waste Management Arrangements [RD123]. These will be developed and will be in place prior to the Generation Phase of the Power Station During the Development and Construction Phases, comprehensive arrangements will be developed to operate the plant used to manage and dispose of radioactive waste. The development of such arrangements will ensure that a fit-for-purpose set of documents is in place for the management of radioactive waste following the commencement of commissioning and operational actvities It is anticipated that an updated suite of arrangements will be required to enable management of the radioactive waste generated on a bulk scale during the Decommissioning Phase. However, the Radioactive Waste Management Arrangements [RD123] are deemed suitably robust and comprehensive to cover the general requirements for decommissioning at this stage of the Project Monitoring Arrangements 775. Horizon is in the process of developing an Environmental Monitoring Strategy which outlines a framework for implementing and managing environmental monitoring obligations expected to arise from regulatory requirements and voluntary commitments. This Strategy will inform the preparation of environmental monitoring plans and programmes for the various environmental permits granted under EPR16, including EP-RSR The monitoring arrangements for the Power Station will develop as Horizon progresses throughout the Project Lifecycles Phases. Requirements for each Lifecycle Phase will reflect the activities being undertaken. With respect to monitoring and the EP-RSR, the two key areas are: Development of an environmental monitoring programme [FA RSR-8] which will be used for retrospective dose assessments (discussed in Section 6); and, Specification of equipment and development of monitoring requirements for SSC which include in-process monitoring, direct monitoring of discharges, and characterisation of radioactive waste [FA RSR-3] [FA RSR-4] [FA RSR-5] [FA RSR-6]. This will form the basis for the monitoring requirements given in the Technical Specifications that will be used to operate the Power Station Horizon will ensure, via the design process, that environmental monitoring requirements are identified and used to achieve a fully compliant Power Station design. Horizon will also ensure that space, access and utilities keep options open for monitoring as the design advances. It is recognised that additional monitoring requirements will be identified and specified as the Power Station design develops A pre-construction radiological environment baseline for the site has been established for the purposes of the Development Consent Order (DCO). For the purposes of compliance Page 238 of 255

253 with the EP-RSR, a pre-operational radiological environmental baseline for the site and surroundings of the Power Station will be established prior to the Commissioning Phase Horizon will commit to the periodic review of monitoring plans and programmes. As part of the review, Horizon will ensure that the most up-to-date standards and guidance are employed so that the techniques and instrumentation in operation represent BAT Notifications and Reporting 780. During operations, Horizon will report information to NRW, such as sampling and monitoring data. This will be carried out via the Independent Assurance Team in line with the requirements of the environmental permits Horizon will formally notify NRW of any exceptional events. These include, but are not limited to: Malfunctions and accidents leading, or potentially leading to significant pollution accidents; Breaching a limit or QNL specified in the EP-RSR; and, Significant adverse environmental effects that could reasonably be seen to result from the operation of the facility Horizon has arrangements in place for informing regulatory stakeholders of site incidents and emergencies. In advance of EP-RSR Grant, Horizon will ensure that its procedures for reporting and notification relating to radioactive substances activities will meet the formal notification requirements of the EP-RSR Horizon s arrangements for reporting and notification will continue to develop beyond EP- RSR Grant. As the level of activities increase on the site, processes will be expanded to ensure appropriate information is at all times provided to NRW. Following EP-RSR Grant, reporting and notification will be on-going until EP-RSR Surrender following completion of decommissioning and dispatch from site of all radioactive wastes to the GDF Assurance 784. Assurance consists of internal independent review and external independent verification respectively. The latter makes use of competent external organisations, industry peer reviews and regulatory reviews and inspections. The exercise is aimed at identifying areas at risk of non-compliance with the EP-RSR The Head of Function accountable for Assurance has accountability within Horizon for the framework and delivery against the NORA Mandate 30 [RD124]. The mandate is granted by the Chief Executive Officer to give a full independent remit to Horizon to assure its activities. Functions include: Establishment of the Independent Nuclear Safety Review and Independent Environmental Review Procedures and Personnel; Management of notification and reporting requirements; and, Presentation to the Horizon Board of Directors of an independent view of performance. 30 The NORA Mandate is expected to be updated to become the Assurance Mandate before the end of Page 239 of 255

254 NSSEPs have been established to guide Horizon's development and form the basis for assurance activities Horizon's internal audit function is controlled by the Audit Programme Coordinator (APC) within the Corporate Services Department. The APC reports directly to the Risk and Audit Committee and Corporate Services Director Independent review is built into Horizon processes for the development of the design and organisation. It is undertaken by experienced individuals who have an appropriate technical or regulatory background The assurance activities undertaken by Horizon are mature but will continue to develop as a key part of its compliance approach throughout the lifetime of the Project. The approach to assurance and associated procedures will be updated and improved based on operational experience and feedback Quality Assurance 789. Effective quality assurance arrangements underpin all of Horizon s activities and support compliance to related EP-RSR requirements. These arrangements are integral to the design process and the Safety and BAT Cases which define the techniques and technology used to ensure nuclear safety and protection of the environment. The application of quality processes ensures the integrity of information related to discharges and doses, and therefore stakeholder confidence that the impact of the development is within permitted bounds during operations The Head of Quality Assurance & Management Systems has responsibility within Horizon for the framework for assuring quality (although all employees and contractors are responsible for delivering quality at each Phase). Horizon s management arrangements are accredited to ISO The project delivery process is run alongside formal quality planning arrangements where delivery by competent individuals is carried out in accordance with procedures which have been developed and approved using a robust system of review and challenge. Horizon projects are executed in accordance with approved quality plans to ensure that contract, specification and code requirements are followed Horizon uses a significant number of contractors and therefore its IC capability is being developed to scrutinise documentation provided by the supply chain as an essential part of ensuring quality. Quality plans from suppliers will be reviewed in accordance with the Supplier Quality Documents Review procedure [RD125]. Where required, this will involve review and mark-up from the appropriate environmental Technical Lead and/or Subject Matter Leads acting as IC for Horizon Horizon operates a Corrective Action Programme (CAP) where events and issues are captured and corrective actions are assigned to resolution owners. Non-conformities are screened by the Heads of Function, and Resolution Owners are appointed to correct and prevent recurrence. The CAP also aims to record and disseminate good practice within Horizon Management review exists to ensure that Horizon leadership takes an active interest in quality management. This is based upon a set of performance indicators developed to assist in the demonstration that good environmental performance is being achieved. Page 240 of 255

255 795. Quality assurance will remain key in all future activities. When Horizon procures and constructs SSC for the Power Station through the EPC Contractor, it will ensure that the quality expectations of the supply chain are defined and sufficient information is available for review by Horizon to assure that these have been met The quality arrangements for the EPC contractor are key in ensuring that the as-designed plant is being delivered, accepted, installed and commissioned using specifications in accordance with applicable industry standards. In particular, information related to factory acceptance and site acceptance testing prior to radioactive materials being used is key. Records management will also be a key activity co-ordinated and defined through the quality management process Incidents and Emergency Response 797. During the Phases prior to fuel coming to site, Horizon s emphasis is on developing emergency arrangements to deal with conventional hazards related to a large construction site and programme. Contingency arrangements will be developed by the contractors against risk assessment to ensure that all risks have adequate and proportionate emergency arrangements in place to ensure the safety of staff, the public and the environment. These contingency arrangements or plans will be supported by an overarching Horizon response structure which will manage the strategic and tactical level responses during these Phases of the Project The contingency arrangements will eventually be replaced by the Site Operational Emergency Response Arrangements which will be built from the overarching Horizon response structure and will then be developed to focus on any incident or emergency from an operating nuclear site including radiological and environmental response The EP-RSR conditions will come into effect following EP-RSR Grant and this will require that arrangements to deal with accidents be expanded to cover radioactive substances. Any radioactive substances used before the site is operational, will be covered by written Contingency Plans to ensure an adequate emergency response. When the site is operational the full emergency plan and arrangements will also include responses to radiological and environmental emergencies Decommissioning 800. As previously highlighted, prior to the commencement of decommissioning, Horizon will apply for a variation to its EP-RSR. The decommissioning of the Power Station will give rise to significant volumes of solid radioactive waste to be disposed of to appropriately permitted facilities, as well as gaseous and aqueous radioactivity that will be discharged to the environment. The majority of the solid waste will comprise components of the reactor systems which have become activated or contaminated by the operation of the Power Station, but there will also be some secondary waste generated by decontamination and dismantling activities, and abatement of gaseous and aqueous discharges. Discharges from decommissioning operations will be driven by decontamination of active systems and removal, after treatment, of the Power Station s operational water inventory The E&WM is responsible for setting Horizon s strategies and processes regarding management of waste and discharges. These policies are implemented via numerous processes and procedures which themselves will ensure control of all decommissioning waste management activities. Page 241 of 255

256 802. Horizon is developing a technical plan which demonstrates that decommissioning the Power Station in accordance with those policies is technically feasible given currently available techniques, identifies the cost associated with so doing, and the mechanisms for funding decommissioning and waste management when the costs fall due. This will be presented in the form of the Funded Decommissioning Programme (FDP) which will comprise the Decommissioning Waste Management Plan (DWMP) and the Funding Arrangements Plan (FAP) In addition to the DWMP and FAP it is anticipated that Waste Transfer Contracts (WTCs) will be agreed with the Skdetary of State for Business, Energy and Industrial Strategy (BEIS) to provide for the transfer of title to the ILW and spent fuel generated at the Power Station to Government upon payment of a fee The E&WM and Radiation Protection and Environment functions are engaged with the Power Station design to investigate options that represent relevant good practice and BAT. This will reduce the risk to people and the environment posed by the eventual decommissioning of the Power Station. An example of this is selection of construction materials to reduce the levels of trace elements in materials that can become activated and will ultimately be disposed of as radioactive waste The records obtained throughout the lifetime of the Power Station (including the oversight of the design, construction and modification processes) will be used throughout the Generation Phase to update the DWMP. This will include the selection of BAT for decommissioning work and securing all relevant regulatory permissions During the Decommissioning Phase, a decommissioning plan will be implemented: all waste will be disposed of to permitted facilities (as part of the revised EP-RSR) and the site returned to an end-state which will be agreed with appropriate regulatory bodies as part of the detailed decommissioning planning work. Arrangements to carry out these activities will be developed and implemented in advance of conducting activities The DWMP will be a live document which will be reviewed and updated to take account of design change or operational experience. The Energy Act 2008 requires that once an FDP, and the constituent DWMP and FAP, have been approved by the Secretary of State, Annual and Quinquennial Reports will need to be prepared by Horizon for submission to BEIS. These will, respectively, provide: A summary of the changes to the decommissioning activities and costs over the previous year (except in the year that a Quinquennial Report is issued); and, A detailed and comprehensive analysis to ensure that the objective of the FDP, which is to ensure sufficiency of the independent fund, continues to be met. The details of the review process will be set out in the FAP. Their undertaking is a legal requirement under the Energy Act It is anticipated that the first review of the approved DWMP will be undertaken around the time of First Criticality. Page 242 of 255

257 9 Forward Work Plan 808. This section presents a Forward Work Plan (FWP) which lists Horizon s commitments to the implementation of further management arrangements and demonstration of Best Available Techniques (BAT) pertinent to this Application for an EP-RSR (Table 9.1). Those commitments which originate from the main body of the Application are designated FA RSR (Forward Action Radioactive Substances Regulation), while those which derive from the BAT Case (Appendix C) are designated FA BC (Forward Action BAT Case) The arrangements associated with BAT will be created, developed, trialled and implemented ahead of radioactive substance activities being undertaken. It is anticipated that the commitments will be incorporated into the EP-RSR in whole, in part, or as modified, and will thus become formal requirements additional to the standard EP-RSR conditions The implementation timings are proposed with respect to the Project Lifecycle Phases (see Section 2.5 for further details). The timings relate to the Development, Construction, Commissioning and Generation Phases of Unit 1 of the Power Station. Unit 2 will be developed on a schedule approximately 16 months behind. The implementation timings presented in Table 9.1 represent the completion milestone for each commitment. The work towards this milestone will commence with the identification of a Horizon owner who will develop a programme of work to complete the commitment; this programme will be shared with the regulators prior to the work commencing. An indication of the likely overall programmes for commencement and delivery of the FWP is presented in Figure Horizon has scheduled its activities to ensure compliance with the requirements of the EP-RSR as the need arises. The approach is planned, and reflects both the complex and interdependent nature of the various permitting, consenting and licensing regimes for nuclear new build developments, and the fact that the Application for the EP-RSR is being made significantly in advance of commencement of the permitted activities. Further information will continue to be produced as Horizon s organisation and the Power Station design develop between the EP-RSR Application and commencement of operations As noted in Section 8, the Generic Design Assessment (GDA) process, which is on-going, will result in the imposition of a number of Assessment Findings which Horizon will then have an obligation to address. The Assessment Findings will only become formal when they are published in the Environment Agency (EA)/Natural Resources Wales (NRW) Decision Document. The publication date is currently anticipated to be December It is expected that the findings will largely be based upon those presented in the EA/NRW Consultation Document [RD126]. The findings may be included within the EP-RSR or addressed independently by Horizon. Page 243 of 255

258 Table 9.1 Summary of Forward Work Plan Commitments Ref. Source Commitment Timescales FARSR-1 Section 1 Horizon will review and report the findings of the UK ABWR Step 4 technical assessment undertaken by the ONR, andf will ensure that any findings relevant to the EP- RSR are captured and addressed in the Wylfa Newydd BAT Case. FARSR-2 Section 3 Horizon will undertake formulation development work to ensure that the packaged waste volumes of ILW from the Power Station will be minimised as far as reasonably practicable, and to support RWM in a disposability assessment through a Letter of Compliance (LoC) Submission. FARSR-3 Section 6 Horizon will prepare a report which demonstrates that the equipment and arrangements ultimately proposed for the sampling, monitoring, and analysis of radioactive gaseous, aqueous, and solid waste discharges and disposals from the Power Station represent BAT. The report will include consideration of equipment special requirements, suitable and safe access, and sampling arrangements. FARSR-4 Section 6 Horizon will prepare a report evaluating the requirement for in-process radiation monitoring as a means of demonstrating compliance with its EP-RSR, and will demonstrate that any in-process monitoring thus identified represents BAT. FARSR-5 Section 6 Horizon will establish radiation threshold levels for the R/B stack gaseous discharge monitoring system, above which an alarm will be triggered. Horizon will also develop response procedures to such an alarm. FARSR-6 Section 6 Horizon will confirm the monitoring arrangements for the Service Building (S/B) HVAC discharge, as well as all other gaseous discharges outlets not assessed at GDA. FARSR-7 Section 6 Horizon will establish a radiological environmental baseline for the Power Station site and its surroundings. FARSR-8 Section 6 Horizon will determine the BAT with which to monitor the environment and submit its findings to NRW. The development of the programme will be informed by the results of radiological effects modelling along with the results from the monitoring programme for the Existing Power Station and the results presented in the RIFE (Radioactivity in Food and the Environment) report. Throughout the design of the monitoring programme Horizon will engage with NRW to ensure that the programme meets their expectations. 12 months prior to the start of the Construction Phase Prior to the start of the Generation Phase 12 months prior to the start of the Commissioning Phase 12 months prior to the start of the Construction Phase 12 months prior to the start of the Commissioning Phase 12 months prior to the start of the Commissioning Phase (for each individual facility with a gaseous discharge outlet) Prior to the start of the Commissioning Phase Prior to the start of the Commissioning Phase Page 244 of 255

259 Ref. Source Commitment Timescales FARSR-9 Section 8 Horizon will complete a Readiness Review at the end of each Lifecycle Phase and prior to the commencement of the next. Each review will provide reassurance of Horizon s suitability to undertake the activities defined in the next Project Phase. Successful completion of a review will be a prerequisite for the release of the associated Hold Point, thereby allowing transition to the next Phase. Following completion of a readiness review the resulting report will be provided to NRW. FARSR-10 Section 8 Horizon will prepare individual BAT Cases for the design and intended operation of the Power Station facilities that will generate, process, store and dispose of solid, aqueous and gaseous radioactive wastes. These facilities include the Lower Activity Waste Management Facility (LAWMF), ILW Storage Facility, and Spent Fuel Storage Facility. The BAT Cases will be provided to NRW prior to the procurement of equipment. FARSR-11 Section 8 Horizon will demonstrate that it is capable of complying fully with all of the operational requirements of the EP-RSR in advance of placing fuel in the reactor. Comprehensive arrangements will be in place for the management of radioactive waste, including disposal in line with the EP- RSR requirements. FABC-1 BAT Case Horizon shall develop a strategy for the management of failed fuel in the Spent Fuel Pool (SFP) (Expected Events only). As part of this, Horizon must substantiate the assumption that failed fuel rods will not release fission products once in the SFP leading to an increase in radioactivity in the SFPs. FABC-2 BAT Case Horizon must define schemes, operating regimes and maintenance instructions to ensure adequate isolation (in regards to supporting leak tightness) is achieved. FABC-3 BAT Case Horizon will establish a contract with an offsite active laundry service provider for its reusable coveralls. FABC-4 BAT Case Horizon will set up commercial service-level agreements with those permitted waste service providers that have been identified as representing an optimum waste route. This will be implemented to facilitate off-site transfer of radioactive waste. FABC-5 BAT Case Horizon must review, determine and underpin its selection of, if any, pre-treatment/commissioning techniques which may be employed at the Power Station. FABC-6 BAT Case Horizon must undertake a revised direct shine dose assessment of the Spent Fuel Storage Facility to ensure that it does not challenge, from a dose perspective, the concept of ALARA. At the end of each Lifecycle Phase and prior to the start of the next Lifecycle Phase. 12 months prior to the start of the construction of each facility. 6 months prior to the start of the Commissioning Phase. Prior to the start of the Construction Phase. 12 months prior to the start of the Commissioning Phase. 3 months prior to the start of the Commissioning Phase. 6 months prior to the Commissioning Phase. 12 months prior to the start of the Commissioning Phase. 12 months prior to the start of the Construction Phase. Page 245 of 255

260 Ref. Source Commitment Timescales FABC-7 BAT Case Horizon must undertake further work to consider optimisation of the use of the ILW Storage Facility for the storage of decommissioning wastes. It is currently envisaged that a separate ILW store for decommissioning waste arisings will be built at the end of the plant s life but this must be confirmed. Within 10 years of the start of the Generation Phase. Page 246 of 255

261 Figure 9.1 Indicative Programme for the Delivery of the Forward Work Plan Commitments Page 247 of 255

262 [This page is intentionally blank] Page 248 of 255

263 10 References Table 10.1 Schedule of References Ref. No. [RD1] [RD2] Title The Environmental Permitting (England and Wales) Regulations 2016, SI 2016/1154. Department of Energy and Climate Change, UK Strategy for Radioactive Discharges, [RD3] Planning Act [RD4] The Infrastructure Planning (Environmental Impact Assessment) Regulations 2017, 2017 No.572. [RD5] Town and Country Planning Act [RD6] Marine and Coastal Access Act [RD7] [RD8] [RD9] [RD10] [RD11] [RD12] The Justification of Practices Involving Ionising Radiation Regulations 2004, SI 2004/1769. Justification Application UK ABWR Nuclear Reactor, Nuclear Industry Association, December The Justification Decision (Generation of Electricity by the UK ABWR Nuclear Reactor) Regulations 2015, SI 2015/ Process and information Document for Generic Assessment of Candidate Nuclear Power Plant Designs, Version 3, October Environment Agency. [RD13] Nuclear Installations Act [RD14] ONR Licensing Nuclear Installations 4 th Edition Jan [RD15] Consolidated Version of the Treaty Establishing the European Atomic Energy Community (2012/C 327/01). [RD16] Natural Resources Wales. How to apply for an environmental permit Part RSR-B3 New bespoke radioactive substances activity permit (nuclear site open sources and radioactive waste). Guidance Notes. NRW-EPG-RSR-B3 Version 1, December 2015 [RD17] accessed 28/03/2017 [RD18] [RD19] LLW Repository Ltd., Guidance on decision making for management of wastes close to the LLW and ILW categorisation boundary that could potentially cross the LLW boundary, NWP/REP/016 Issue 2, February 2013 Council directive 1999/31/EC of 26th April 1999 on the landfill of waste Page 249 of 255

264 [RD20] [RD21] [RD22] [RD23] [RD24] Hitachi-GE, GEP E7 - Quantification of Discharges and Limits, GA (HE-GD-0004), Rev. F, July 2016 NRW, Criteria for setting limits on the discharge of radioactive waste from nuclear sites, 2012 Environment Agency, Discharges from boiling water reactors A review of available discharge data, July 2016 UK ABWR, GDA GEP, Radioactive Waste Management Arrangements, GA , Rev G, Fundamentals of the management of radioactive waste. An introduction to the management of higher-level radioactive waste on nuclear licensed sites. Guidance from the Health and Safety Executive, the Environment Agency and the Scottish Environment Protection Agency to nuclear licensees. December [RD25] DEFRA, Guidance on the legal definition of waste and its application, August 2012 [RD26] [RD27] [RD28] [RD29] [RD30] [RD31] [RD32] [RD33] [RD34] [RD35] The management of higher activity radioactive wastes on nuclear licensed sites, Joint regulatory guidance, 2015 Generic Design Assessment: Disposability Assessment for Wastes and Spent Fuel arising from Operation of the UK ABWR Part 1: Main Report & Part 2: Supporting Data Nuclear industry guidance an aid to the design of ventilation of radioactive areas, Ref: NVF/DG001 Issue NNVF Discussion Document - Guidance on safe change filter housing design. Issue 01 VMG DD003 June 2015 LLWR Ltd, Packaging Services Brochure Environmental Permitting Regulations (2010): Criteria for Setting Limits on the Discharge of Radioactive Waste from Nuclear Sites, Environment Agency, v1.0, June 2012 Commission recommendation of 18 December 2003 on standardised information on radioactive airborne and liquid discharges into the environment from nuclear power reactors and reprocessing plants in normal operation, Euratom, 2004 (2004/2/Euratom). Corrigendum to Commission Recommendation 2004/2/Euratom of 18 December 2003 on standardised information on radioactive airborne and liquid discharges into the environment from nuclear power reactors and reprocessing plants in normal operation, Euratom, Sampling airborne radioactive materials from the stacks and ducts of nuclear facilities, BS ISO 2889:2010. Stationary source emissions Measurement of velocity and volume flow rate of gas streams in ducts, ISO 10780:1994. [RD36] Equipment for continuous monitoring of radioactivity in gaseous effluents Part 1: General requirements, BS EN :2004. [RD37] Equipment for continuous monitoring of radioactivity in gaseous effluents Part 3: Specific requirements for radioactive noble gas monitors, BS EN :2004. Page 250 of 255

265 [RD38] [RD39] [RD40] [RD41] [RD42] Performance Standard for Organisations Undertaking Radioanalytical Testing of Environmental and Waste Waters, Environment Agency, 2012e. Minimum requirements for the Self-Monitoring of Effluent Flow, Environment Agency, February Performance Standards and Test Procedures for Continuous Emissions Monitoring Systems For gaseous, particulate and flow-rate monitoring systems, Environment Agency, July General requirements for the competence of testing and calibration laboratories, BS EN ISO/IEC 17025:2005. Determination of the characteristic limits (decision threshold, detection limit and limits of the confidence interval) for measurements of ionizing radiation -- Fundamentals and application, ISO 11929:2010 [RD43] Regulatory Guidance Series, No RSR 1; Radioactive Substances Regulation Environmental Principles, Version 3, Natural Resources Wales, September [RD44] [RD45] [RD46] [RD47] [RD48] [RD49] [RD50] [RD51] [RD52] [RD53] [RD54] [RD55] Sampling requirements for stack emission monitoring, Technical Guidance Note (Monitoring) M1, Natural Resources Wales, Monitoring of Radioactive Releases to Atmosphere from Nuclear Facilities, Technical Guidance Note (Monitoring) M11, Environment Agency, 1999a. Monitoring of Radioactive Releases to Water from Nuclear Facilities, Technical Guidance Note (Monitoring) M12, Environment Agency, 1999b. Technical Guidance Note M3: How to assess monitoring arrangements for emissions to air in EPR permit applications, Environment Agency, Version 2, January EA and SEPA, Radiological Monitoring Technical Guidance Note 1: Standardised Reporting of Radiological Discharges from Nuclear Sites, May, 2010, Version 1.0. EA and SEPA, Radiological Monitoring Technical Guidance Note 2: Environmental, December, 2010, Version 1.0. RSR : Principles of optimisation in the management and disposal of radioactive waste, Environment Agency, Rev 2.0, April Nuclear Industry Safety Directors Forum, Best Available Techniques (BAT) for the Management of the Generation and Disposal of Radioactive Wastes, A Nuclear Industry Code of Practice, Issue 1, December Environmental Baseline Report Radiological Monitoring, WD S5-PDC- REP-00007, Rev 0.5. Centre for Environment, Fisheries and Aquaculture Science, Radioactivity in Food and the Environment, 2015, RIFE-21, Prepared on behalf of Environment Agency, Food Standards Agency, Food Standards Scotland, Natural Resources Wales, Northern Ireland Environment Agency and the Scottish Environment Protection Agency. ICRP Publication 103. The 2007 Recommendations of the International Commission on Radiological Protection. Ann. ICRP 37(2 4). ICRP R. Allott, D. Copplestone, P. Merrill & S. Oliver. Habitats assessment for radioactive substances Environment Agency Science report: SC060083/SR1. May 2009 Page 251 of 255

266 [RD56] [RD57] [RD58] [RD59] [RD60] [RD61] [RD62] [RD63] [RD64] [RD65] [RD66] [RD67] [RD68] [RD69] [RD70] [RD71] Copplestone, D, et al. Environment Agency, Impact Assessment of ionising radiation on Wildlife. R&D publication Principles for the Assessment of Prospective Public Doses arising from Authorised Discharges of Radioactive Waste, Environment Agency, August Council Directive 96/29/Euratom of 13 May 1996 Laying Down Basic Safety Standards for the Protection of the Health of Workers and the General Public Against the Dangers Arising from Ionizing Radiation. Official Journal of the European Communities, L159, Volume 39, 29 June Assessing Dose of the Representative Person for the Purpose of Radiation Protection of the Public and Optimisation of Radiological Protection: Broadening the Process. ICRP Publication 101. Ann ICRP, 36, No 3, Application of the 2007 Recommendations of the ICRP to the UK: Advice from the Health Protection Agency. Documents of the Health Protection Agency, Radiation, Chemicals and Environmental Hazards, July Principles for the exemption of radiation sources and practices from regulatory control. Safety Series No 89. IAEA (1988). Application of the Concepts of Exclusion, Exemption and Clearance Safety Guide. IAEA Safety Standards Series No.RS-G-1.7. IAEA (2004). Department of Energy and Climate Change and Welsh Assembly Government (2009). Statutory Guidance to the Environment Agency concerning the regulation of radioactive discharges into the environment. Smith, J.G.,Simmonds, J.R, The Methodology for Assessing the Radiological Consequences of Routine Releases of Radionuclides to the Environment Used in PC-CREAM 08, HPA-RPD-058, Jones, J.A. (1983) The fifth report of a Working Group on Atmospheric Dispersion: Models to Allow for the Effects of Coastal Sites, Plume Rise and Buildings on Dispersion of Radionuclides and Guidance on the Value of Deposition Velocity and Washout Coefficients. NRPB-R157, National Radiological Protection Board, Chilton, Met Office Numerical Weather Prediction models (Parameters for Cemaes coast). Parameter Values Used in Coastal Dispersion Modelling for Radiological Assessments. Report SC060080/R3. Environment Agency, February Radiological Habits Survey: Wylfa Environment Report RL 03/14. CEFAS 2014 Radiological Habits Survey: Wylfa Environment Report RL 02/05. CEFAS 2005 Radiological Habits Survey: Wylfa Environment Report RL 03/10. CEFAS 2010 Radioactivity in Food and the Environment, Page 252 of 255

267 [RD72] [RD73] [RD74] [RD75] [RD76] [RD77] [RD78] [RD79] [RD80] Wylfa Environmental Statement. Part One, Section 8. The Legislative and Regulatory Framework, Radioactive Discharges and Emissions and Nuclear Safety, Magnox 2008 Guidance on the Assessment of Radiation Doses to Members of the Public due to the Operation of Nuclear Installation under Normal Conditions. HPA-RPD-019. Health Protection Agency HPA Guidance on the application of dose coefficients for the embryo, foetus and breastfed infant in dose assessments for members of the public, Oatway, W.B et al, HPA publication RCE-5, ADMS 5 Atmospheric Dispersion Modelling System User Guide, Version 5.1, May 2015 DCPAK3.02 Keith F. Eckerman and Richard W. Leggett Oak Ridge National Laboratory as obtained on 01/02/2016 from National Dose Assessment Working Group (2011). Guidance on short term release assessments. NDAWG Guidance Note 6. Initial radiological assessment methodology part 2 user report Science Report: SC030162/SR2 Environment Agency May 2006 Los Alamos National Laboratory Monte Carlo Code Group [RD81] [RD82] [RD83] [RD84] [RD85] [RD86] [RD87] [RD88] Map of the North Anglesey Marine possible SAC. Available from: Copplestone, D., Hingston, J., Real, A. The development and purpose of the FREDERICA radiation effects database. J. Environ. Radioact. 99, Beresford, N., Brown, J., Copplestone, D., et al. D-ERICA: an integrated approach to the assessment and management of environmental risks from ionising radiation. Description of purpose, methodology and application The ERICA Tool available at Accessed March ICRP Publication 108. Environmental Protection the Concept and Use of Reference Animals and Plants. Ann. ICRP 38 (4-6). ICRP, Ulanovsky, A., Prohl, G., Gomez-Ros, J.M. Methods for calculating dose conversion coefficients for terrestrial and aquatic biota.j. Environ. Radioact., 99, Brown, J.E., Alfonso, B., Avila, R., Beresford, N.A., Copplestone, D., Pröhl, G., Ulanovsky A. The ERICA Tool. J. Environ. Radioact., 99, International Atomic Energy Agency Safety Reports Series No.19 Generic Models for Use in Assessing the Impact of Discharges of Radioactive Substances to the Environment IAEA, Vienna Page 253 of 255

268 [RD89] Avila et al. The selection of parameter values in studies of environmental radiological impacts. Journal of Radiological Protection [RD90] Impact Assessment of ionising radiation on Wildlife. R&D publication 128 Environment Agency Version 2 as revised by SKN_CEN Belgian Nuclear Research Centre, Version [RD91] [RD92] [RD93] [RD94] Vives i Batlle, J., Jones, S.R. and Copplestone, D. A method for estimating 41 Ar, 85, 88 Kr and 131 m, 133 Xe doses to non-human biota. Journal of Environmental Radioactivity, 144, , June Environment Agency. Air emissions risk assessment for your environmental permit Wylfa Newydd Project Pre-Application Consultation - Stage One. Preliminary Environmental Information Report -Volume 1. Natural Resources Wales. Possible new marine Special Areas of Conservation and Special Protection Areas in Wales. February Accessible from: [RD95] Environment Agency. Radioactivity in Food and the Environment RIFE 20. October [RD96] [RD97] Environment Agency. RSR permitting prospective radiological assessments for human health and wildlife (habitats) Operational instruction 338_04 Issued 16/02/2012 Countryside Council for Wales. Site of Special Scientific Interest: Citation for Gwynedd/Anglesey Tre r Gof SSSI. Accessible from: [RD98] Preliminary Environmental Information Report (Version 2, Sept. 2015), Chapter 19, Surface Water and Groundwater [RD99] Environment Agency. Criteria for Setting Limits on the Discharge of Radioactive Waste from Nuclear Sites. Version 1.0, June [RD100] Simmonds, J.R., Lawson, G., Mayall, A. Methodology for Assessing the Radiological Consequences of Routine Releases of Radionuclides to the Environment, RP72, [RD101] Wylfa Newydd Project Pre-Application Consultation Stage One. Preliminary Environmental Information Report Volume 1. [RD102] Wylfa Environmental Statement Part One Section 8. The Legislative and Regulatory Framework, Radioactive Discharges and Emissions and Nuclear Safety [RD103] IAEA-TECDOC Specific Activity Models And Parameter Values For Tritium, C14 And Cl36. In: Quantification of Radionuclide Transfer in Terrestrial and Freshwater Environments for Radiological Assessments. IAEA, [RD104] OSPAR, Towards the Radioactive Substances Strategy Objectives. Third Periodic Evaluation. Radioactive Substances Series. Publication 445/2009. [RD105] RSR Environmental Permit Compliance Matrix Wylfa Newydd, OD057-S5-NL- REP-00017, Rev 0.8. Page 254 of 255

269 [RD106] EA Guidance - How to comply with your environmental permit for radioactive substances on a nuclear licenced site, GEHO0812BUSS-E-E, 478_10, Version 2, Environment Agency, August [RD107] Licence and Permit Compliance Management Procedure, HG-D-02-PRC , Rev 1. [RD108] Radioactive Substances Regulation: Management Arrangements at Nuclear Sites (MANS), Version 2, Environment Agency, April [RD109] Regulatory Guidance Series, No RSR 1; Radioactive Substances Regulation Environmental Principles, Version 3, Environment Agency, September [RD110] Nuclear, Safety, Security and Environmental Principles, HG-M-05-POL , Rev 3. [RD111] Consolidated Compliance Requirements, OD058-S2-MSH-REG-00001, Rev 4.0. [RD112] Nuclear Site Licence Application - Nuclear Baseline Report, WD S5- NLREP [RD113] Management of Nuclear Baseline Change Procedure, HG-S-02-PRC , Rev 3. [RD114] Defining the Organisation Procedure, HG-S-02-PRC , Rev 2. [RD115] Competency Assessment Procedure, HG-S-02-PRC , Rev 3. [RD116] Management of Records Process System Description, HG-S-04-PSD , Rev 2. [RD117] Joint Regulatory Guidance - The management of higher activity radioactive waste on nuclear licensed sites, Revision 2, Office of Nuclear Regulation; the Environment Agency; the Scottish Environment Protection Agency and Natural Resources Wales to nuclear licensees, February [RD118] Management of Best Available Techniques (BAT) Methodology, HG-D-01-PSD , Rev 1. [RD119] Corporate Radioactive Waste Adviser (RWA) Body Membership Procedure, HG-M- 05-PRC , Rev 1. [RD120] Corporate RWA Provision of Advice Procedure, HG-M-05-PRC , Rev 1. [RD121] Design Change Control Process, HG-D-01-PRO , Rev 1. [RD122] Configuration Management, HG-D-01-PSD , Rev 1.0 [RD123] Radioactive Waste Management Arrangements - Radioactive Waste, Spent Fuel and Decommissioning, HNP-S3-EWM-REP-00017, Rev 1. [RD124] The Nuclear Oversight function Mandate, HG-M-05-MAN , Rev 3. [RD125] Supplier Quality Documents Review, HG-D-01-PRC , Rev 1. [RD126] [RD127] Generic design assessment of Hitachi-GE Nuclear Energy Limited's UK Advanced Boiling Water Reactor, Consultation document, 12 December Page 255 of 255

270 8 CONTACT US: If you have any questions or feedback regarding the Wylfa Newydd Project you can contact us on our dedicated Wylfa Newydd freephone hotline and address, by calling on or ing Horizon Nuclear Power Sunrise House 1420 Charlton Court Gloucester Business Park Gloucester, GL3 4AE T +44 (0) All material in this document is, unless specified otherwise, copyright of Horizon Nuclear Power Wylfa Ltd and may not be reproduced without prior permission. Any unauthorised use or copying of the material may violate trademark, copyright and other proprietary rights and civil and criminal statutes. The material shall not be used in any manner that infringes any trademark, copyright or other proprietary rights and Horizon Nuclear Power Wylfa Ltd reserves all rights with respect to such unauthorised use. WN0908-HZCON-PAC-REP-00003