Coversheet SITE EVALUATION OF THE OPG NEW NUCLEAR AT DARLINGTON - PART 2: DISPERSION OF RADIOACTIVE MATERIALS IN AIR AND WATER

Transcription

1 Coversheet OPG Proprietary Document Number: NK054-REP Sheet Number: Revision: N/A R001 Title: SITE EVALUATION OF THE OPG NEW NUCLEAR AT DARLINGTON - PART 2: DISPERSION OF RADIOACTIVE MATERIALS IN AIR AND WATER Ontario Power Generation Inc., This document has been produced and distributed for Ontario Power Generation Inc. purposes only. No part of this document may be reproduced, published, converted, or stored in any data retrieval system, or transmitted in any form or by any means (electronic, mechanical, photocopying, recording, or otherwise) without the prior written permission of Ontario Power Generation Inc. Site Evaluation of the OPG New Nuclear at Darlington - Part 2: Dispersion of Radioactive Materials in Air and Water NK054-REP R Project ID: AMEC NSS: P1093/RP/001 R05 OPG Proprietary Prepared by: [AMEC NSS Per attached] Verified by: [AMEC NSS Per attached] Elizabeth Alderson Badi-Uz-Zaman Khan Arindam Mukherjee Mark Gerchikov Jimin Peng AMEC NSS Date Yung Cheung AMEC NSS Date Reviewed by: [AMEC NSS Per attached] Michael Lee Robert Zimmermann Marcello Oliverio Richard J. Fluke AMEC NSS Date Reviewed for QA by: [AMEC NSS Per attached] Charles Gordon AMEC NSS Date Approved by: Recommended for OPG Acceptance by: [AMEC NSS Per attached] Ron Henry AMEC NSS John Marczak Manager Safety Analysis Review Dept Darlington New Nuclear Project Date Date Accepted for OPG Use by: Bob Goodman Director - Engineering Darlington New Nuclear Project Date Associated with REP N-TMP R001 (Microsoft XP)

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3 Revision Summary Rev Date Author Comments R00 August 25, 2008 Mark Gerchikov et al. Framework Report Issued to Client R01 November 03, 2008 Mark Gerchikov et al. Report Issued to Client R02 January 21,2009 Mark Gerchikov et al. Final Report Incorporating Client s Comments R03 February 27,2009 Mark Gerchikov et al. Updated Report Issued to Client R04 March 20,2009 Mark Gerchikov et al. Updated Report Incorporating Client s Comments R05 August 2009 Mark Gerchikov et al. Revised to address internal and external client comments Confidentiality, Copyright and Intellectual Property Notice 2009 This document and its contents are strictly confidential. It has been produced by for Ontario Power Generation Inc. (OPG) under the terms of the Nuclear Safety and Technology Support and Services Agreement dated December 13, 2005 between OPG and AMEC NSS. Rights of copying and of ownership and use of the intellectual property in or embedded in this document are solely determined by the terms of such Agreement. No part of this document shall be used, reproduced, published, converted or stored in any data retrieval system or transmitted in any form or by any means (electronic, mechanical, photocopying, recording or otherwise) other than in accordance with and subject to such Agreement. If you are not the intended recipient please notify the Contracts Manager, AMEC NSS, (416) or return by post to, 700 University Avenue H4, Toronto, Ontario M5G 1X6. P1093/RP/001 R05 Page 2 of 132

4 Table of Contents Page EXECUTIVE SUMMARY INTRODUCTION Project and Task Background Objectives Normal Operations Accidents Regulatory Framework Overview CNSC RD IAEA NS-G IAEA NS-R CNSC RD Document Structure GENERAL SITE DESCRIPTION AND CHARACTERISTICS Site location and description Programme for meteorological investigation Meteorological and atmospheric monitoring Regional Climatology Temperature Precipitation Wind Lake Effect Atmospheric Stability Atmospheric Pressure Monitoring Programme for Surface Water and Groundwater Lake Water Groundwater Population distribution Uses of Land and Water in the Region Agriculture Industry Transportation Fishing Biological Data Baseline ambient radioactivity and Pre-existing Hazardous substances ASSESSMENT OF DISPERSION OF RADIOACTIVE MATERIALS FOR NORMAL OPERATIONS P1093/RP/001 R05 Page 3 of 132

5 3.1 Radioactive Source Parameters for Normal Discharges to Air Radioactive Source Parameters for Normal Discharges to Surface Water and Groundwater Identification of Potential Critical Groups and Their Characterization Selection of Potential Critical Groups Characterization of Potential Critical Groups Determination of average dilution factor in fish home range Dairy cow s diet Radioactive Effluent Dose Consequence for Normal Operations Environmental Pathway Model Pathway Analysis Model and its Implementation Assessment Results Sensitivity Analysis Radioactive Effluent Dose Consequence for Refurbishment and Decommissioning Radiological Implications from Decommissioning and Refurbishment Source Term and Dose Assessment for Refurbishment and Decommissioning Operations Assessment of Consequences from Radioactive Waste and Used Fuel Management ASSESSMENT OF DISPERSION OF RADIOACTIVE MATERIALS FOR ACCIDENTAL DISCHARGES Derivation of Accident Scenario and Release Characteristics Representative Source Term for Radioactive Airborne Releases Core Inventory Derivation of Release Fractions for RSGB Releases Source Term for Accidental Radioactive Discharges to Surface Water and Groundwater Offsite Public Dose Consequence for Radioactive Airborne Release Accidents Dose Targets and Limits Dose Consequence Calculations RSGB Dose Consequence Results Offsite Public Dose Consequence for Radioactive Waterborne Release Accidents Impact of Mitigation on Dose Consequence for Accidents CONSIDERATION OF THE FEASIBILITY OF AN EMERGENCY PLAN Objectives for Emergency Planning Existing Emergency Plans Protective Action Levels Comparison with IAEA Requirements and Recommendations from Other Agencies Potential Impact on Emergency Planning of OPG s New Nuclear at Darlington QUALITY ASSURANCE PROGRAMME CONCLUSIONS REFERENCES P1093/RP/001 R05 Page 4 of 132

6 APPENDIX Appendix A: Abbreviations Appendix B: Extracts from IAEA GS-R P1093/RP/001 R05 Page 5 of 132

7 Figures Page Figure 2.1-1: Darlington Location. [OPG 2004]...20 Figure 2.1-2: Darlington Site with the Surrounding Regions. [OPG 2004]...21 Figure 2.1-3: Site Access Locations [OPG 2004]...22 Figure 2.3-1: Map of Darlington Monitoring Wells [CH2MHill 2009a]...40 Figure 2.4-1: Population Distribution Grid for the Area Surrounding the Proposed Darlington Nuclear Site...42 Figure 2.4-2: Land Use Prediction to 2031[Regional Municipality of Durham 2008]...43 Figure 2.5-1: Darlington Environmental Monitoring Locations [OPG 2008c]...49 Figure : Locations of potential critical groups...57 Figure : Environmental transfer model [CSA 2008]...61 Figure 4.4-1: Variation of Committed Effective Dose with Distance for RSGB SRF Release EARLY 7 Day Figure 4.4-2: Variation of Dose to Thyroid with Distance for the RSBG SRF Release EARLY 7 Day Figure 4.4-3: Variation of Committed Effective Dose with Distance for RSGB LRF Release - LATE Figure 5.2-1: Darlington Specific Response Sectors in Primary Zone P1093/RP/001 R05 Page 6 of 132

8 Tables Table 2.2-1: Data collected at the Darlington Site from [AMEC NSS 2008a]...28 Table 2.2-2: Data collected at the Bowmanville Station from [OPG 2004]...29 Table 2.2-3: Precipitation data from nearby sites [OPG 2004]...29 Table 2.2-4: Triple Joint Wind Frequencies Stability Class A [AMEC NSS 2008a].30 Table 2.2-5: Triple Joint Wind Frequencies Stability Class B [AMEC NSS 2008a].31 Table 2.2-6: Triple Joint Wind Frequencies Stability Class C [AMEC NSS 2008a].32 Table 2.2-7: Triple Joint Wind Frequencies Stability Class D [AMEC NSS 2008a].33 Table 2.2-8: Triple Joint Wind Frequencies Stability Class E [AMEC NSS 2008a].34 Table 2.2-9: Triple Joint Wind Frequencies Stability Class F [AMEC NSS 2008a].35 Table : Wind Speed Frequencies Grand Total [AMEC NSS 2008a]...35 Table 3.1-1: Estimated maximum Airborne Emission to the Environment [OPG 2009d]...51 Table 3.1-2: Airborne Release Parameters [OPG 2009d]...52 Table 3.2-1: Estimated maximum waterborne emission to the environment [OPG 2009d]...53 Table 3.2-2: Discharge Rate of Liquid Effluents [OPG 2009d]...54 Table 3.2-3: Aquatic Plume Parameters [OPG 2008c]...55 Table : Summary of Potential Critical Groups [OPG 2008c]...56 Table : Transfer Compartments and their Units [CSA 2008]...63 Table : Transfer Parameters and their Units [CSA 2008]...64 Table : Radionuclides of Importance Released from ACR Reactor [OPG 2008c]*...67 Table : Limiting radionuclides in particulate released from ACR-1000 reactor...68 Table : Limiting Radionuclide in Beta-Gamma Released from ACR-1000 Reactor via Water pathway...69 Table : Doses to Potential Critical Groups Resulting from the Operation of AP1000 Reactor (Once-Through Cooling)...71 Page P1093/RP/001 R05 Page 7 of 132

9 Table : Doses to Potential Critical Groups Resulting from the Operation of EPR Reactor (Once-Through Cooling)...72 Table : Doses to Potential Critical Groups Resulting from the Operation of ACR-1000 Reactor (Once-Through Cooling)...73 Table : Doses to Potential Critical Groups Resulting from the Operation of AP1000 reactor (Cooling Tower Option)...74 Table : Doses to Potential Critical Groups Resulting from the Operation of EPR reactor (Cooling Tower Option)...75 Table : Doses to Potential Critical Groups Resulting from the Operation of ACR-1000 reactor (Cooling Tower Option)...76 Table : Effect of temperature change on public dose...78 Table : Effect of flow rate change on public dose...78 Table : Doses to hypothetical groups...79 Table 3.5-1: Comparative Analysis of Radionuclide Inventory During Refurbishment and Decommissioning...82 Table 3.5-2: Pickering A Releases to Air During Refurbishment of Units 3 and 4 (Bq)...84 Table 3.5-3: Pickering A Releases to Water During Refurbishment of Units 3 and 4 (Bq)...84 Table : EPR Core Inventory From Vendor Data (AREVA 2007)...88 Table : Baseline Release...91 Table : RSGB Releases Normalised against SRF and LRF...92 Table 4.4-1: Protective Action Levels...93 Table 4.4-2: Health Canada Recommended Intervention Levels...94 Table 4.4-3: Variation of Committed Effective Early Whole Body Doses with Distance for RSGB SRF...98 Table 4.4-4: Variation of Early Equivalent Dose to the Thyroid with distance for RSGB SRF Release...99 Table 4.4-5: Variation of Late Committed effective Whole Body Doses with Distance for RSGB LRF Release Table 4.4-6: Dose (msv) by Pathway at 1 km from Release Point for RSGB LRF Table 4.4-7: Total Event Committed Effective Doses for RSGB LRF P1093/RP/001 R05 Page 8 of 132

10 Table Protective Action Levels (PALs) P1093/RP/001 R05 Page 9 of 132

11 EXECUTIVE SUMMARY OPG has commenced the federal approvals process for site preparation, construction, and operation of up to four additional nuclear reactors for the purpose of generating up to 4,800 MW of electricity at the Darlington Nuclear Site ( the site ). An application for a License to Prepare the Site was submitted to the Canadian Nuclear Safety Commission (CNSC) in September, The three reactors that are currently under consideration are: Advanced CANDU Reactor (ACR-1000) The Advanced CANDU Reactor (ACR-1000) is a Pressurized Hybrid Reactor (PHR) by Atomic Energy of Canada Limited. The ACR-1000 is based on the Canadian Deuterium Uranium (CANDU) technology. It uses light water (H 2 O) to cool the fuel and heavy water (D 2 O) for the moderator. The ACR-1000 plant (planned) has a net electrical output of approximately 1085 MW. It is a Generation III+ reactor. Areva US EPR (EPR) Areva's US EPR Reactor (EPR) is an evolutionary Pressurized Water Reactor (PWR) by AREVA NP. It has a rated electrical power of 1,580 MW and a design life of 60 years. The reactor operates with enriched uranium fuel (enrichment of up to 5%). The EPR is classified as a Generation III+ reactor. Advanced Passive (AP1000) Reactor The Advanced Passive (AP1000) reactor is another PWR by Westinghouse. Its safety systems operate passively, using natural forces such as gravity and natural circulation in order to function. The AP1000 output is 1037 MWe (net). It operates with enriched uranium fuel (enrichment of up to approximately 4.5%). It is a Generation III+ reactor. AMEC NSS has been contracted by Ontario Power Generation (OPG) to evaluate the Darlington Site for suitability for the new nuclear power plant. This Task 2 Dispersion of Radioactive Material in Air and Water report is produced as part of the Site Evaluation of the OPG New Nuclear at Darlington to satisfy the requirements of Canadian Nuclear Safety Commission (CNSC) Regulatory Document 346 (RD-346) Site Evaluation for New Nuclear Power Plants. RD-346 is consistent with, and was developed based largely on, the following six IAEA Safety Guidelines for which aligned work sub-packages are being completed: NS-G-3.1, External Human Induced Events in Site Evaluation for Nuclear Power Plants [IAEA 2002a] NS-G-3.2, Dispersion of Radioactive Material in Air and Water and Consideration of Population Distribution in Site Evaluation for Nuclear Power Plants [IAEA 2002b] P1093/RP/001 R05 Page 10 of 132

12 NS-G-3.3, Evaluation of Seismic Hazards for Nuclear Power Plants [IAEA 2002c] NS-G-3.4, Meteorological Events in Site Evaluation for Nuclear Power Plants [IAEA 2002d] NS-G-3.5, Flood Hazard for Nuclear Power Plant on Coastal and River Sites [IAEA 2002e] NS-G-3.6: Geotechnical Aspects of Site Evaluation and Foundations for Nuclear Power Plants [IAEA 2002f] This report addresses the requirements that are specified by the IAEA Safety Guide NS-G-3.2 [IAEA 2002b] including: Assessment of risks to the public for routine releases of radioactivity. Accident assessments for bounding release scenarios. Assessment of adequacy of Emergency Response plans to meet new reactor risk. It also addresses the requirements specified in CNSC regulatory document RD-346 [CNSC 2008a] in relation to dispersion of radioactive materials, and IAEA NS-R-3, Site Evaluations for Nuclear Installations [IAEA 2003]. All work to evaluate the site s compliance with the above site evaluation requirements was carried out in accordance with the AMEC NSS Quality Assurance system, which is fully compliant with the ISO 9001 standard. It is also compliant with the requirements of CSA Z and applicable portions of CSA N and N standards. The study presents five key findings confirming that it is feasible to construct a new nuclear power plant at the site to satisfy the above requirements: 1. Monitoring data required to assess radiological impact from radioactive releases into the environment and the baseline information on environmental contamination, hydrogeology, meteorology and population distribution and habits have been collated. The site provides comprehensive datasets surpassing the requirements of IAEA and CNSC expectations with data collated over several decades to monitor performance of the operational nuclear generating station. These have been recently complemented with additional information as part of the Environmental Assessment which is being conducted by Ontario Power Generation. 2. Representative releases of radioactive materials have been estimated for both normal operations and accidents. These estimates were made on the basis of information provided by reactor vendors and, in the case of abnormal releases, on the basis of the Canadian regulatory requirements RD-337 [CNSC 2008b] specifying frequency and credible threshold releases. It is assumed that any P1093/RP/001 R05 Page 11 of 132

13 reactor design licensed for construction in Canada will satisfy these requirements. 3. An assessment of atmospheric, surface water and groundwater dispersion of radioactive materials in the environment was conducted for both normal operations and accidental releases. Doses to the most exposed members of the public have been estimated. 4. Estimated maximum annual doses for normal operations are 5 µsv for both once-through cooling option and cooling tower option, which represent a small fraction of the dose limit of 1 msv. This demonstrates that both options are feasible and can satisfy Canadian and IAEA requirements. Additional sensitivity studies were conducted to gauge the impact on dose if in the future population groups were to be located in locations with maximum exposure. Even such a conservative scenario resulted in estimated doses significantly below the regulatory limit. 5. Stylized accident release scenarios based on the RD-337 Small Release and Large Release Frequency safety goals were modeled to predict possible doses to the public. These results are consistent with predicted conditions for Emergency Planning. The overall impact from the New Nuclear at Darlington on emergency planning from an accident event has been considered in [OPG 2009c]. The results indicate that sheltering for an event such as that modeled could be mandatory to around 2 km from the point of release, while the option of sheltering may be considered for out to 10 km. Evacuation however would only require consideration for less than 2 km from the point of release. The mandatory evacuation dose is only exceeded within 500 m from the point of release. The results indicate that if early phase evacuation takes place, approximately half of the total predicted dose for the release over 1 year is avertable. The relocation limit for dose over 1 year is not exceeded beyond 1 km from the point of release, and as such, no local residents would require permanent relocation. It is possible that temporary relocation measures will be required within this 1 km area for a time immediately following the release and for as long as 1 year. This relocation would only apply to permanent local residents and not to workers or businesses. Based on the RD-337 Safety Goal Based (RSGB) release analyses, the existing Darlington Nuclear Emergency Plan is adequate for the OPG New Nuclear at Darlington as well. P1093/RP/001 R05 Page 12 of 132

14 1.0 INTRODUCTION 1.1 Project and Task Background OPG has commenced the federal approvals process for site preparation, construction, and operation of up to four additional nuclear reactors for the purpose of generating up to 4,800 MW of electricity at the Darlington Nuclear Site ( the site ). An application for a License to Prepare the Site was submitted to the Canadian Nuclear Safety Commission (CNSC) in September, The three reactors that are currently under consideration are: Advanced CANDU Reactor (ACR-1000) The Advanced CANDU Reactor (ACR-1000) is a Pressurized Hybrid Reactor (PHR) by Atomic Energy of Canada Limited. The ACR-1000 is based on the Canadian Deuterium Uranium (CANDU) technology. It uses light water (H 2 O) to cool the fuel and heavy water (D 2 O) for the moderator. The ACR-1000 plant (planned) has a net electrical output of approximately 1085 MW. It is a Generation III+ reactor. Areva US EPR (EPR) Areva's US EPR Reactor (EPR) is an evolutionary Pressurized Water Reactor (PWR) by AREVA NP. It has a rated electrical power of 1,580 MW and a design life of 60 years. The reactor operates with enriched uranium fuel (enrichment of up to 5%). The EPR is classified as a Generation III+ reactor. Advanced Passive (AP1000) Reactor The Advanced Passive (AP1000) reactor is another PWR by Westinghouse. Its safety systems operate passively, using natural forces such as gravity and natural circulation in order to function. The AP1000 output is 1037 MWe (net). It operates with enriched uranium fuel (enrichment of up to approximately 4.5%). It is a Generation III+ reactor. AMEC NSS has been contracted by Ontario Power Generation to evaluate the Darlington Site for suitability for the new nuclear power plant. This Task 2 Dispersion of Radioactive Material in Air and Water report is produced as part of the Site Evaluation of the OPG New Nuclear at Darlington to satisfy the requirements of Canadian Nuclear Safety Commission (CNSC) Regulatory Document 346 (RD-346) Site Evaluation for New Nuclear Power Plants. RD-346 is consistent with, and was developed based largely on, the following six IAEA Safety Guidelines for which aligned work sub-packages are being completed: NS-G-3.1, External Human Induced Events in Site Evaluation for Nuclear Power Plants [IAEA 2002a] P1093/RP/001 R05 Page 13 of 132

15 NS-G-3.2, Dispersion of Radioactive Material in Air and Water and Consideration of Population Distribution in Site Evaluation for Nuclear Power Plants [IAEA 2002b] NS-G-3.3, Evaluation of Seismic Hazards for Nuclear Power Plants [IAEA 2002c] NS-G-3.4, Meteorological Events in Site Evaluation for Nuclear Power Plants [IAEA 2002d] NS-G-3.5, Flood Hazard for Nuclear Power Plant on Coastal and River Sites [IAEA 2002e] NS-G-3.6: Geotechnical Aspects of Site Evaluation and Foundations for Nuclear Power Plants [IAEA 2002f] This report addresses the requirements that are specified by CNSC guidance document RD-346 [CNSC 2008a] in relation to dispersion of radioactive materials, and by the IAEA Safety Guide NS-G-3.2 [IAEA 2002b] including: Assessment of risks to the public for routine releases of radioactivity. Accident assessments for bounding release scenarios. Assessment of adequacy of Emergency Response plans to meet new reactor risk. It also addresses the requirements specified in NS-R-3; Site Evaluations for Nuclear Installations [IAEA 2003] in relation to dispersion of radioactive materials. 1.2 Objectives The objectives are as follows: Normal Operations Objective 1. The evaluation of the transport and diffusion of effluents discharged into the atmosphere for radioactive materials to assess the suitability of the Darlington Nuclear site. Objective 2. The evaluation of the transport and diffusion of effluents discharged into the hydrosphere for radioactive materials to assess the suitability of the Darlington Nuclear site. Objective 3. The uses of land and water in the region of the site should be investigated as part of the environmental assessment and to demonstrate feasibility of the emergency planning. P1093/RP/001 R05 Page 14 of 132

16 1.2.2 Accidents Objective 4. The study of the population distribution to evaluate the potential radiological impacts from accidents and to assist in the demonstration of the feasibility of emergency measures. Objective 5. Demonstration of the feasibility of an emergency plan. 1.3 Regulatory Framework Overview CNSC RD-346 All requirements specified in the CNSC Guidelines [CNSC 2008a] in relation to dispersion of radioactive materials are consistent with those identified by IAEA NS-G-3.2 [IAEA 2002b], with the exception of an additional statement relating to baseline data sample sizes: Baseline data is expected to be of sufficient sample size and duration to conduct hypothesis testing against post-commissioning (follow-up) monitoring data, with sufficient power to detect relevant effect sizes. [CNSC 2008a] The radiological characterization of the site and environs is very extensive due to the data collected for the on-going operation of the site [OPG 2008d] and the additional characterization done in support of the EA [OPG 2009b]. It is expected that this will provide sufficient baseline information for radiological hypothesis testing analysis going forward IAEA NS-G-3.2 IAEA Guide [IAEA 2002b] contains recommendations related to the following activities: Radioactive source parameters for normal and accidental discharge to air Programme for meteorological investigation, collection, analysis and presentation of meteorological data Atmospheric Dispersion Modeling Programme for monitoring of surface water and groundwater. Radioactive source parameters for normal and accidental discharge to surface water and groundwater Modeling of radionuclide dispersion in surface water Modeling of dispersion and retention of radionuclides in groundwater Land use, water use and population distribution in the region. P1093/RP/001 R05 Page 15 of 132

17 Consideration of the Feasibility of an Emergency Plan and Quality Assurance Programme IAEA NS-R-3 With respect to dispersion of radioactive materials this IAEA NS-R-3 Guide [IAEA 2003] does not provide any recommendations in addition to those specified in the IAEA Guide NS-G-3.2 [IAEA 2002b] CNSC RD-337 Although compliance with requirements of the regulatory document Design of New Nuclear Power Plants [CNSC 2008b] is outside the scope of this assessment, it provides the design criteria for Design Basis Accident (DBA) frequencies. For the purposes of this Site Evaluation Study, it is implicitly assumed that any new nuclear plant will comply with these regulatory requirements. Section of the regulatory document is reproduced below: Safety Goals Qualitative Safety Goals A limit is placed on the societal risks posed by nuclear power plant operation. For this purpose, the following two qualitative safety goals have been established: Individual members of the public are provided a level of protection from the consequences of nuclear power plant operation such that there is no significant additional risk to the life and health of individuals; and Societal risks to life and health from nuclear power plant operation are comparable to or less than the risks of generating electricity by viable competing technologies, and should not significantly add to other societal risks. Quantitative Application of the Safety Goals For practical application, quantitative safety goals are established to achieve the intent of the qualitative safety goals. The three quantitative safety goals are: 1. Core damage frequency; 2. Small release frequency; and 3. Large release frequency. A core damage accident results from a postulated initiating event (PIE) followed by failure of one or more safety system(s) or safety support system(s). Core damage frequency is a measure of the plant s accident preventive capabilities. P1093/RP/001 R05 Page 16 of 132

18 Small release frequency and large release frequency are measures of the plant s accident mitigative capabilities. They also represent measures of risk to society and to the environment due to the operation of a nuclear power plant. Core Damage Frequency: The sum of frequencies of all event sequences that can lead to significant core degradation is less than 10-5 per reactor year. Small Release Frequency: The sum of frequencies of all event sequences that can lead to a release to the environment of more than Becquerel of iodine-131 is less than 10-5 per reactor year. A greater release may require temporary evacuation of the local population. Large Release Frequency: The sum of frequencies of all event sequences that can lead to a release to the environment of more than Becquerel of cesium-137 is less than 10-6 per reactor year. A greater release may require long term relocation of the local population. 1.4 Document Structure This report is organized as follows: Section 2 - General Site Description and Characteristics: This is the introduction to the report and contains a general description of the plant. It will mention different monitoring programs that are on the site and the population and land use within a 100 km radius. Section 3 - Assessment of Dispersion of Radioactive Materials for Normal operation: This section describes selection of bounding source terms for discharges to air and water under the normal operations and assessment of the resulting doses to potential critical groups. It contains the analysis of assessment results and evaluates consequences resulting from refurbishment and de-commissioning operations. Section 4 - Assessment of Dispersion of Radioactive Materials for Accidental Discharges: This section describes air and water source terms for beyond design basis accidents (BDBA) and design basis accidents (DBA). It will describe the dose calculation model and provide a detailed analysis of the dose consequences. Section 5 - Consideration of the Feasibility of an Emergency Plan: This section contains an assessment of how the Emergency Plan for Darlington is impacted by the calculated dose consequences. Section 6 - Quality Assurance Programme: P1093/RP/001 R05 Page 17 of 132

19 This section contains a description of the quality system under which the project was executed. It also contains the qualification status of every data stream that was used in this project. Section 7 - Conclusion This section provides a summary of the findings. P1093/RP/001 R05 Page 18 of 132

20 2.0 GENERAL SITE DESCRIPTION AND CHARACTERISTICS 2.1 Site location and description The Darlington Nuclear Site, where it is proposed the OPG New Nuclear at Darlington (NND) would be built, is located on the north shore of Lake Ontario, about 65 km east of the City of Toronto, in the Municipality of Clarington, Region of Durham in Ontario, Canada. The site is at latitude 43º 53' north and longitude 78º 43' west. The entire site (including Darlington Nuclear Generating Station - DNGS) comprises a land parcel of about 485 hectares. Maps showing the site with the neighbouring area are provided as Figures and The site is bounded by the South service road to the north and Lake Ontario to the south. To the west, the site is bounded by Solina Road. The St. Marys Cement plant occupies the land east of the site. The site is situated in an undulating to moderately rolling limestone till plain. The previously irregular terrain has been graded in the existing powerhouse area to an elevation of about 100 m. The surface elevation rises towards the north with a mean elevation of 122 m just south of the railway tracks. To the north of the tracks, the ground is irregular ranging from 120 to 128 m elevation. A higher ridge, starting from the shore just east of Raby Head, extends diagonally across the site in a north-westerly direction with levels ranging up to 15 m above the surrounding ground. On the west half of the site, the land slopes from 100 m elevation at the Darlington powerhouse to 130 m at the east end of the site. Offshore from the site, the lake bottom slopes away gradually reaching a depth of 6 m at about 425 m from shore and 14 m at 1.2 km from shore. An undulating terrain usually increases atmospheric turbulence near ground level during times of moderate or strong winds, resulting in lower pollution concentrations at locations near the station. The site is well-supplied by access roads for supply of off-site emergency aid and for ease of evacuation of non-essential personnel in case of an emergency. The site may be easily reached by car or rail. The multi-lane Macdonald-Cartier Freeway (The Highway of Heroes section of the 401) runs east/west, immediately north of the site. There are three controlled exits to the Macdonald-Cartier Freeway (Highway of Heroes), directly through Holt Road and two others less than 3 km on either side of the NGS. Rail access can also be provided by the Canadian National Railway's main line, which bisects the site in an approximate east to west direction. The tracks are fenced and gated across the entire site. A rail siding area has been provided at the east boundary limit of the OPG Property. Figure shows the major access and physical features of the site. P1093/RP/001 R05 Page 19 of 132

21 Figure 2.1-1: Darlington Location. [OPG 2004] P1093/RP/001 R05 Page 20 of 132

22 P1093/RP/001 R05 Figure 2.1-2: Darlington Site with the Surrounding Regions. [OPG 2004] Page 21 of 132

23 Figure 2.1-3: Site Access Locations [OPG 2004] P1093/RP/001 R05 Page 22 of 132

24 THIS PAGE IS INTENTIONALLY LEFT BLANK P1093/RP/001 R05 Page 23 of 132

25 2.2 Programme for meteorological investigation Meteorological and atmospheric monitoring OPG has been gathering on-site weather data at the site since 1991 [OPG 2004]. The meteorological tower at the site is located just north of the facility, just southeast of the intersection of Highway 401 and Holt Road (main access to the site). The tower has no significant obstructions from nearby buildings. Meteorological data available from the site consist of wind speeds at two heights and temperature at one height. Data are stored in a data logger and transmitted to a personal computer every fifteen minutes where the data are stored. A parallel fifteen minute back up of the same data is recorded by the Plant Information (PI) interface server, located at the Pickering Training Centre. There are two anemometers on the monitoring tower, one at the standard ten-metre height, the other at fifty metres. The combined wind speed and direction sensors are RM Young Wind Monitors, model 05305, at both the ten and fifty metre heights. Each set of wind speeds is sampled at a three second interval. The data logger records sixty- and fifteen-minute averaged wind speeds, in km/h, and directions, in degrees, from both anemometers. A Campbell Scientific Model 107 temperature probe measures the air temperature at a height of ten metres (in compliance with IAEA guidance). Although the guidance requires temperature measurement to be taken at the same height as the wind speed and recommend heights of 10 m and 50 m above ground, it should be noted that Meteorological Services Canada (MSC) measures temperature at a height of 2 m. For this reason, the MSC data sets are used here for discussion purposes only. Where data sets are required for calculation purposes, the on-site meteorological data measurement have been used. The temperature is sampled at a three-second interval and the data logger records sixty-minute and fifteen-minute averaged temperatures. The fifteen-minute averaged data sets are used in dispersion models for short duration release consequence prediction and the sixty-minute averaged data sets are used for prolonged and long term release modelling. Between 2001 and 2004, the observed wind direction data from the site meteorological tower appears to show a slight shift in direction 1 from north to west in comparison with past historical and MSC data for the same period. A performance assessment of the on-site meteorological tower has been carried out and the root cause rectified; the subsequent data collection, post 2005 has not shown such deviations. For this reason, the weather data sets for the five-year period from 1996 to 2000 have been used for this assessment as the most recent contiguous five-year 1 Sensitivity analysis was carried out on the data and the difference in accident impact results due to use of weather data were found to be approximately 3%. P1093/RP/001 R05 Page 24 of 132

26 period of data. Recent annual data observations show that this basis is supportable and that future five-year data sets will be consistent with this. Other than the monitoring station in Darlington, meteorological data were collected from Environment Canada maintained stations [MSC 2008] that were present within 100 km from Darlington as follows: Daily Data Belleville Bloomfield Cobourg STP Oshawa WPCP Thornhill Grandview Toronto Island Toronto Lester B. Pearson International Airport Hourly Data Peterborough Airport Central Toronto (Precipitation post 2000) Toronto Island Toronto Lester B. Pearson International Airport Trenton Airport (Precipitation Post 2000) Trenton Municipal Airport Bowmanville Mozert (Precipitation only) Meteorological data has been collected from these nearby MSC stations over the period for which Darlington Meteorological Station data sets are in question. Given that the meteorological data averages for Trenton, Toronto Island, Toronto (Pearson) and Peterborough stations, in particular, are similar to those of the data averages for the Darlington meteorological tower, it is considered that these four locations form a representative set for current meteorological data at the site. The proposed location of the land fill from the new building locations is to the east of the existing tower location and is built up to a maximum height of 140 m. The guardhouse to the south will be relocated during construction and it may be necessary to consider the relocation of the meteorological tower at the same time, purely for logistical purposes. At that time, consideration should be given to ensuring that the tower is not relocated to a position where building wake effects will influence the meteorological data obtained. Objective S2.12 of the [IAEA 2002b] states that for the programme for meteorological investigation: The meteorological data collected should be compatible in terms of their nature, scope and precision with the methods and models in which they will be used in P1093/RP/001 R05 Page 25 of 132

27 evaluating the radiation exposure of the public and the radiological impact on the environment for assessment against each regulatory objective. Humidity, air pressure and precipitation are currently not logged on-site by the meteorological tower; however the information is readily available from Environment Canada stations as listed above Regional Climatology The meteorology in the vicinity of site is affected by meso-scale factors consisting of the general circulation of air masses and the effects of the Great Lakes, and microscale factors that include off-shore winds (for coastal areas due to diurnal temperature changes), terrain and topography. Meso-scale factors affect meteorology beyond 10 km from the point of interest. Micro-scale factors affect weather within 10 km from the point of interest Temperature The area has warm summers and cold winters. The mean annual temperature at the site is 7.9 C. Mean daily temperatures fall below zero in the winter months (December through March). The coldest recorded hourly temperature measurement was C, whereas the warmest recorded hourly temperature was 31.5 C. The coldest month was January, with mean daily temperatures in the vicinity of -5.5 C. Summer temperatures averaged 17.7 C, or higher. The highest daily mean temperature recorded was 20.0 C, which occurred in July. Site temperature data from the period of are summarized in Table The data sets are consistent with temperatures recorded at the MSC stations in the surrounding regions for the same and subsequent periods Precipitation Wind Precipitation is quite uniformly distributed throughout the year (see Table 2.2-2) with a slight maximum in the fall season. Total annual precipitation averages about 850 mm, of which about one-tenth occurs as snowfall [OPG 2004]. The greatest 24 hour rainfall recorded in August during the period of is 81.2 mm and the greatest 24 hour snowfall recorded in January is 29 mm (as water). These results are based on data recordings taken at Bowmanville Mozert station, the nearest measuring point to the site between 1971 and Precipitation data sets post 2000 are available from Trenton and Toronto (Pearson). Results from 2000 to 2006 are presented in [AMEC NSS 2009a]. Also a comparison of the precipitation values recorded in the nearby stations has been provided in Table for a quick comparison. Table to Table provide the joint frequencies of wind speed and wind direction from hourly data collected from the on-site 50 m meteorological tower for P1093/RP/001 R05 Page 26 of 132

28 each stability class (See Section 2.2.7). The summary of the wind speed frequencies is provided in Table The data were measured at the 10 m height. The wind directions indicated are the directions from which the wind is blowing. The prevailing winds at the site blow from the west, northwest or southwest over 45% of the time. In addition, there is a high frequency (55%) of light winds (<3 m/s or 10.5 km/h). Winds are lightest in the summer and strongest in the winter and spring Lake Effect The proximity of the station to the lake affects the meteorology near the plant due to the lake effect. Lake effect results from temperature differences between land and water. A Thermal Internal Boundary Layer (TIBL) phenomenon develops as a result of this. In the spring and summer, when the skies are clear and the geostrophic winds are light, a strong temperature gradient between the air over the land and the air over Lake Ontario begins to form in the morning. This is because the air over the ground is heated more rapidly to a higher temperature than the air over the water. The cooler air over the water is denser and flows inland at low levels, producing a lake breeze. At night, the situation is reversed. The land and the air above it cool faster than the water and the air above. The denser air is then over land and the low-level air flows from the land to the lake, creating a land breeze. The lake breeze usually has a much greater intensity than the land breeze. In the fall and winter, the lake is generally warmer than land resulting in more frequent land breezes. In warm seasons, due to solar heating, the air over land is often warmer than that over water. When cold stable lake air flows over warmer land, the resulting upward heat flux gives rise to a TIBL. This TIBL grows in depth with distance inland as the stable air is advected over land and adjusts to changes in surface roughness, heat, and moisture input. The depth of the TIBL is typically hundreds of metres and extends about 10 km inland before a new equilibrium is reached. For emission sources near the ground, pollutants emitted into the unstable boundary layer would also result in higher than expected ground level concentrations during on-shore flows because the stable layer aloft would limit vertical diffusion (compared to the situation when there is no TIBL) Atmospheric Stability The most commonly used method of categorizing the amount of atmospheric turbulence present is categorization of turbulence into six stability classes. The turbulent nature of the atmosphere strongly affects the concentration of contaminants downwind of the release point. A highly turbulent atmosphere is referred to as Stability Class A and occurs under warm sunny conditions. A highly stable atmosphere is referred to as Stability Class F and occurs typically under night-time conditions and low wind speeds. A neutral atmosphere, referred to as Stability Class D, is representative of average turbulence conditions and occurs typically under cloudy, P1093/RP/001 R05 Page 27 of 132

29 windy conditions. All other things being equal, downwind contaminant concentrations for cold ground-level releases are highest when the atmosphere is highly stable (F-stability) and lowest when the atmosphere is highly unstable (A-stability). Various schemes have been developed for predicting stability class. A widely accepted method and one used by OPG is the Sigma Theta (σθ) method. It is based primarily on the standard deviation of continuous measurements of wind direction, but also on the time of day and wind speed [Ontario Hydro, 1984]. The meteorological data used to predict stability class are measured at the 10 m and 50 m elevations of the on-site tower. The meteorological assessment that qualifies the site as acceptable to build a new nuclear power plant is contained in Site Evaluation Report on Meteorological Events [AMEC NSS 2009a]. Table 2.2-1: Data collected at the Darlington Site from [AMEC NSS 2008a] Daily Mean Mean Months (1996- Mean Daily Max Daily Min 2000) ( C) ( C) ( C) January February March April May June July August September October November December Year P1093/RP/001 R05 Page 28 of 132

30 Table 2.2-2: Data collected at the Bowmanville Station from [OPG 2004] Monthly Average Daily Extremes Month Precipitation Rain Snow Precipitation Rain Snow (mm) (mm) (mm) (mm) (mm) (mm) Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec Annual Table 2.2-3: Precipitation data from nearby sites [OPG 2004] Precipitation (mm) Month Toronto Oshawa Port Hope January February March April May June July August September October November December Year P1093/RP/001 R05 Page 29 of 132

31 Table 2.2-4: Triple Joint Wind Frequencies Stability Class A [AMEC NSS 2008a] Wind Direction (wind blowing from) Wind Speed, u (m/s) u 2 2 < u 3 3 < u 4 4 < u 5 5 < u 6 u > 6 Total Frequency (%) at 10 m Height N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW Total A P1093/RP/001 R05 Page 30 of 132

32 Table 2.2-5: Triple Joint Wind Frequencies Stability Class B [AMEC NSS 2008a] Wind Direction (wind blowing from) Wind Speed, u (m/s) u 2 2 < u 3 3 < u 4 4 < u 5 5 < u 6 u > 6 Total Frequency (%) at 10 m Height N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW Total B P1093/RP/001 R05 Page 31 of 132

33 Table 2.2-6: Triple Joint Wind Frequencies Stability Class C [AMEC NSS 2008a] Wind Direction (wind blowing from) Wind Speed, u (m/s) u 2 2 < u 3 3 < u 4 4 < u 5 5 < u 6 u > 6 Total Frequency (%) at 10 m Height N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW Total C P1093/RP/001 R05 Page 32 of 132

34 Table 2.2-7: Triple Joint Wind Frequencies Stability Class D [AMEC NSS 2008a] Wind Direction (wind blowing from) Wind Speed, u (m/s) u 2 2 < u 3 3 < u 4 4 < u 5 5 < u 6 u > 6 Total Frequency (%) at 10 m Height N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW Total D P1093/RP/001 R05 Page 33 of 132

35 Table 2.2-8: Triple Joint Wind Frequencies Stability Class E [AMEC NSS 2008a] Wind Direction (wind blowing from) Wind Speed, u (m/s) u 2 2 < u 3 3 < u 4 4 < u 5 5 < u 6 u > 6 Total Frequency (%) at 10 m Height N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW Total E P1093/RP/001 R05 Page 34 of 132

36 Table 2.2-9: Triple Joint Wind Frequencies Stability Class F [AMEC NSS 2008a] Wind Direction (wind blowing from) Wind Speed, u (m/s) u 2 2 < u 3 3 < u 4 4 < u 5 5 < u 6 u > 6 Total Frequency (%) at 10 m Height N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW Total F Table : Wind Speed Frequencies Grand Total [AMEC NSS 2008a] Wind Speed, u (m/s) u 2 2 < u 3 3 < u 4 4 < u 5 5 < u 6 u > 6 Total Frequency (%) at 10 m Height P1093/RP/001 R05 Page 35 of 132

37 2.2.8 Atmospheric Pressure Hourly observations of atmospheric pressure, which were adjusted to sea level equivalent, were obtained from four MSC weather stations located near the Darlington Nuclear Generating Station. This information is presented in detail in [AMEC NSS 2009a] and indicates a maximum atmospheric pressure of hpa and a minimum of hpa over the period P1093/RP/001 R05 Page 36 of 132

38 2.3 Monitoring Programme for Surface Water and Groundwater Lake Water Lake Currents and Waves Water movement near the site is predominantly along the shore, occurring for 73% of the time (35% to west and 38% to east) [OPG 2002]. Onshore and offshore movement occurs about 15% of the time [OPG 2004]. The mean current speed during the spring-summer-fall periods was 1.8 cm/s. Mean speeds recorded during the winter period were higher, averaging almost 2.8 cm/s. The maximum speed was 56 cm/s, as recorded during the winter of 1982 [Ontario Hydro, 1992]. Water levels The mean water level is at elevation m relative to the International Great Lakes Datum [AMEC NSS 2009b]. Minimum and maximum daily means are estimated at m and m respectively. The site is protected from high lake levels by the protected face of the shoreline which was built to elevation 79 m, 1 m above site grade level and about 3.0 m above the highest water level ever recorded. This shoreline will provide an adequate safety barrier against the severest anticipated combination of spring runoff and wave action [OPG 2004]. Flooding analysis has indicated that for the New Nuclear at Darlington, identified potential flood hazards can all be mitigated through conventional engineering means and methods [AMEC NSS 2009b]. Ice Conditions in Lake Ontario Ground observations of the lake since 1971 near the site revealed no extensive ice cover, only shore and slush ice, which has developed into ridges extending up to 30 m away from the shore [OPG 2004]. Water Temperature Water temperatures have been recorded in Lake Ontario, from 0.2 to 2.2 km offshore from the site, since Water temperatures are discussed in detail in the Surface Water Technical Supporting Documentation of the Environmental Assessment for the New Nuclear at Darlington [Golder 2009b] and the Darlington Safety Report [OPG 2004]. Major up-welling and down-welling events, resulting in temperature changes of over 10 C in a period of several days, were recorded each year during the 17 years the measurements were taken. In a typical year, events of this magnitude occurred about three times during July to September [Ontario Hydro, 1991]. Further information on Lake Hydrology is provided in the Site Evaluation Report on Flood Hazards [AMEC NSS 2009b]. P1093/RP/001 R05 Page 37 of 132

39 The specific requirements of the IAEA Guide NS-G-3.2 have been covered in a variety of reports and are referenced below: The general shore and bottom configuration in the region, and unique features of the shoreline in the vicinity of the discharge. Data on bathymetry out to a distance of several kilometres, and data on the amount and character of sediments in the shallow shelf waters: Site Evaluation Report on Flood Hazards [AMEC NSS 2009b] Speeds, temperatures and directions of any near shore currents that could affect the dispersion of discharged radioactive material. Measurements should be made at appropriate depths and distances, depending on the bottom profile and the location of the point of discharge: [AMEC NSS 2009b] The duration of stagnation and characteristics of current reversals. After stagnation, a reversal in current usually leads to a large scale mass exchange between inshore and offshore waters that effectively remove pollutants from the shore zone: [AMEC NSS 2009b] The thermal stratification of water layers and its variation with time, including the position of the thermocline and its seasonal changes: [AMEC NSS 2009b] The load of suspended matter, sedimentation rates and sediment distribution coefficients, including data on sediment movements characterized by defining at least the areas of high rates of sediment accumulation: Aquatic Environment, Existing Environmental Conditions TSD NND Environmental Assessment [Golder & SENES 2009a] The background levels of activity in water, sediment and aquatic food due to natural and artificial sources: Radiological Environmental Monitoring Program [OPG 2008c] Seasonal cycles of phytoplankton and zooplankton, with at least the periods of their presence and cyclical evolutions of their biomass: [Golder & SENES 2009a] Spawning periods and feeding cycles of major fish species: [Golder & SENES 2009a] Groundwater Groundwater Monitoring Program Since 2002, OPG has undertaken a voluntary program for a reconnaissance-level groundwater impact evaluation in the area. These wells are monitored for: Metals (aluminium, antimony, arsenic, barium, beryllium, bismuth, boron, cadmium, calcium, caesium, chromium (Total), cobalt, copper, iron, lead, lithium, magnesium, manganese, molybdenum, nickel, phosphorus, potassium, P1093/RP/001 R05 Page 38 of 132

40 selenium, silver, sodium, strontium, thallium, thorium, tin, titanium, tungsten, uranium, vanadium, zinc, zirconium) Anions (chloride, sulphate, bromide, fluoride, nitrate, nitrite, phosphate) Alkalinity * Organics * (base neutral and acid extractible which includes polycyclic aromatic hydrocarbons and phenolics) Petroleum Hydrocarbons (Fractions F1 (C6 to C10), F2 (>C10 to C16), F3 (>C16 to C34) and F4 (>C34)) BETX (benzene, ethylbenzene, toluene, xylene) PCBs * Radionuclides (Tritium, Sr-90 *, C-14 *, Cs-134, Cs-137, Co-60, K-40, Th series, U series) * Monitoring of alkalinity, organics, PCBs, Sr-90 and C-14 was discontinued in 2008 since no abnormal readings have ever been detected, nor would they be expected to occur. In the beginning of 2007 there were 20 monitoring wells and 8 Solinst multilevel wells. For the Environmental assessment on the site, 76 new monitoring wells have been setup. One multilevel well has 2 intake zones in the bedrock and the other multilevel well has 3 groundwater intake zones in the overburden material. Further information on the most recent groundwater monitoring programme at Darlington Nuclear Site is provided in the Geological and Hydrogeological Technical Support Document for Darlington New Build Environmental Assessment [CH2MHill 2009a] and the Site Evaluation Study Report Geotechnical Aspects [AMEC NSS 2009c]. Locations of the Monitoring wells existing in Darlington post-commencement of Environmental Assessment study are provided in Figure on the next page. Multilevel wells are marked with a MW and the standard boreholes are marked with a DN. P1093/RP/001 R05 Page 39 of 132

41 Figure 2.3-1: Map of Darlington Monitoring Wells [CH2MHill 2009a] P1093/RP/001 R05 Page 40 of 132

42 2.4 Population distribution The current population distribution for the region immediately surrounding the site was generated by Statistics Canada using their Geographical Information system (GIS). It is based on the most recent census information (2006 for Canada and 2000 for the USA). The 2006 population data are summarized in [OPG 2009a]. Figure shows the grid used to represent demographic distribution for a 100 km circular zone around the proposed site for OPG New Nuclear at Darlington based on the 2006 area. There are relatively few people living within 4 km of the station, give that within the immediate 8 km radius of the station the area is primarily rural with the exception of the city of Bowmanville. The population increases substantially beyond 40 km with the inclusion of the City of Toronto. For the purposes of Emergency Response planning activities over the lifetime of the proposed OPG New Nuclear at Darlington, it is necessary to consider the projected population distributions. This can be done with some accuracy, for projections of around years, by using municipal plans for land use, projected development and economic growth. This information is based on the municipal land use planning projects proposed in the Growing Durham report [Regional Municipality of Durham 2008]. The resulting land use prediction is shown in Figure The recommended growth scenario [Regional Municipality of Durham 2008] concludes that the majority of land areas north of the Darlington Nuclear site will be assigned for future employment (industrial) between 2031 and The majority of residential growth to 2031 is expected to be within the current urban areas of Courtice and Bowmanville through increase in population density. The Recommended Growth Scenario has identified land north of the Canadian Pacific Railway corridor in the vicinity of the site as future residential growth areas between 2031 and OPG has, through collaboration and formal input into municipal planning, influenced changes to the land use structure to ensure that Emergency Response capability is not diminished. As a result, the recommended growth scenario supports unimpeded emergency plan implementation through 2031 [OPG 2009c]. It is expected that the emergency preparedness consultative processes will continue and evacuation time estimates will be jointly reviewed and revised as required beyond P1093/RP/001 R05 Page 41 of 132

43 ` Oro - Medonte Ramara NNW Carden N Sommerville Harvey NNE Springwater NW Eldon Fenelon Verulam Dummer NE Whitechurch- Aurora Stouffville Adjala- New Tosorontio W Halton Hills WNW Caledon Essa New Tecumseh Brampton Barrie Innisfil Bradford - West Guillembury King Mississauga Vaughan Etobicoke East Gwillimbury Newmarket Richmond Hill North York Georgina Markham East York York Toronto Uxbridge Scarborough Brock Pickering Ajax ` Mariposa Whitby Oshawa Lindsay Ops Manvers Clarington 4-km 8-km Emily Cavan Hope Smith Peterborough Port Hope 16-km 24-km 32-km South Monaghan Hamilton 40-km Duoro Otanabee Coburg Asphodel Haldimand 60-km Percy Cramahe Colborne Seymour Lake Ontario 80-km Murray Rawdon ENE 100-km Sidney E WSW Milton ESE Oakville Burlington Hamilton Niagara County Orleans County Monroe County SE Glanbrook Stoney Creek SW Grimsby Lincoln Niagara-onthe-Lake St. Catherines West Lincoln Thorold SSW Niagara Falls Erie County Genessee County SSE Kilometers S Figure 2.4-1: Population Distribution Grid for the Area Surrounding the Proposed Darlington Nuclear Site P1093/RP/001 R05 Page 42 of 132

44 P1093/RP/001 R05 Figure 2.4-2: Land Use Prediction to 2031[Regional Municipality of Durham 2008] Page 43 of 132

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46 2.5 Uses of Land and Water in the Region This section addresses Objective Agriculture Agricultural data presented in this report were derived primarily from the Darlington Nuclear Generating Station (NGS) Safety Report [OPG 2004] and the Site Specific Survey Report [OPG 2006]. Inventories of site-specific agricultural data, which are pertinent to the food chain pathway radiological analysis, are vegetable, food crops, livestock and dairy products. The first major category is further divided into four groups based on phenotypic and agricultural transport characteristics. These include: Leafy vegetables Exposed produce Protected produce Wheat and grains Leafy vegetables (e.g.: cabbage and lettuce) have a broad flat leaf surface that may directly intercept deposition material from the atmosphere. In this group, the edible portion of the plant is primarily concerned with vegetative growth (leaves and stems). Exposed produce refers to vegetables and fruits that may also intercept airborne deposition material like leafy vegetables, but the edible surfaces are relatively smaller for deposition interception. The edible portions are typically concerned with reproductive functions (seeds and fruits). Protected produce items are not directly exposed to airborne material because they grow underground, or if above ground, their edible portions are protected by pods, shells, or non-edible skins or peels. The edible portions of the protected produce are typically the reproductive or storage organs. Wheat and grains are similar to protected produce but they are used both as human food and livestock feeds. The second major category includes livestock food items of beef, pork, poultry, mutton, milk, and eggs. These are of concern because of animals grazing on contaminated vegetation or of feeding contaminated plant material to animals Industry There are no industrial plants manufacturing hazardous materials within 8 km of the site, although there is a water treatment plant which stores quantities of chlorine. OPG will maintain the monitoring of the area development and will provide formal input to planning and construction activities to ensure sustainability in the area. St. Marys Cement Company owns the calcining facility and quarry on the property to the east of the site. Routine blasting and the possible detonation of stored explosives P1093/RP/001 R05 Page 45 of 132

47 are possible external hazards that could affect the site; however, the design will take this into account. Industry in the Bowmanville community has primarily been light to medium manufacturing. The two largest plants produce rubber products and iron castings. There are three major pipelines near the site. Two pipelines carrying natural gas are located about 10 km from the site and one pipeline carrying refined oil products is located about 8 km from the site. The City of Oshawa is a major centre of heavy manufacturing activity with about 90 manufacturing plants employing over 20,000 people. The largest segment of this employment is associated with the automobile industry. The 1993 Official Plan for the Regional Municipality of Durham [Durham 1993] designates approximately 3100 hectares as industrial lands in the region. For more information, refer to the Darlington Safety Report [OPG 2004] and the Site Specific Survey Report [OPG 2006]. It is proposed that a Durham York waste park will be constructed within Clarington Energy Park to the west of the site on a site of 12 hectares [MOE 2006]. The proposal includes facilities for the management of municipal solid waste, including incinerator and ash-processing facilities. Once it is fully operational, it will have the capacity to process 400,000 tonnes/year of waste Transportation There are 4 forms of transportation available near the site: Rail Canadian National Railway s Toronto to Montreal main line passes through the exclusion area around the Darlington station. An average of 53 trains per day use this track and, for freight trains, each has an average of 72 cars. The rail line is dual track and the type and condition of track are excellent. The track is curved and has a gentle slope. There are no level crossings on the site. Details of rail related hazard analysis are provided in Site Evaluation Report Summary Report: Site Evaluation of External Human Induced Events [AMEC NSS 2009d]. There are no significant hazards from rail incidents that can not be mitigated. Airport There are two airports within a distance of 40 km (Greenbank and Oshawa). Details of hazard analysis from airport related incidents are provided in Site Evaluation Report Summary Report: Site Evaluation of External Human Induced Events [AMEC NSS 2009d] There are no significant hazards from aircraft incidents that can not be mitigated. Water - A large number of ships, from small pleasure craft to large lake and ocean vessels ply Lake Ontario. Winter conditions limit this traffic to about 8 months of the year. The larger cargo vessels move along shipping lanes, which are designated by the Ministry of Transport, and the nearest approach is about 27 km from the site. The ports of Oshawa, Whitby, and Cobourg are nearest to the site and receive small lake P1093/RP/001 R05 Page 46 of 132

48 2.5.4 Fishing vessels, gasoline being the most dangerous cargo carried. St. Marys Cement Company Limited owns a pier to the east of the site. Large lake vessels are able to use this dock to load cement or unload gypsum/coal. There are no significant hazards from water related incidents that can not be mitigated. Details are provided in Site Evaluation Report Summary Report: Site Evaluation of External Human Induced Events [AMEC NSS 2009d]. Road - The nearest major road is the Macdonald-Cartier Freeway (the Highway of Heroes section of the 401). Details of road-related hazard analysis is provided in Site Evaluation Report Summary Report: Site Evaluation of External Human Induced Events [AMEC NSS 2009d]. There are no significant hazards from road related incidents that can not be mitigated. Major cold water streams are along the northern shore of Lake Ontario. These streams are concentrated within 50 km of the site. The major species of fish found either inhabiting or migrating up these streams during the spawning season are the salmonoids: coho salmon, Chinook salmon, rainbow trout, brown trout, and lake trout. Highest fishing activities are found at the branches of these streams. Sport fish likely to be caught and consumed in lakeshore area include rainbow trout, salmon, lake trout, brown trout, and walleye. They are highly mobile, migrate about the entirety of Lake Ontario, are unlikely to be strongly attracted to any feature of the lake bottom diffuser, and feed primarily on schooling fish species that are equally migratory and therefore would not be exposed to the discharge for extended periods [Barry Myler, 2008]. For more information, refer to the Darlington Safety Report [OPG 2004] and the Site Specific Survey Report [OPG 2006] Biological Data The biological data that have been collected in and around the region are collated in the Site Specific Survey Report [OPG 2006] and the REMP Report [OPG 2008c]. These data have been used in the pathway analysis. For further details, refer to Section 3.3 and Section 3.4 of this report Baseline ambient radioactivity and Pre-existing Hazardous substances There has been an ongoing program to monitor the radioactivity in the environment in the region surrounding Darlington. It is summarized in detail in the Radiological Environmental Monitoring Program [OPG 2008c]. It covers: External Gamma dose rates in the air Tritium Oxide and C-14 concentrations in air Tritium Oxide concentrations in precipitation and gross beta activity in deposition rates via dry/wet fallout Tritium Oxide, C-14 and I-131 concentrations in milk P1093/RP/001 R05 Page 47 of 132

49 Tritium oxide and gross beta activity concentration in drinking water and lake water Tritium Oxide, Organically bound Tritium (OBT), C-14 and gamma emitting radionuclide concentrations in aquatic sample (e.g.: fish (HTO, C-14, gamma), lake sediment (C-14, gamma), lake water and beach sand (gamma).) Tritium Oxide, gamma and C-14 concentrations in terrestrial samples (e.g.: fruit (HTO, C-14), vegetable (HTO, C-14) and honey (gamma).) Figure displays the sampling locations used in this monitoring program. In 2007 an additional study of Baseline Assessment of Radiation and Radioactivity [AMEC NSS 2009e] was initiated to complement the Radiological Environmental Monitoring Program [OPG 2008c] that was already in place by collecting additional samples to ensure that all feasible human consumption pathways are included and extending the number of analyzed radionuclides. REMP sampling program [OPG 2008c] has collected samples over a period of in excess of ten years. Additional sampling has been conducted for the OPG New Nuclear Environmental Assessment to collect additional radionuclide data for the specified reactor types. Triplicate samples and field and trip blanks have been collected over a period of 8 months and the sampling program is on-going. The quantity of baseline data will be sufficient to conduct a meaningful comparison with the future monitoring data. P1093/RP/001 R05 Page 48 of 132

50 Figure 2.5-1: Darlington Environmental Monitoring Locations [OPG 2008c] P1093/RP/001 R05 Page 49 of 132

51 3.0 ASSESSMENT OF DISPERSION OF RADIOACTIVE MATERIALS FOR NORMAL OPERATIONS 3.1 Radioactive Source Parameters for Normal Discharges to Air Objective 1 is addressed in this section. Airborne radioactive materials will be generated during the normal operation of the OPG New Nuclear at Darlington. Airborne materials discharges could come from the following areas: Reactor building Used fuel storage bay area (and used fuel reception bays) Decontamination centre D 2 O handling area (ACR-1000) Active ventilation exhausts (e.g.: from solid and liquid waste management, chemical lab, etc.) All active or potentially active gases vapour and airborne particulate contained in the ventilation exhaust flow will be treated through a series of high efficiency filters and monitored before the air is released to the environment via the exhaust stack. The estimated maximum emissions from the three reactor designs are tabulated in Table [OPG 2009d]. The information associated with the projected airborne release parameters from the proposed OPG New Nuclear at Darlington is summarized in Table P1093/RP/001 R05 Page 50 of 132

52 Table 3.1-1: Estimated maximum Airborne Emission to the Environment [OPG 2009d] Emissions from different reactor types (Bq/y) AP1000 (4 units) EPR(3 units) ACR-1000 (4 units) H E+13 H E+13 H E+14 * C E+12 C E+11 C E+12 Iodine Iodine Iodine I E+10 I E+08 I E+07 I E+10 I E+09 Noble gas Noble gas Noble gas Kr-85m 5.33E+12 Kr-85m 1.67E+13 Total 2.36E+14 Kr E+14 Kr E+15 Kr E+12 Kr E+12 Kr E+12 Kr E+13 Xe-131m 2.66E+14 Xe-131m 3.89E+14 Xe-133m 1.29E+13 Xe-133m 2.00E+13 Xe E+14 Xe E+14 Xe-135m 1.04E+12 Xe-135m 1.55E+12 Xe E+13 Xe E+14 Xe E+11 Xe E+12 Ar E+12 Ar E+12 Particulate Particulate Particulates Cr E+07 Cr E+07 Total 1.90E+08 Mn E+07 Mn E+06 Co E+06 Co E+05 Co E+09 Co E+07 Co E+09 Co E+07 Fe E+07 Fe E+06 Sr E+08 Sr E+07 Sr E+08 Sr E+06 Zr E+08 Zr E+06 Nb E+08 Nb E+06 Ru E+07 Ru E+06 Ru E+07 Ru E+04 Sb E+06 Sb E+04 Cs E+08 Cs E+06 Cs E+07 Cs E+06 Cs E+08 Cs E+06 Ba E+07 Ba E+05 Ce E+06 Ce E+06 * From [OPG 2008a]. Corresponds to the maximum emission of Tritium to air without the use of Tritium Removal Facility. The activity of individual radionuclides for the category of noble gas and particulate is not available for ACR-1000 reactor design [OPG 2009d]. P1093/RP/001 R05 Page 51 of 132

53 Table 3.1-2: Airborne Release Parameters [OPG 2009d] Reactor types Parameters AP1000 EPR ACR-1000 Elevation of release point (m) Volumetric emission rate (m 3 /s) Building height (m) Gas temperature (ºC ) Calculations of doses have been done with an ambient air temperature of 20 C consistent with previous assessments [OPG 2008c], which is the highest daily mean temperature as discussed in Section 2. This results in a conservative estimate of doses to the public because it will reduce the height of the air plume due to the decreased buoyancy of the plume and correspondingly increase the contaminant transfer factor (from the source to receptor). 3.2 Radioactive Source Parameters for Normal Discharges to Surface Water and Groundwater This section addresses Objective 2. Liquid radioactive effluent will be generated during the normal operation. The liquid radioactive effluent will be, after appropriate treatment, discharged to Lake Ontario. The location of the effluent discharge, assumed to be the point source for both once-through cooling option and cooling tower option, is illustrated in Figure The estimated maximum radioactive emissions to water from three reactor designs are summarized in Table [OPG 2009d]. It should be noted that the activity of individual radionuclide for the category of gross beta and gamma is not available for ACR-1000 reactor design [OPG 2009d]. The discharge rate and aquatic plume parameters associated with the projected waterborne release from the proposed OPG New Nuclear at Darlington are given in Table and Table Under normal operational conditions, all liquid radioactive materials will be collected and treated prior to discharge to the environment (Lake Ontario). There will be no direct discharge of liquid radioactive materials to groundwater under normal operations. 2 Revision 1 of [OPG 2009d] was used for the assessment. The preliminary analysis shows that using EPR specific building height will result in lower public dose than the previous estimate, i.e., the previous estimate represents a more conservative assessment. P1093/RP/001 R05 Page 52 of 132

54 Table 3.2-1: Estimated maximum waterborne emission to the environment [OPG 2009d] Emissions from different reactor types (Bq/y) AP1000 (4units) EPR(3 units) ACR-1000 (4 units) H E+14 H E+14 H E+15 ** C E+11 * C E+11 * C E+10 Gross betagamma Gross beta-gamma Gross beta-gamma Na E+08 Na E+08 Total 5.60E+10 Cr E+08 Cr E+08 Mn E+08 Mn E+07 Fe E+08 Fe E+07 Fe E+07 Fe E+07 Co E+08 Co E+08 Co E+07 Co E+07 Zn E+07 Zn E+07 W E+07 W E+07 Np E+07 Np E+07 Br E+06 Br-84 0 Rb E+07 Rb-88 0 Sr E+07 Sr E+06 Sr E+06 Sr-90 0 Sr E+06 Sr E+06 Y-91m 1.48E+06 Y-91m 0 Y E+07 Y E+07 Zr E+07 Zr E+07 Nb E+07 Nb E+07 Mo E+07 Mo E+08 Tc-99m 8.14E+07 Tc-99m 1.89E+08 Ru E+08 Ru E+08 Rh-103m 7.30E+08 Rh-103m 2.79E+08 Ru E+10 Ru E+09 Rh E+10 Rh E+09 Ag-110m 1.55E+08 Ag-110m 4.88E+07 Ag E+07 Ag E+06 Te-129m 1.78E+07 Te-129m 6.66E+06 Te E+07 Te E+06 Te-131m 1.33E+07 Te-131m 3.44E+07 Te E+06 Te E+06 I E+09 I E+09 Te E+07 Te E+07 I E+08 I E+08 I E+08 I E+09 I E+08 I Cs E+09 Cs E+08 I E+08 I E+09 Cs E+07 Cs E+07 P1093/RP/001 R05 Page 53 of 132

55 Emissions from different reactor types (Bq/y) AP1000 (4units) EPR(3 units) ACR-1000 (4 units) Cs E+09 Cs E+08 Ba-137m 1.84E+09 Ba-137m 3.64E+08 Ba E+08 Ba E+08 La E+09 La E+08 Ce E+07 Ce E+06 Ce E+07 Ce E+07 Pr E+07 Pr E+06 Ce E+08 Ce E+08 Pr E+08 Pr E+08 All others 2.96E+06 All others 2.22E+06 * No data on waterborne emission of C-14 were provided in OPG report for PWR reactor designs [OPG 2009d]. Based on published data [Kunz C. 1985], the discharge of C-14 in liquid was estimated at 1.0E11 Bq/y (or 3.19E+3 Bq/s) for PWR reactors assuming the total electrical output of 5124 GW year. ** From [OPG 2008a]. Corresponds to the maximum emission of Tritium to water without the use of Tritium Removal Facility. Table 3.2-2: Discharge Rate of Liquid Effluents [OPG 2009d] Reactor type Cooling options EPR (3 AP1000 (4units) units) 3 ACR-1000(4 units) Once through(l/s) Mechanical draft cooling(l/s)) Natural draft cooling(l/s)) Revision 1 of [OPG 2009d] was used for this assessment. It is expected that the impact of using EPR specific discharge rate on the total dose is negligible. 4 It is expected that the impact of using EPR specific discharge rate on the public dose is very small. P1093/RP/001 R05 Page 54 of 132

56 Table 3.2-3: Aquatic Plume Parameters [OPG 2008c] Water plume parameters Value Effluent Once-through cooling 1.15 recirculation factor Cooling tower 1 Proportionality Once-through cooling 7.10E-06 coefficient Cooling tower 3.39E-7 Initial dilution Once-through cooling 5 7 Cooling tower 1 Velocity of flow to West(m/s) 0.10 Fraction of time current to West 0.35 Velocity of flow to East(m/s) 0.16 Fraction of time current to East Initial dilution of a once through cooling system is 7 due to the diffuser. Without a diffuser the dilution is 1. The above results imply that a diffuser is included in the design. P1093/RP/001 R05 Page 55 of 132

57 3.3 Identification of Potential Critical Groups and Their Characterization Selection of Potential Critical Groups Doses received by individual members of the public, as a result of a given radionuclide release, will vary depending on factors such as proximity to the release, dietary and behavioural habits, age and metabolism, and variations in the environment. To determine the radiological effects on members of the public, a concept of critical groups is adopted in this study. A critical group is a fairly homogeneous group of people whose location, characteristics or diet, cause them to receive doses higher than the average received by typical people in all other groups in the exposed population. Based on the Darlington Radiological Environmental Monitoring Program (REMP) review [OPG 2008c], which was conducted in 2008, eleven Potential Critical Groups (PCGs) have been identified in this study. These PCGs were updated for the NND location, summarized in Table and are illustrated in Figure Table : Summary of Potential Critical Groups [OPG 2008c] No. Potential critical group 6 Wind sector (direction to)- distance from OPG NND Note 1 Farm WNW 2.8 km Existing PCG used in REMP 2 Dairy Farm N km The closest Dairy farm to the NND site 3 Industrial (St. Marys Cement) NE km Existing PCG used in REMP 4 West East Beach Residents ENE 2.2 km Existing PCG used in REMP 5 Camper W 5.2 km Existing PCG used in REMP 6 Bowmanville Residents NE km Existing PCG used in REMP 7 Fisher E 1.1 km Revised from an existing PCG used in OPG s REMP 8 Oshawa Residents WNW 7.3 km Existing PCG used in REMP 9 Rural Residents NE 1.8 km Existing PCG used in REMP 10 New resident NNW km New identified PCG 11 New industrial NW km New identified PCG 6 The representative location of potential critical group 1-9 was based on a previous work [OPG 2008c]. It is expected that the dose to workers in Clarington Energy Park will be bounded by that for St. Marys Cement plant which is down wind and closest to the NND. P1093/RP/001 R05 Page 56 of 132

58 Location of Outfall Used in the Study Figure : Locations of potential critical groups P1093/RP/001 R05 Page 57 of 132

59 Compared to the existing Darlington NGS s REMP program, the following changes have been implemented: 1. The location of the Fisher Potential Critical Group has been amended to place in the area which is close to the NND, located 1100 m offshore from the new plant. 2. One additional Industrial PCG and one Residential PCG were included to reflect future land use. These two groups represent the future locations of urban residents and industrial site which are the closest to the NND on the basis of the Durham Regional Population Growth estimate [Regional Municipality of Durham 2008]. The locations of these two groups are illustrated in Figure As different age classes may have quite different habits, intake rates, and dose coefficients, people in each PCG are grouped into three age classes as follows [CSA 2008]. one year old infant, 10 year old child, and Adult Characterization of Potential Critical Groups Consistent with the previous work [OPG 2008c, 2008d], the following assumptions were made regarding the characterization of the potential critical groups: All potential critical groups, except fisher group, industrial /commercial group and camper group, live at the location 100 percent of the time. Fisher group only resides at the location for 1 percent of the time. Industrial/Commercial groups are considered to be at the location for only 23 percent of the time since they do not reside at the location. Camper group live at the location for 50% of the time. The fraction of local food intake is as determined by the site specific survey [OPG 2006]. Drinking water is consumed from local sources as determined by the site specific survey. Only dairy farm residents ingest local cow s milk. No local grain products are consumed by humans. Beach recreational activities for all potential critical groups except fisher group and industrial/commercial groups are assumed to take place at Darlington Provincial Park. Detailed characteristics of potential critical groups, including local water and food usage and food and water intake rates, can be found in the site survey, Darlington REMP review report and CSA standard [OPG 2006 and 2008b, and CSA 2008]. It was assumed that the characteristics of two additional groups identified for this study, new resident group and new industrial group, were consistent with those of Bowmanville residents and the existing Industrial (St. Marys Cement) group, respectively. P1093/RP/001 R05 Page 58 of 132

60 3.3.3 Determination of average dilution factor in fish home range For the site evaluation purpose, the concept of home range was adopted to determine the concentration in fish due to waterborne emission. This assumed that fish would only be immersed in average water concentrations over a home range area. A recent study on the fish home range recommended that a 2 km x 2 km area would be a conservative estimate of the species of concern [SENES 2009a]. On this basis, the average dilution factors (DF) in this area were estimated for both once-through cooling and cooling tower options for all three designs with the assumption that the home range was centered at outfall. The equation used to calculate the average dilution factor is given below. The parameters used in the calculation of the average dilution factor are summarized in Table Average DF= 1000 D F ( x) dx [CSA 2008] dβ πκ Where D F (x)= 0.17 QV U C x i + D O 1.17 Table : Parameters Used for the Estimate of Average DF [CSA 2008, OPG 2009d, SENES 2009a] Parameters Once-through Cooling tower Initial dilution, D Average water depth in this range, d, (m) Recirculation factor, β Proportionality coefficient, κ 7.1E E-07 Discharge rate, Q V, ACR (L/s) AP1000/EPR Flow speed, Uc, (m/s) Range, x i (m) The average dilution factors in this range are summarized in Table P1093/RP/001 R05 Page 59 of 132

61 Table : Average Dilution Factors in Home Range Area Designs Once-through Dilution Factor Cooling tower Dilution Factor Dairy cow s diet ACR AP1000/EPR For the site evaluation purpose, it was assumed that the dairy cow s diet consisted of forage and generic feed crops. Furthermore, it was conservatively assumed that the forage accounted for 30% of the dry weight of the total feed consumed by the dairy cow and the generic feed crops accounted for 70% [SENES 2009b]. It was also assumed that the hold up time for forage and generic feed crops was one day and three months, respectively [SENES 2009b]. 3.4 Radioactive Effluent Dose Consequence for Normal Operations Environmental Pathway Model Members of the public could be exposed to radiation and radioactivity via a variety of pathways, which may include: (a) Air inhalation (b) Air immersion (external exposure) (c) Water ingestion (d) Water immersion (external exposure swimming and bathing) (e) Soil external exposure or soil ground shine (f) Terrestrial plant ingestion fruits and vegetables (g) Terrestrial animal product ingestion milk, meat, eggs (h) Aquatic animal ingestion fish (i) Sediment external exposure beach ground shine (j) Incidental soil and sediment ingestion A generalized model of environmental radioactivity transport and human exposure pathways is illustrated in Figure P1093/RP/001 R05 Page 60 of 132

62 \ Figure : Environmental transfer model [CSA 2008] P1093/RP/001 R05 Page 61 of 132

63 Using the concept of compartments, each environmental source/receptor is presented as a numbered compartment and the quantity in compartment is denoted by X i. Transfer from compartment to compartment j is characterized by a transfer parameter P ij and the amount present in compartment j under steady-state conditions due to transfer from compartment to compartment j is therefore P ij X i. The magnitude of the quantity (concentration or dose) represented by any compartment j is X = P j i ij X i Where the summation is over all compartment transfer into compartment j. The compartments and transfer parameters, along with their corresponding units are summarized in Tables and P1093/RP/001 R05 Page 62 of 132

64 Table : Transfer Compartments and their Units [CSA 2008] Compartment Number Compartment Name Units* 0 Source Bq s -1 ** 1 Atmosphere Bq m -3 ** 1a Atmosphere (HTO from HT) Bq m -3 2 Surface Water (lake, river or ocean) Bq L -1 2p Surface Water (pond) Bq L -1 2w Ground Water (well) Bq L -1 3area Soil Surface Bq m -2 3mass Bulk Soil Bq kg -1 dw 3spw Soil Water Bq L -1 4 Forage and Plant Produce Bq kg -1 fw 5 Animal Produce Bq kg -1 fw 5m Mother s Milk *** Bq L -1 6 Aquatic Animals (fish and shellfish) Bq kg -1 fw 7 Aquatic Plants Bq kg -1 fw 8 Sediment Bq kg -1 dw 9 Dose Sv a -1 Notes: * Units involve mass (kg), fresh weight (fw) or dry weight (dw) as indicated. ** For noble gases, source units are BqMeV s-1, atmosphere units are BqMeV m-3. *** CSA N288.1 does not explicitly address mother s milk as a transfer compartment. The parameters for this compartment have been taken from COG DRL Guidance document [Ecometrix Inc. 2008]. P1093/RP/001 R05 Page 63 of 132

65 Table : Transfer Parameters and their Units [CSA 2008] Transfer Parameter From Compartments To Parameter Units P 01 Source Atmosphere s m -3 P 11a Atmosphere (HT) Atmosphere (HTO)* Unitless P 3area1 Surface Soil Atmosphere m 2 m -3 P 3mass1 Bulk Soil Atmosphere kg dw m -3 P 12p Atmosphere Surface Water (pond) m 3 L -1 P 13area Atmosphere Surface Soil m 3 m -2 P 13mass Atmosphere Bulk Soil m 3 kg -1 dw P 13spw Atmosphere Soil Water m 3 L -1 P 14 Atmosphere Forage and Plant Produce m 3 kg -1 fw P 15 Atmosphere Animal Produce m 3 kg -1 fw P(i) 19 Atmosphere Dose (inhalation) Sv a -1 Bq -1 m 3 P(e) 19 Atmosphere Dose (immersion) Sv a -1 Bq -1 m 3 P 02 Source Surface Water (lake, river, ocean) s L -1 P 3spw1 Soil Water Atmosphere L m -3 P 3area2p Surface Soil Surface Water (pond) m 2 L -1 P 3area2w Surface Soil Groundwater (well) m 2 L -1 P 3area3spw Surface Soil Soil Water m 2 L -1 P 3spw2w Soil Water Groundwater (well) Unitless P 3spw2p Soil Water Surface Water (pond) Unitless P 23area Surface Water Surface Soil L m -2 P 23mass Surface Water Bulk Soil L kg -1 dw P 23spw Surface Soil Soil Water Unitless P 24 Surface Water Forage and Plant Produce L kg -1 fw P1093/RP/001 R05 Page 64 of 132

66 Transfer Parameter From Compartments To Parameter Units P 25 Surface Water (lake, river) Animal Produce L kg -1 fw P 2p5 Surface Water (pond) Animal Produce L kg -1 fw P 2w5 Well Water Animal Produce L kg -1 fw P 26 Surface Water Aquatic Animal L kg -1 fw P 27 Surface Water Aquatic Plants L kg -1 fw P 28 Surface Water Sediment L kg -1 dw P(i) 29 Surface Water Dose (ingestion) Sv a -1 Bq -1 L P(i) 2w9 Well Water Dose (ingestion) Sv a -1 Bq -1 L P(e) 29 Surface Water Dose (immersion) Sv a -1 Bq -1 L P(e) 2w9 Well Water Dose (immersion) Sv a -1 Bq -1 L P 3mass4 Bulk Soil Forage and Plant Produce kg dw kg -1 fw P 3mass5 Bulk Soil Animal Produce kg dw kg -1 fw P(i) 3mass9 Bulk Soil Dose (ingestion) Sv a -1 Bq -1 kg dw P(e) 3area9 Surface Soil Dose (groundshine) Sv a -1 Bq -1 m 2 P 45 Forage and Plant Produce Animal Produce kg fw kg -1 fw P 49 Forage and Plant Produce Dose (ingestion) Sv a -1 Bq -1 kg fw P 59 Animal Produce Dose (ingestion) Sv a -1 Bq -1 kg fw P 69 Aquatic Animals Dose (ingestion) Sv a -1 Bq -1 kg fw P 79 Aquatic Plants Dose (ingestion) Sv a -1 Bq -1 kg fw P(i) 89 Sediment Dose (ingestion) Sv a -1 Bq -1 kg dw P(e) 89 Sediment Dose (beachshine) Sv a -1 Bq -1 kg dw Notes: * P 11a is a composite transfer parameter encompassing transfer of HT to soil, oxidation of HT to HTO by soil microbes, re-emission of HTO from soil to atmosphere, and dispersion of HTO in the atmosphere. For C-14 and radioiodine only For HTO only P1093/RP/001 R05 Page 65 of 132

67 3.4.2 Pathway Analysis Model and its Implementation The environmental transfer model discussed above has been programmed in a computer code called Integrated Model for Probabilistic Assessment of Contaminated Transport (IMPACT) [Ecometrix Inc 2006a]. The IMPACT program is a customizable tool that allows the user to assess the transport and fate of contaminants through a user-specified environment. It also enables the quantification of the human exposure to those environmental contaminants and the calculation of DRLs for nuclear facilities (power generating stations, research reactors, waste management facilities). It covers all potential exposure and release scenarios, including atmospheric and aquatic pathways. Quality assurance during code development has followed CSA standard N286.7 [CSA 1999] and CNSC Guideline G-149 [CNSC 2000]. The code verification and validation was documented in Tool Qualification Report [Ecometrix Inc, 2006b]. In this assessment, the IMPACT code (version 5.2.2) was used for the calculation of doses to members of the public. The code is essentially consistent with CSA N , reflecting the updates in scientific developments related to the understanding of environmental transport models and human dosimetry. Detailed discussion of various mathematical models and key model parameters comprising the IMPACT database can be found in CSA N288.1 [CSA 2008] Assessment Results This section summarizes dose predictions for potential critical groups using IMPACT code (version 5.2.2) assuming a life time of sixty years for all designs considered in the assessment. Selection of limiting radionuclides As the report [OPG 2009d] did not provide the breakdown of radionuclides in particulate (air emission) and gross beta-gamma (liquid effluent) for ACR-1000 design, the limiting radionuclide in particulate and gross beta-gamma for this design must be determined before the detailed calculation of dose to the public. Table provides the radionuclides of importance released from CANDU reactor [OPG 2008c], which will be used for ACR reactor for the determination of the limiting radionuclide in air (particulate) and water (gross beta-gamma). It should be noted that this method of using the limiting radionuclide, the radionuclide resulting in the highest dose to the public per unit release, will result in a conservative estimate. P1093/RP/001 R05 Page 66 of 132

68 Table : Radionuclides of Importance Released from ACR Reactor [OPG 2008c]* Radionuclides of importance in particulate in air emission Co-60 Cs-134 Cs-137 Hg-203 Ru-106 Sb-124 Sc-46b Zn-65 Radionuclides of importance in water Gross Beta-gamma emission Co-60 Cs-137 Eu-154 Fe-59 Gd-153 Mn-54 Nb-95 Sb-124 Sb-125 Zn-65 Zr-95 Note: * The report [OPG 2009d] does not provide the list of radionuclides in particulate and gross beta-gamma released from ACR reactor design. The individual radionuclides presented in this table are based on a list of radionuclides released from CANDU reactor with an assumption that the emission of radionuclide from ACR-1000 and CANDU reactor will be similar. To determine the limiting radionuclide to represent gross measurement of particulate in air and beta-gamma in water emission, the doses to each potential critical group selected were estimated using IMPACT assuming a unit release rate (1 Bq/second) of each radionuclide identified above. The results are presented in Table and Table Those limiting radionuclides identified through this exercise will be used in the subsequent dose calculations. P1093/RP/001 R05 Page 67 of 132

69 Table : Limiting radionuclides in particulate released from ACR-1000 reactor Potential Critical Group Age group Limiting Radionuclide Annual Dose (Sv) Bowmanville Residents 1 yr old Co E yr old Cs E-10 Adult Cs E-10 Camper 1 yr old Co E yr old Co E-10 Adult Cs E-10 Dairy Farm 1 yr old Cs E yr old Cs E-10 Adult Cs E-09 Fisher 1 yr old Ru E yr old Ru E-13 Adult Ru E-13 Farm 1 yr old Co E yr old Cs E-09 Adult Cs E-09 New Industry Adult Co E-10 New Residents 1 yr old Co E yr old Co E-10 Adult Cs E-10 Oshawa Residents 1 yr old Co E yr old Co E-10 Adult Cs E-10 Rural Residents 1 yr old Co E yr old Co E-10 Adult Cs E-10 St. Marys Adult Co E-10 West East Beach Residents 1 yr old Co E yr old Co E-09 Adult Cs E-09 Note: + means the contribution from its progeny is included From the preceding table, the one year old child in West East Beach Residents group had the limiting dose from particulate due to Co-60. Therefore, the limiting radionuclide in particulate (airborne emission) released from ACR-1000 reactor design is assumed to be Co-60. P1093/RP/001 R05 Page 68 of 132

70 Table : Limiting Radionuclide in Beta-Gamma Released from ACR-1000 Reactor via Water pathway Annual Potential Critical Group Age group Limiting Radionuclide Dose (Sv) Bowmanville Residents 1 yr old Co E yr old Co E-12 Adult Cs E-12 Camper 1 yr old Zn E yr old Cs E-12 Adult Cs E-12 Dairy Farm 1 yr old Fe E yr old Fe E-12 Adult Fe E-12 Fisher 1 yr old Zn E yr old Cs E-11 Adult Cs E-11 Farm 1 yr old Fe E yr old Fe E-12 Adult Fe E-12 New Industry Adult Cs E-13 New Residents 1 yr old Fe E yr old Fe E-12 Adult Cs E-12 Oshawa Residents 1 yr old Co E yr old Cs E-12 Adult Cs E-12 Rural Residents 1 yr old Fe E yr old Fe E-12 Adult Fe E-12 St. Marys Adult Cs E-13 West East Beach Residents 1 yr old Co E yr old Co E-12 Adult Cs E-12 Note: + means the contribution from its progeny is included From the preceding table, the Adult in Fisher group had the limiting dose from gross beta-gamma in water emission resulting from Cs-137. Therefore, the limiting radionuclide in gross beta-gamma (waterborne emission) released from ACR-1000 reactor design is assumed to be Cs-137. Estimated doses to members of the public Doses to members of the public, represented by potential critical groups, were calculated for three reactor designs using IMPACT based on the emission data described previously. Two cooling options, including once through cooling option and cooling tower option, were considered separately for each design. The results are P1093/RP/001 R05 Page 69 of 132

71 discussed below. It should be noted that the following radionuclides were not included for dose calculation. Co-57 (in particulate) released from AP1000 and EPR accounts for 0.02% and 0.65% of total activity, respectively. It is expected the dose associated with the emission of this radionuclide only accounts for a small fraction of total dose resulting from the emission of particulate and therefore the exclusion of Co-57 for dose calculation has a negligible effect on the total dose estimated. Some waterborne radionuclides including Ag-110, Br-84, Ce-143, Pr-143, Sr-91, Te-129, Te-129m, Te-131, Te-131m, W-187, Y-91m, Y-93, and the category of all others for which the information of individual radionuclides was not provided by vendors were not modeled in IMPACT. The sum of the activity (released from the AP1000 or EPR) from these radionuclides is 0.5% and 1% of the gross beta-gamma activity, respectively. It is expected the doses associated with the emission of these radionuclides only account for a small fraction of total dose resulting from the waterborne emission and therefore the exclusion of these radionuclides for dose calculation has a negligible effect on the total dose estimated. Other radionuclides released to water are Ba-137m, Pr-144, and Rh-106, in total accounting for 35% for AP1000 and 18% for EPR of the total gross betagamma, and have very short half lives (Rh-106: 29.8 s; Ba-137m: 2.5 min; and Pr-144: 17.3 min). Therefore, it is expected that these isotopes will not be available for human intake due to decay to the stable elements (Rh-106 Pd-106, Ba-137m Ba-137, Pr-144 Nd-144) prior to release to the lake and in the environment. Therefore, these three radionuclides will not be considered in the subsequent dose calculations. Once-through cooling option The doses to each of eleven PCG, including three age classes in each PCG, are presented in Table through Table for once-through cooling operation. P1093/RP/001 R05 Page 70 of 132

72 Table : Doses to Potential Critical Groups Resulting from the Operation of AP1000 Reactor (Once-Through Cooling) Dose due to airborne emission(msv) Dose due to waterborne emission(msv) Total annual dose (msv) Potential Critical Group Age group Bowmanville Residents 1 yr old 5.15E E E yr old 3.98E E E-04 Adult 3.83E E E-04 Camper 1 yr old 4.96E E E yr old 3.91E E E-04 Adult 3.74E E E-04 Dairy Farm 1 yr old 4.90E E E yr old 1.93E E E-03 Adult 1.40E E E-03 Fisher 1 yr old 2.43E E E yr old 2.04E E E-05 Adult 1.87E E E-05 Farm 1 yr old 2.37E E E yr old 1.80E E E-03 Adult 1.77E E E-03 New Industry Adult 4.54E E E-05 New Residents 1 yr old 7.28E E E yr old 5.76E E E-04 Adult 5.46E E E-04 Oshawa Residents 1 yr old 3.38E E E yr old 2.71E E E-04 Adult 2.56E E E-04 Rural Residents 1 yr old 7.45E E E yr old 5.69E E E-04 Adult 5.46E E E-04 St. Marys Adult 1.62E E E-04 West East Beach Residents 1 yr old 1.66E E E yr old 1.34E E E-03 Adult 1.28E E E-03 P1093/RP/001 R05 Page 71 of 132

73 Table : Doses to Potential Critical Groups Resulting from the Operation of EPR Reactor (Once-Through Cooling) Dose due to airborne emission(msv) Dose due to waterborne emission(msv) Total annual dose (msv) Potential Critical Group Age group Bowmanville Residents 1 yr old 1.86E E E yr old 1.50E E E-04 Adult 1.54E E E-04 Camper 1 yr old 1.93E E E yr old 1.56E E E-04 Adult 1.57E E E-04 Dairy Farm 1 yr old 7.59E E E yr old 3.78E E E-04 Adult 3.42E E E-04 Fisher 1 yr old 1.82E E E yr old 1.43E E E-05 Adult 1.42E E E-05 Farm 1 yr old 8.05E E E yr old 6.46E E E-04 Adult 6.70E E E-04 New Industry Adult 3.78E E E-05 New Residents 1 yr old 3.30E E E yr old 2.64E E E-04 Adult 2.67E E E-04 Oshawa Residents 1 yr old 1.70E E E yr old 1.37E E E-04 Adult 1.37E E E-04 Rural Residents 1 yr old 3.34E E E yr old 2.66E E E-04 Adult 2.69E E E-04 St. Marys Adult 6.16E E E-05 West Each Beach Residents 1 yr old 1.16E E E yr old 9.13E E E-04 Adult 9.13E E E-04 P1093/RP/001 R05 Page 72 of 132

74 Table : Doses to Potential Critical Groups Resulting from the Operation of ACR-1000 Reactor (Once-Through Cooling) Dose due to airborne emission(msv) Dose due to waterborne emission(msv) Total annual dose (msv) Potential Critical Group Age group Bowmanville Residents 1 yr old 5.42E E E yr old 5.13E E E-04 Adult 5.70E E E-04 Camper 1 yr old 5.30E E E yr old 5.04E E E-04 Adult 5.34E E E-04 Dairy Farm 1 yr old 1.80E E E yr old 1.24E E E-03 Adult 1.26E E E-03 Fisher 1 yr old 4.63E E E yr old 4.30E E E-05 Adult 4.05E E E-04 Farm 1 yr old 2.70E E E yr old 2.46E E E-03 Adult 2.72E E E-03 New Industry Adult 1.12E E E-04 New Residents 1 yr old 9.28E E E yr old 8.71E E E-04 Adult 9.09E E E-04 Oshawa Residents 1 yr old 4.38E E E yr old 4.12E E E-04 Adult 4.29E E E-04 Rural Residents 1 yr old 1.08E E E yr old 9.77E E E-04 Adult 1.03E E E-03 St. Marys Adult 2.91E E E-04 West East Beach Residents 1 yr old 3.62E E E yr old 3.21E E E-03 Adult 3.29E E E-03 The largest estimated dose for the three proposed designs for the once-through cooling option is approximately 5 µsv/y for a one year old infant in the dairy farm group. This value is far below the regulatory dose limit, representing only 0.5 percent of the limit of 1 msv/y. Airborne emission was estimated to be the major contributor to the public dose. Cooling tower option The doses to each of eleven PCG, including three age classes in each PCG, are presented in Table through Table for cooling tower option. Mechanical draft cooling method is used in this study to illustrate the effect of cooling tower P1093/RP/001 R05 Page 73 of 132

75 option. It is expected that the Natural draft cooling method will produce similar results as the only difference between two cooling tower methods, from the dose calculation viewpoint, is the effluent discharge rate which is however similar for the two methods as shown in Table Table : Doses to Potential Critical Groups Resulting from the Operation of AP1000 reactor (Cooling Tower Option) Dose due to airborne emission(msv) Dose due to waterborne emission(msv) Total annual dose (msv) Potential Critical Group Age group Bowmanville Residents 1 yr old 5.15E E E yr old 3.98E E E-04 Adult 3.83E E E-04 Camper 1 yr old 4.96E E E yr old 3.91E E E-04 Adult 3.74E E E-04 Dairy Farm 1 yr old 4.90E E E yr old 1.93E E E-03 Adult 1.40E E E-03 Fisher 1 yr old 2.43E E E yr old 2.04E E E-04 Adult 1.87E E E-04 Farm 1 yr old 2.37E E E yr old 1.80E E E-03 Adult 1.77E E E-03 New Industry Adult 4.54E E E-05 New Residents 1 yr old 7.28E E E yr old 5.76E E E-04 Adult 5.46E E E-04 Oshawa Residents 1 yr old 3.38E E E yr old 2.71E E E-04 Adult 2.56E E E-04 Rural Residents 1 yr old 7.45E E E yr old 5.69E E E-04 Adult 5.46E E E-04 St. Marys Adult 1.62E E E-04 West East Beach Residents 1 yr old 1.66E E E yr old 1.34E E E-03 Adult 1.28E E E-03 P1093/RP/001 R05 Page 74 of 132

76 Table : Doses to Potential Critical Groups Resulting from the Operation of EPR reactor (Cooling Tower Option) Dose due to airborne emission(msv) Dose due to waterborne emission(msv) Total annual dose (msv) Potential Critical Group Age group Bowmanville Residents 1 yr old 1.86E E E yr old 1.50E E E-04 Adult 1.54E E E-04 Camper 1 yr old 1.93E E E yr old 1.56E E E-04 Adult 1.57E E E-04 Dairy Farm 1 yr old 7.59E E E yr old 3.78E E E-04 Adult 3.42E E E-04 Fisher 1 yr old 1.82E E E yr old 1.43E E E-04 Adult 1.42E E E-04 Farm 1 yr old 8.05E E E yr old 6.46E E E-04 Adult 6.70E E E-04 New Industry Adult 3.78E E E-05 New Residents 1 yr old 3.30E E E yr old 2.64E E E-04 Adult 2.67E E E-04 Oshawa Residents 1 yr old 1.70E E E yr old 1.37E E E-04 Adult 1.37E E E-04 Rural Residents 1 yr old 3.34E E E yr old 2.66E E E-04 Adult 2.69E E E-04 St. Marys Adult 6.16E E E-05 West East Beach Residents 1 yr old 1.16E E E yr old 9.13E E E-04 Adult 9.13E E E-04 P1093/RP/001 R05 Page 75 of 132

77 Table : Doses to Potential Critical Groups Resulting from the Operation of ACR-1000 reactor (Cooling Tower Option) Dose due to airborne emission(msv) Dose due to waterborne emission(msv) Total annual dose (msv) Potential Critical Group Age group Bowmanville Residents 1 yr old 5.42E E E yr old 5.13E E E-03 Adult 5.70E E E-03 Camper 1 yr old 5.30E E E yr old 5.04E E E-04 Adult 5.34E E E-04 Dairy Farm 1 yr old 1.80E E E yr old 1.24E E E-03 Adult 1.26E E E-03 Fisher 1 yr old 4.63E E E yr old 4.30E E E-03 Adult 4.05E E E-03 Farm 1 yr old 2.70E E E yr old 2.46E E E-03 Adult 2.72E E E-03 New Industry Adult 1.12E E E-04 New Residents 1 yr old 9.28E E E yr old 8.71E E E-03 Adult 9.09E E E-03 Oshawa Residents 1 yr old 4.38E E E yr old 4.12E E E-04 Adult 4.29E E E-03 Rural Residents 1 yr old 1.08E E E yr old 9.77E E E-03 Adult 1.03E E E-03 St. Marys Adult 2.91E E E-04 West East Beach Residents 1 yr old 3.62E E E yr old 3.21E E E-03 Adult 3.29E E E-03 For the cooling tower option, the estimated bounding dose is approximately 5 µsv/y, representing 0.5 percent of the dose limit of 1 msv/y. The critical group who will be exposed to such a dose is a one year old infant in the dairy farm group. As for the once-through cooling option, airborne emission was estimated to be the major contributor to the public dose for the cooling tower option. It should be noted that the value of home range used to derive the average dilution factor for that area is very conservative. The sport fish such as salmon, lake trout, rainbow trout, brown trout, and walleye are so highly mobile, migrate about the entire Lake and are unlikely to be strongly attracted to any feature of the lake P1093/RP/001 R05 Page 76 of 132

78 bottom diffuser. This conservative assumption could result in the overestimate of the dose to all groups who consumed fish caught locally Sensitivity Analysis Sensitivity analyses were conducted for all three designs to account for uncertainty in public dose associated with some condition changes in the future. The factors considered in the sensitivity analysis include: Temperature change Hydrological condition (change of Lake Ontario current speed) Changes in human habits The annual dose due to airborne emission is estimated to be 4.9E-3 msv for both the once through cooling option and cooling tower option. However, the annual dose due to waterborne emission is higher for the cooling tower option (3.11E-6 msv) than the once through cooling option (3.48E-7 msv). Overall, the doses from the cooling tower option are slightly higher than from the once through cooling option. Therefore for the sensitivity study, the cooling tower option was used for all cases as this is the most conservative option. The results of the sensitivity analyses are summarized in the following paragraphs. It should be noted that for each of three factors considered, the discussion will focus on the critical group who is expected to receive the highest total dose, taking into account all three reactor designs. Temperature change To investigate the effect of temperature change on the public dose, it was assumed that the ambient air temperature would increase to 25ºC from 20ºC in the future. The estimated bounding doses are summarized in Table As shown in the table, when the temperature increases, the dose resulting from airborne emission will increase from 4.90E-3 msv/y to 5.84E-3 msv/y. This is expected because the elevated ambient air temperature will result in the decrease of the height of the air plume due to the decreased buoyancy of the plume; correspondingly, the contaminant transfer factor (from the source to receptor) will increase. However, the results show that the total dose will continue to represent a small fraction of the regulatory limit of 1 msv/y. P1093/RP/001 R05 Page 77 of 132

79 Table : Effect of temperature change on public dose Temperature Dose due to airborne emission (msv/y) Dose due to waterborne emission (msv/y) Total doses (msv/y) 20ºC 4.90E E E ºC 5.84E E E-03 Flow rate The current flow rates in Lake Ontario could be changed in the future. The effect of current velocity changes on public dose was investigated. For the sensitivity analysis, it was assumed that the flow rate would increase by 100% (i.e.: the velocity of flow to west and east was assumed to increase to 0.2 m/s and 0.32 m/s respectively, compared to 0.1 m/s and 0.16 m/s used in the previous analysis). The estimated bounding doses (the corresponding critical group is adult in fisher group) are summarized in Table Table : Effect of flow rate change on public dose Current flow rate Dose due to airborne emission (msv/y) Dose due to waterborne emission (msv/y) Total doses (msv/y) Current flow rate of 0.1 m/s and 0.16 m/s 4.05E E E-03 Current flow rate of 0.2 m/s and 0.32 m/s 4.05E E E-03 As shown in the table, the dose to the critical group (adult in fisher group) due to waterborne emission would increase should the flow rate increase. This is corresponding to the decreased dilution factor. However, the total dose is still well below the regulatory limit of 1 msv/y. Hypothetical groups Two hypothetical groups were designed to ensure that conservative estimates of exposure are derived to provide an upper bound value of doses incorporating feasible changes in potential group location and habits over the next 100 years. The hypothetical groups included one dairy farm group and one urban resident group. It was assumed that both hypothetical groups lived in the vicinity of the West East Beach. This location was selected as the closest to the point of emission beyond St Marys quarry site within the eastern wind sector on the basis of assessment P1093/RP/001 R05 Page 78 of 132

80 summarized in Section 3.3. It has been estimated that doses to the public will be higher in the eastern wind sector due to predominant wind direction. Although no dairy farming or urban resident group is located in this area at present and there are no existing plans that would result in such changes, this sensitivity test was designed to ensure that unknown future changes are accounted for. The two hypothetical groups were assumed to reside at this location 24 hours a day, 365 days a year. Their intake of local food and water were consistent with the results of site specific survey for this population type thus representing realistic consumption habits. The maximum doses for this assessment for the most exposed age class, taking into account all three reactor designs, are presented in Table Table : Doses to hypothetical groups Potential critical group Age class Dose due to airborne emission (msv/y) Dose due to waterborne emission (msv/y) Total doses (msv/y) Hypothetical urban resident 1 yr old 2.96E E E-03 Hypothetical dairy farm 1 yr old 1.76E E E-02 From the table, the highest dose is less than 0.02 msv/y received by a one year old infant in the hypothetical dairy farm group. This accounts for approximately 2 % of the dose limit of 1 msv/y. It should be noted that this dose represents a very conservative bounding estimate which is based on the assumption of having a dairy farming community in the maximum exposure location at some point in the future. It has thus been demonstrated that even under this artificial scenario individual doses to the public will be below the limit. 3.5 Radioactive Effluent Dose Consequence for Refurbishment and Decommissioning Decommissioning of a nuclear facility involves the following broad phases: System drying and decontaminate Isolate for lay up state Disassemble and remove waste Monitor for radiological impact, decontaminate and release to the environment in the end state as per regulatory acceptance criteria P1093/RP/001 R05 Page 79 of 132

81 Decommissioning activities can have radiological safety implications for occupational radiation exposures, environmental emissions, waste generation and disposal resulting in public dose. The extent of the radiological safety implications during decommissioning depends upon the specific decommissioning phase, the facility design, operating history, and magnitude and characteristics of the radiological source term when facility was shut down at the end of plant life. In addition, it greatly depends upon the technology employed to disassemble, decontaminate and remove radiological components. Generally, the radiological source term in a nuclear power plant during the decommissioning phase or refurbishment will be significantly less compared to operating plant conditions as the spent fuel has been removed for interim storage and no new source term is being produced resulting from fission of nuclear fuel in the reactor. Noble Gases, C-14, Radio-Iodine s, Tritium, Fission and Activation products are all subject to decay in the period between reactor shut-down and the time when a particular decommissioning or refurbishment phase takes place. By the end of the dismantling and site restoration phases, the site will be free of industrial hazards. All radioactive contamination in excess of the established levels and other hazardous material will be removed from the site. The site will meet the criteria established by the CNSC according to the license requirements. A nuclear power plant design includes a planned phase in the operating life of the plant to undergo a major rehabilitation of the plant components and structures to extend plant s operating life to enhance equipment reliability and safety Radiological Implications from Decommissioning and Refurbishment There are two radiological safety areas that are impacted during any decommissioning and rehabilitation of a nuclear power plant. Public radiation exposure from effluent emissions to air and water Individual and occupational radiation exposures from in plant activities The information presented below covers only planned operations as per the facility design and does not address implications for accidental conditions. The decommissioning activities are carried out at the end of the plant s useful life and after a long period (about 50 years) to allow for radioactive decay of the source term to reduce the radiological hazards. The time selected is such that all short term radionuclides decay away leaving only long lived radionuclides. Nuclear power plant refurbishment takes place after a much shorter shut-down period than in the case of decommissioning, which nevertheless results in the decay of some short-lived radionuclides. The table below provides a comparative analysis of radionuclide inventory during normal operations, refurbishment and decommissioning. Also, the table compares the absence of pathways during decommissioning and rehabilitation activities. Most of the P1093/RP/001 R05 Page 80 of 132

82 pathways will cease to exist and overall radiological emissions will be a small fraction compared to when plant was operating. Generally, this is true for all reactor designs. P1093/RP/001 R05 Page 81 of 132

83 Table 3.5-1: Comparative Analysis of Radionuclide Inventory During Refurbishment and Decommissioning Radionuclide category Facility Life Cycle Phases and Environmental Impact/Emission Plant Operating Prolonged Shut Down for Refurbishment Tritium Yes Yes (Much reduced depending upon the plant type and system performance) Noble Gases Yes None None Radio-Iodine Yes None* None* C-14 (gaseous) Yes Relatively a small fraction None Airborne Particulates Yes Yes Yes Gamma exposure Yes Reduced level due to shutdown conditions * Long lived isotopes are present but inventories are insignificant. Decommissioning Activities Significantly reduced due to decay and system dryness Much reduced due to decay Source Term and Dose Assessment for Refurbishment and Decommissioning Operations From Table it can be seen that both Refurbishment and Decommissioning are associated with lower inventory than reactor operations. However, both refurbishment and decommissioning may involve intrusion into the reactor core thus making normally isolated contaminated components accessible as potential sources for environmental releases. ACR design presents the bounding case for refurbishment as the scope of refurbishment for PWR designs is likely to be limited to replacement of steam generators. While the design life of the PWR steam generators is 60 years, their replacement during plant lifetime cannot be completely ruled out. Steam generator replacement is a well developed process with contamination well contained within the vessels thus giving rise to only minimal releases. Refurbishment of ACR reactors is assumed to be consistent in scope to that carried out in the past for CANDU plants. Other potential refurbishment work might involve replacement of contaminated components within the reactor vault. Environmental releases are largely dependent on the processes implemented for contamination control, decontamination and effluent processing. These have been developed significantly since the first refurbishment at Pickering NGS took place in 1980s and early 1990s and for new ACR refurbishments the resulting releases are likely to be significantly smaller than in the case of Pickering. Refurbishment of Units 3 and 4 at Pickering A Nuclear Generating Station took place between June 1989 (shut down Unit 3) and August 9, 1991 (return to service) and between August 15, 1991 (shut down Unit 4) and March 28, 1993 (return to service). Based on comparative analysis of environmental releases for the period between 1989 and 1993 with post-refurbishment releases (Table 3.5-2), it can be concluded that P1093/RP/001 R05 Page 82 of 132

84 releases to air during refurbishment at Pickering A were 2-3 times higher than during normal operations. Metals, copper and zinc, were being eroded from the condenser tubes made of brass alloy, which, which resulted in abnormally large releases of metals to water from the Pickering Nuclear Generating Station. This abnormal event was independent of refurbishment activities. However Pickering discharges for years 1990 through 1993 provided in Table include releases from operational units as well as those that were refurbished during this period. From Table it can be seen that all radio-iodines and noble gas releases were due to operations of the remaining units and only H-3, particulate and C-14 releases had contribution from the units that were being refurbished. Decommissioning operations leading to releases of radioactive substances will be consistent in scope to a comprehensive refurbishment project involving dismantling and removal of core components such as that carried out at Pickering A. However decommissioning operations involving handling of radioactive materials will be carried out after a significant shut-down resulting in substantial decay of short-lived radionuclides and smaller releases to environment compared to refurbishment. In summary, a bounding assumption can be made that environmental releases during refurbishment will be increased by three and twenty times for releases to air and water respectively for H-3, particulates and C-14, although the actual refurbishment and decommissioning releases taking into account modern technologies for minimizing releases are likely to be substantially lower. The release of noble gases and radioiodines will be negligible. Under normal operations noble gases and radioiodines were estimated to contribute about 80% of the total dose to the critical group (for the bounding release scenario). Assuming two units are being refurbished and two units are in operation, the resulting effective dose during the refurbishment can be estimated as 4 µsv/y. This is based on the 5 µsv/y bounding dose for the once-through cooling option for normal operation of four units, lack of noble gas and radioiodine releases and a bounding assumption that releases of the remaining radionuclides will increase by a factor of 3 and 20 for airborne and waterborne discharges respectively. P1093/RP/001 R05 Page 83 of 132

85 Table 3.5-2: Pickering A Releases to Air During Refurbishment of Units 3 and 4 (Bq) Radionuclide * 2,3, for Average 1993 Current Annual Emissions for PNGS A 7 H E E E E E E+14 I E E E E E E+07 Particulates 2.85E E E E E E+08 Noble Gases, Bq-MeV 4.07E E E E E E+14 C E E E E E E+12 *No information available for the fourth quarter; annual release for 1991 was estimated based on discharge data for January through September OPG 1991a 2 OPG 1991b 3 OPG 1991c 4 OPG 1991d 5 OPG OPG OPG 2007 Table 3.5-3: Pickering A Releases to Water During Refurbishment of Units 3 and 4 (Bq) Radionuclide * 2,3, for Average Current Annual Emissions for PNGS A 7 Tritium Oxide 4.07E E E E E E+13 Gross Beta-Gamma 4.81E E E E E E+09 *No information available for the fourth quarter; annual release for 1991 was estimated based on discharge data for January through September OPG 1991a 2 OPG 1991b 3 OPG 1991c 4 OPG 1991d 5 OPG OPG OPG 2007 P1093/RP/001 R05 Page 84 of 132

86 3.6 Assessment of Consequences from Radioactive Waste and Used Fuel Management The radioactive waste and used fuel management systems include - Liquid waste treatment facility, which may include heavy water processing plant; - Solid waste conditioning and storage facility; - Used fuel storage facility. Operation of these facilities will result in airborne emissions to atmosphere and liquid releases to Lake Ontario. Radioactive releases to air will mainly comprise tritium, C-14 and noble gases. They will be processed through dehumidifiers, activated charcoal delay beds and filters before release. Radioactive releases to water will be mainly comprised of tritium, Carbon 14 and other beta-gamma emitters. All effluents will be collected in holding tanks and released following treatment in facilities typically including filtration, reverse osmosis and sorption. All releases to water will be monitored prior to discharge. Based on studies conducted by OPG [OPG 2009b], a preliminary conclusion can be reached that releases from radioactive waste and used fuel management systems represent a small fraction of releases due to normal operations and that accident consequences from operating such systems are negligible compared to severe accidents considered for nuclear power plant operation. P1093/RP/001 R05 Page 85 of 132

87 4.0 ASSESSMENT OF DISPERSION OF RADIOACTIVE MATERIALS FOR ACCIDENTAL DISCHARGES This section addresses Objective 4. Accidental discharges can result from internally or externally initiated events. Externally initiated man made events are addressed in [AMEC NSS 2009d] and externally initiated natural events are addressed in [AMEC NSS 2009a; AMEC NSS, 2009b; AMEC NSS, 2009c]. The criteria for assessing and addressing the risk from accidental radiological releases for new Nuclear build projects are provided in RD-337 [CNSC 2008b]. As stated in Section 1.0, RD-337 sets Qualitative and Quantitative Safety Goals, in particular: Small Release Frequency (SRF) The sum of frequencies of all event sequences that can lead to a release to the environment of more than Becquerel of iodine-131 (I-131) is less than 10-5 per reactor year. A greater release may require temporary evacuation of the local population. Large Release Frequency (LRF) The sum of frequencies of all event sequences that can lead to a release to the environment of more than Becquerel of caesium-137 (Cs-137) is less than 10-6 per reactor year. A greater release may require long term relocation of the local population. RD-337 establishes requirements regarding containment performance following a severe accident to the effect that Containment maintains its role as a leak-tight barrier for a period that allows sufficient time for the implementation of off-site emergency procedures following the onset of core damage. Containment also prevents uncontrolled releases of radioactivity after this period. In order to satisfactorily demonstrate that the proposed OPG New Nuclear at Darlington will comply with RD-337, (with respect to the impact of protective measures on the public), while also demonstrating compliance with the emergency response requirements of RD-346 it is necessary to develop an accident release scenario model to provide results that can be used to assess the impacts of an accidental release. 4.1 Derivation of Accident Scenario and Release Characteristics Final versions of the Safety Analysis and PRAs have not yet been completed for the reactors taking into consideration Canadian regulatory requirements and specific site characteristics. The EA TSD on Malfunctions and Accidents [SENES 2009c] provides details on the design features in the new reactors which will reduce the likelihood and P1093/RP/001 R05 Page 86 of 132

88 minimize the consequences of any accidents. An assessment was performed to evaluate scenarios corresponding to the RD-337 safety goal release thresholds to assess the potential impact of such release in terms of radiation dose to the public. These are not design basis cases and are intended to be indicative only, to show that the intent of RD-337 with respect to the impact of protective measures on the public can be met. For this, RD-337 Safety Goal Based (RSGB) Releases to the environment were derived and used in dose calculations. The new reactors must comply with the RD-337 safety goals. This sets limits on the performance of the reactors with respect to accident frequency and consequences of off-site releases. The release limits associated with the LRF and SRF goals can be considered to be the maximum that would be expected for a single event of a 10-6 per year or 10-5 per year frequency respectively, for a plant that just meets the RD-337 requirements. A stylized accident radioactive release scenario was created, with a representative relative isotopic abundance based on actual reactor core behaviour. The release to the environment (also referred to as source term) was normalized to each of the SRF and the LRF threshold release values for I-131 and Cs-137, respectively. These releases, the RSGB Small Release and the RSGB Large Release respectively, were used to determine the potential dose to the public by postulating an event in these categories corresponding to the Large and Small Release thresholds. For the purposes of this analysis, the containment leak-tight period required by RD-337 has been represented as a delay in release of 24 hours after reactor shutdown. The process by which the detailed source term used in the dose calculations was derived is described in the following sections. 4.2 Representative Source Term for Radioactive Airborne Releases For the purposes of modeling an airborne radiological release to the public, the source term for each event is obtained from a combination of core inventory (CI) and a set of release fractions (RF). Source = CI x RF The RF is normally applied to a group of chemically and physically similar isotopes. For example, group 6, which is named for Ruthenium, includes not only Ruthenium isotopes, but also Rhodium isotopes. The RSGB cases are derived on the following basis: The reactor with the highest burn up rate will produce the most conservative starting isotope mix for the core inventory. The release fractions are normalised to generate a release that meets the threshold requirements of both the RD-337 SRF and LRF releases to provide the RSGB cases. P1093/RP/001 R05 Page 87 of 132

89 Physical and chemical processes that would normally lead to retention in containment during the assumed 24-hour leaktight period and thereby reduce the quantities available for subsequent release are neglected. It should be noted that the above basis does not directly reflect the reactor designs nor represent a realistic accident event, since the containment designs will provide for mitigation of the release (this has not been assumed for the RSGB cases). No credit is therefore taken for mitigation in this assessment, other than the initial 24 hour delay period prior to release. Based on the above, the BDBA scenario on which to base the isotopic mix for the RSGB cases for OPG New Nuclear at Darlington has been selected on the basis that the event is the largest contributor to the Large Release Frequency Accident (as defined in [NRC 2007]) (contributing 66%). The calculated frequency of this event is well below the RD-337 LRF limit Core Inventory The EPR core inventory corresponds to the largest physical core inventory (i.e., it experiences highest burn-up rate and uses the highest enrichment of the three proposed designs at 100% power). The 60 dose-significant isotopes in the core inventory, as used in the assessment, are presented in Table The remaining isotopes are not considered in the assessment because they do not contribute to dose for one or more of the following reasons: Very short half life; Extremely long half life, in effect, stable; Very small quantity. Table : EPR Core Inventory From Vendor Data (AREVA 2007) Radionuclide Parent MACCS2 7 Isotope Group Half-Life (s) Core Inv. (Bq) EPR Co-58 None E E+00 * Co-60 None E E+00 * Kr-85 None E E+16 Kr-85m None E E+18 Kr-87 None E E+18 Kr-88 None E E+18 Rb-86 None E E+16 Sr-89 None E E+18 7 Information on MACCS2 accident consequence assessment package is provided in Section P1093/RP/001 R05 Page 88 of 132

90 Radionuclide Parent MACCS2 7 Isotope Group Half-Life (s) Core Inv. (Bq) EPR Sr-90 None E E+17 Sr-91 None E E+18 Sr-92 None E E+18 Y-90 Sr E E+17 Y-91 Sr E E+18 Y-92 Sr E E+18 Y-93 None E E+18 Zr-95 None E E+18 Zr-97 None E E+18 Nb-95 Zr E E+18 Mo-99 None E E+18 Tc-99m Mo E E+18 Ru-103 None E E+18 Ru-105 None E E+18 Ru-106 None E E+18 Rh-105 Ru E E+18 Sb-127 None E E+17 Sb-129 None E E+18 Te-127 Sb E E+17 Te-127m None E E+16 Te-129 Sb E E+18 Te-129m None E E+17 Te-131m None E E+17 Te-132 None E E+18 I-131 Te-131m E E+18 I-132 Te E E+18 I-133 None E E+19 I-134 None E E+19 I-135 None E E+18 Xe-133 I E E+19 Xe-135 I E E+18 Cs-134 None E E+18 Cs-136 None E E+17 Cs-137 None E E+17 Ba-139 None E E+18 Ba-140 None E E+18 La-140 Ba E E+18 La-141 None E E+18 La-142 None E E+18 P1093/RP/001 R05 Page 89 of 132

91 Radionuclide Parent MACCS2 7 Isotope Group Half-Life (s) Core Inv. (Bq) EPR Ce-141 La E E+18 Ce-143 None E E+18 Ce-144 None E E+18 Pr-143 Ce E E+18 Nd-147 None E E+18 Np-239 None E E+20 Pu-238 Cm E E+16 Pu-239 None E E+15 Pu-240 Cm E E+15 Pu-241 None E E+17 Am-241 None E E+15 Cm-242 None E E+17 Cm-244 None E E+17 * Although not present in the source term, a value of 1 is assigned to allow the code to include build-up of Cobalt during the decay chain process Derivation of Release Fractions for RSGB Releases RD-337 stipulates the threshold releases to the environment for SRF and LRF events and that containment should prevent releases to the environment, even in the case of a severe accident, for a period of time allowing protective actions to be implemented. This time is taken to be 24 hours. Radioactive material would therefore decay before the release to the environment. This will affect the isotope mix for the RSGB SRF and LRF due to decay during the containment hold-up period. The effect of this can be adjusted for by reverse decay of the I-131 or Cs-137 quantity (for the SRF and LRF cases respectively) and then normalising the release to the predecay quantities of I-131 or Cs-137, as appropriate, to give a release to containment. These will then decay before release to the environment. The quantities (released into containment) for normalisation are shown below: Case Normalising Isotope RD-337 Quantity Released to Environment Starting Quantity Pre-Decay (24 hrs) Released to Environment SRF I x10 15 Bq 1.09x10 15 Bq LRF Cs x10 14 Bq 1.00x10 14 Bq P1093/RP/001 R05 Page 90 of 132

92 The baseline release fractions for the chosen event are presented in Table by MACCS2 fission group. Table also presents the total baseline activity released, for all isotopes in the group for the BDBA event. The baseline release fractions have then been normalised to the pre-decay 8 quantities of 1.09x Bq I-131 (1) and 1.00x Bq Cs-137 (2) for the SRF and LRF cases respectively, as follows: RF ini-131 = Release Fraction for Isotope Group i x (1.09 x10 15 ) (1) Release Fraction for I-131 x Core Inventory of I-131 RF incs-137 = Release Fraction for Isotope Group i x (1.0x10 14 ) (2) Release Fraction for Cs-137 x Core Inventory of Cs-137 Where: RF ini-131 = Normalised Release Fraction for isotope i, normalised to the SRF predecay quantity RF incs-137 = Normalised Release Fraction for isotope i, normalised to the LRF predecay quantity The above equations were used to normalise the baseline release to give the isotope mix release fractions for an RSGB SRF case and an RSGB LRF case release. The resulting RFs and release quantities are presented in Table Note that the quantities released (in Bq) shown in Table are for the group, thus Group 2 includes not only the quantity of I-131, but all the iodine isotopes, and hence the value is greater than 1.00x Bq. Table : Baseline Release Fraction of initial core inventory released to environment as a total for each MACCS2 fission group Group 1 (Xe/Kr) Group 2 (I) Group 3 (Cs) Group 4 (Te) Group 5 (Sr) Group 6 (Ru) Group 7 (La) Group 8 (Ce) Group 9 (Ba) 9.90E E E E E E E E E-03 Quantities released to environment in Becquerels Group 1 (Xe/Kr) Group 2 (I) Group 3 (Cs) Group 4 (Te) Group 5 (Sr) Group 6 (Ru) Group 7 (La) Group 8 (Ce) Group 9 (Ba) 2.4E E E E E E E E E+17 8 The normalisation and reverse decay method is a simplification, however for the generic source term, the methodology is conservative for the SRF because it artificially increases the starting quantities of such short lived isotopes as Tellurium and I-132. For the LRF it has a negligible impact because the half life of Cs-137 is so long that the applied factor is extremely close to unity. P1093/RP/001 R05 Page 91 of 132

93 Table : RSGB Releases Normalised against SRF and LRF Case Fraction of initial core inventory released as a total for each MACCS2 fission group Group 1 (Xe/Kr) Group 2 (I) Group 3 (Cs) Group 4 (Te) Group 5 (Sr) Group 6 (Ru) Group 7 (La) Group 8 (Ce) Group 9 (Ba) SRF Case 5.12E E E E E E E E E-05 LRF Case 4.17E E E E E E E E E-05 Case Quantities Released From Containment In Becquerels Group 1 (Xe/Kr) Group 2 (I) Group 3 (Cs) Group 4 (Te) Group 5 (Sr) Group 6 (Ru) Group 7 (La) Group 8 (Ce) Group 9 (Ba) SRF Case 1.2E E E E E E E E E+14 LRF Case 1.0E E E E E E E E E+14 The RSGB Releases were modeled as a continuous plume, delayed for 24 hours after reactor shutdown, with a duration of 72 hours. This provides a representative scenario for evaluation of potential doses consistent with RD-337 requirements but is not necessarily characteristic of all potential releases for the reactor technologies. 4.3 Source Term for Accidental Radioactive Discharges to Surface Water and Groundwater Accidental waterborne releases from Darlington Nuclear Site to the environment could originate from the following pathways: Accidental surface water discharge Atmospheric fallout from a severe reactor accident Accidental releases to groundwater For the OPG New Nuclear at Darlington, it is assumed that surface water runoff from the nuclear power plant buildings will be collected in the storm water management ponds and then discharged to an existing drainage course or Lake Ontario. A storm water management system will be installed [CH2MHill 2009b] to: 1. Control runoff release rates to levels that may be efficiently accommodated by the management system and the receiving water body; and 2. Store and detain site runoff to ensure discharges meet applicable regulatory water quantity and quality objectives; in line with the Ontario Stormwater Management Planning and Design Manual 2003 [Ontario 2003]. Therefore, the only direct releases to water during accidents would be from airborne releases, and are represented by the impact from fallout to Lake Ontario P1093/RP/001 R05 Page 92 of 132

94 following a severe accident involving the OPG New Nuclear at Darlington as assessed in Section 4.4, which includes ingestion doses from both food and water. In the event of an accident, radioactivity can be accidentally released to the groundwater. This can be mitigated through standard engineering practices for detecting and containing leaks to meet requirements. In addition to any engineered systems, there will be a local lowering of the groundwater table in the vicinity of the nuclear power plant building. This will be due to reduced recharge as well as abstraction of water. Therefore groundwater gradients will be towards the location of the OPG New Nuclear so that any contamination can be safely extracted and treated. 4.4 Offsite Public Dose Consequence for Radioactive Airborne Release Accidents Dose Targets and Limits For the Early or Emergency Phase of an event, Protective Action Levels (PALs) have been specified by the Province of Ontario Nuclear Emergency Plan [Ontario 1999], as shown in Table supported by Health Canada [Health Canada 2003] who have recommended intervention levels, as shown in Table Table 4.4-1: Protective Action Levels [Ontario 1999] Exposure Control Measures* Protective Measure Effective Dose Lower Level Thyroid Dose Effective Dose Upper Level Thyroid Dose Sheltering 1 msv (0.1 rem) 10 msv (1 rem) 10 msv (1 rem) 100 msv (10 rem) Evacuation 10 msv (1 rem) 100 msv (10 rem) 100 msv (10 rem) 1 Sv (100 rem) Thyroid Blocking msv (10 rem) - 1 Sv (100 rem) * Defined in Section 5.3 P1093/RP/001 R05 Page 93 of 132

95 Table 4.4-2: Health Canada Recommended Intervention Levels [Health Canada 2003] Countermeasure Sheltering Evacuation Relocation Stable Iodine Prophylaxis Food Controls Intervention Level (averted dose) 5 msv in 1 day 50 msv in 7 days 50 msv in 1 year (return when <50 msv in a year and < 10 msv in 1 month) 100 msv to thyroid 1 msv from each of 3 food groups For the Late phase of the event, where relocation may be required, [Ontario 1999] sets out dose limits of no more than 20 msv in 1 year for the late dose. Health Canada have a less stringent 50 msv dose in 1 year limit for relocation, as indicated in Table Care should be taken when considering results against these latter values due to the varying limits for differing time periods. The above tables form the limits for assessing the acceptability of risk from the proposed station and the extent required for evaluation of the emergency response plan Dose Consequence Calculations The MACCS2 (MELCOR accident consequence code system) computer code [SANDIA 1990, SANDIA 1997 and SANDIA 1998] is used to estimate the off-site radiological doses and health effects that could result from postulated accidental releases of radioactive materials to the atmosphere. The most recent validated version (Version ) of the computer code known as MELCOR accident consequence code system (MACCS2) was used to assess the off-site consequences of each of the stylized releases. The development of MACCS2 is sponsored by the U.S. Nuclear Regulatory Commission and is carried out by Sandia National Laboratories in conformance with its standard software development procedures [SANDIA 1998, US DOE 2004]. MACCS2 is used for the current OPG Probabilistic Risk Assessments. Therefore, the use of MACCS2 in this assessment enables reasonable comparison of results between the proposed plant and existing OPG stations. MACCS2 models the off-site consequences of a reactor accident that releases plumes of radioactive materials to the atmosphere. After the release, the radioactive gases and aerosols in the plume, while dispersing in the atmosphere, are transported by the prevailing wind. The effects of building wake and release height are taken into account. Radioactive decay and build-up of daughter products are modeled throughout the process and the effects of meteorological conditions, including frequency of such conditions are also modelled. P1093/RP/001 R05 Page 94 of 132

96 Radioactive materials deposited from the plume can contaminate the environment and the population can be exposed to radiation through a number of pathways. In MACCS2, the time period after the accident is divided into three phases: Early phase, Intermediate and long-term phase. The Late dose is derived from the sum of the Intermediate and Long-term phases. The Early phase begins immediately upon the arrival of the first plume and can last up to seven days. MACCS calculates the Committed Effective Dose (CED) with distance for varying periods of residence. The dose is calculated for a member of the most critical group. Along with the dose by distance, the dose by pathway (Cloudshine, Groundshine, Inhalation and Resuspension Inhalation) is also calculated for a member of the critical group who is assumed to continuously reside at a distance of 1 km from the release point. This is a conservative approach relative to any population behaviour. Various mitigative measures can be specified for this phase, including evacuation, sheltering and dose-dependent relocation. The dose to thyroid is also calculated. Following the Early phase, four long-term exposure pathways are modeled in the intermediate and long-term phases: groundshine, resuspension, inhalation and ingestion of contaminated food and ingestion of contaminated drinking water. The intermediate phase is used to represent a period in which post-accident hazard evaluation can be performed. In this case, since no mitigation is being performed, the intermediate phase is set to zero. The long-term phase represents the time period subsequent to the intermediate phase and is set to obtain the required late dose results (in this case to one year and to 50 years). The MACCS2 code calculates the radiological dose that is not avoided by mitigative measures. In this case no mitigative actions are to be modeled, enabling calculation of the potential avoidable dose if immediate evacuation occurred. The release must be defined in terms of the initial inventory of material and the fraction of that material that is released, along with its physical and chemical properties. Many of the input variables in these files use default values based on the reasoning supplied in the MACCS documentation [SANDIA, 1990, SANDIA, 1998]. The source term is described in Section 4.2 above. Location specific parameters have been developed in the first instance from the Darlington NGS Safety Report [OPG 2004]. The physical station parameters and other assumptions made in the model are briefly summarised below. Key Assumptions The mean meteorological conditions were calculated by MACCS based on hourly data collected from the site for the year The following key assumptions have been made in preparing this assessment: P1093/RP/001 R05 Page 95 of 132

97 No mitigation (i.e.: sheltering or evacuation) is taken into account. The doses are calculated assuming a radial grid with 16 sectors centred on each of the 16 compass points. The dose is calculated in the center of each of the elements of the grid. The releases are conservatively assumed to be cold ground level releases. The nearest building downstream of the release (for building wake effects etc.) is assumed to be the existing Darlington Nuclear Generating Station. The release scenario is a single continuous plume lasting 72 hours, modelled as three consecutive 24 hour plumes in MACCS. The CED and the equivalent dose to thyroid have been calculated for the Early RSGB SRF, and the Late RSGB LRF phases. The total CED for the LRF case has also been calculated. The immediate surrounding area is assumed to be low density residential and farmland for the purposes of surface roughness. The population in each grid element is assumed to be 1. This is adequate for determining individual doses. The above assumptions coupled with the source term information were used to form the RSGB scenarios RSGB Dose Consequence Results Early (Emergency) Phase Results Figure shows the 7-day Early committed effective doses to the whole body for the RSGB SRF release. The results are summarised in Table This is the phase which dictates emergency response and shows the PALs for effective dose. Figure shows the same information for the equivalent dose to the thyroid over the same period, and the PAL values for dose to the thyroid. The results are summarised in Table Late Phase Results The Late results are presented for the RSGB LRF cases. Figure shows 1-year and 50-year late mean committed effective doses for a 7-day Early phase, representing the additional dose received (after the first 7 days of the Early phase is complete), over the next 1 year and 50 years respectively, following the event. The results are summarised in Table In addition, Table presents the dose by pathway to a receptor at a distance of 1 km from the point of release for the same period. P1093/RP/001 R05 Page 96 of 132

98 Total Event Results The Total results are presented for the RSGB LRF cases. Table presents the 1-year and 50-year mean total committed effective doses for a 7-day Early phase for the RSGB LRF releases. Again, Table presents the dose by pathway to a receptor at a distance of 1 km for the Total event. P1093/RP/001 R05 Page 97 of 132

99 Table 4.4-3: Variation of Committed Effective Early Whole Body Doses with Distance for RSGB SRF Distance from point of release (km) RSGB SRF Early Effective Whole Body Dose (msv) E E E E E E E E E E E E E E E E-01 P1093/RP/001 R05 Page 98 of 132

100 Table 4.4-4: Variation of Early Equivalent Dose to the Thyroid with distance for RSGB SRF Release Distance from point of release (km) RSGB SRF Equivalent Dose to the Thyroid (msv) E E E E E E E E E E E E E E E E+00 P1093/RP/001 R05 Page 99 of 132

101 Table 4.4-5: Variation of Late Committed effective Whole Body Doses with Distance for RSGB LRF Release Distance from point of release (km) RSGB LRF 1 Year Late Whole Body Dose (msv) RSGB LRF 50 Year Late Whole Body Dose (msv) E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E-01 P1093/RP/001 R05 Page 100 of 132

102 RD337 SRF 3 day Release, 24 hour delay with 7 day EARLY Phase RD337 SRF 3 day Release, 24 hour delay with 7 day EARLY Phase 1000 EVACUATION MANDATORY PAL GREATER THAN 100mSv 100 EVACUATION OPTIONAL/SHELTERING MANDATORY PAL 10 msv to 100 msv 10 Dose (msv) SHELTERING OPTIONAL PAL 1mSv to 10 msv Distance from release point (km) Figure 4.4-1: Variation of Committed Effective Dose with Distance for RSGB SRF Release EARLY 7 Day P1093/RP/001 R05 Page 101 of 132

103 Table 4.4-6: Dose (msv) by Pathway at 1 km from Release Point for RSGB LRF LATE DOSES i Committed Effective Dose by Pathway Thyroid Dose by Pathway Case Groundshine Resuspension Inhalation Total ii Groundshine Resuspension Inhalation Total ii RD-337 RSGB LRF Late Dose (1 year) RD-337 RSGB LRF Late Dose (50 years) TOTAL DOSES i Committed Effective Dose by Pathway Thyroid Dose by Pathway Case Groundshine Resuspension Inhalation Remaining iii Pathways Total Groundshine Resuspension Inhalation Remaining iii Pathways Total RD-337 RSGB LRF Total Dose (1 year Late Phase) RD-337 RSGB LRF Total Dose (50 year Late Phase) Notes: i. Doses are normally rounded to 2 significant figures; therefore sum of pathways may not exactly equal Total. ii. Includes ingestion dose, although the contribution is very small. iii. Remaining pathways include Cloudshine, Inhalation and Ingestion doses. P1093/RP/001 R05 Page 102 of 132

104 RD337 SRF 3 day Release, 24 hour delay with 7 day EARLY Phase EVACUATION MANDATORY PAL GREATER THAN 1Sv THYROID DOSE 1000 EVACUATION OPTIONAL/SHELTERING MANDATORY PAL 100 msv to 1Sv THYROID DOSE 100 Dose (msv) SHELTERING OPTIONAL PAL 10mSv to 100 msv THYROID DOSE Distance from release point (km) Figure 4.4-2: Variation of Dose to Thyroid with Distance for the RSBG SRF Release EARLY 7 Day P1093/RP/001 R05 Page 103 of 132

105 RD337 LRF 3 day Release, 24 hour delay with 7 day EARLY Phase and 1 year late phase RD337 LRF 3 day Release, 24 hour delay with 7 day EARLY Phase and 50 year late phase Dose (msv) Distance from release point (km) Figure 4.4-3: Variation of Committed Effective Dose with Distance for RSGB LRF Release - LATE P1093/RP/001 R05 Page 104 of 132

106 Table 4.4-7: Total Event Committed Effective Doses for RSGB LRF Distance from point of release (km) RSGB LRF Case (msv) TOTAL 1 Year RSGB LRF Case (msv) TOTAL 50 Years E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E-01 P1093/RP/001 R05 Page 105 of 132

107 4.5 Offsite Public Dose Consequence for Radioactive Waterborne Release Accidents As discussed in Section 4.3, consequences from airborne release to water will be bounded by consequences from severe accidents scenarios assessed in Section Impact of Mitigation on Dose Consequence for Accidents The avertable dose is the Early phase dose which would be avoided if the affected population were evacuated. For the Early phase, the RSGB SRF event is used to evaluate the emergency response, since the RD-337RD-337 SRF event is designed to address short term evacuation cases. For an event where the release is held within the reactor containment for 24 hours or more, the evacuated population would not be exposed. In this case, the avertable dose would be the full 7 day early phase dose. In the event of a failure of containment isolation, the time for evacuation and sheltering measures would be shorter, and the avertable dose would be proportional then to the time of exposure, allowing for decay over time, and the time to travel out of the affected area. For the purposes of assessment, the dose calculations extend to a minimum distance of 500 m from the reactor building. From Figure 4.4-1, showing the RSGB SRF Early dose, the lower PAL of 10 msv for evacuation is met for people within approximately 2 km of the point of release. The PAL for mandatory evacuation is 100 msv, which would only occur within the site boundary. The same is true if the thyroid dose and corresponding PALs are considered, as shown in Figure The dose that could be avoided by evacuation would be approximately 25 msv in the 1 km evacuation range and up to around 5 msv out to 3 km. The results indicate that sheltering for an event such as that modeled could be mandatory to around 2 km from the point of release, while the option of sheltering may be considered for out to 10 km. The relocation limit for dose over 1 year is not exceeded beyond 1 km from the point of release, and as such, no permanent residents would require relocation. It is possible that ingestion dose control measures, such as control of milk, livestock and produce would be invoked for a period to minimize long term dose to the population as provided for in the Provincial Nuclear Emergency Plan s Ingestion Control Phase for post-accident mitigation. P1093/RP/001 R05 Page 106 of 132

108 5.0 CONSIDERATION OF THE FEASIBILITY OF AN EMERGENCY PLAN This section addresses Objective 5. From Figure 4.4-1, it can be seen that a dose of around 25 msv might be expected at 1 km from the point of release during the Early phase if no mitigation measures were taken. This section provides an assessment of how the Emergency Plan for Darlington is impacted by the calculated dose consequences. The intent of this section is twofold: To analyze compliance of the existing emergency plans against the IAEA guidance on emergency planning. To provide a judgement regarding the likely impacts of the OPG New Nuclear at Darlington on these plans. 5.1 Objectives for Emergency Planning According to IAEA GS-R-2 ( Preparedness and Response for a Nuclear or Radiological Emergency ), the primary objectives for protection and safety as part of emergency planning are as follows: Protection objective: to prevent the occurrence of deterministic effects in individuals by keeping doses below the relevant threshold and to ensure that all reasonable steps are taken to reduce the occurrence of stochastic effects in the population at present and in the future. Safety objective: to protect individuals, society and the environment from harm by establishing and maintaining effective defences against radiological hazards from sources. Radiation protection objective: To ensure mitigation 9 of the radiological consequences of any accidents. Technical safety objective: To take all reasonably practical measures to prevent accidents in nuclear installations and to mitigate their consequences should they occur; to ensure with a high level of confidence that, for all possible accidents taken into account in the design of the installation, including these of very low probability, any radiological consequences would be minor and below prescribed limits... IAEA GS-R-2 also states that measures taken to achieve these objectives (undertaking interventions) are governed at all times by the following principles: Justification of intervention: Any proposed intervention shall do more good than harm. 9 This is not to imply that the effects would be entirely mitigated through the use of emergency response. Instead, the intent is to maintain doses within the regulatory limits. P1093/RP/001 R05 Page 107 of 132

109 Optimization of intervention: The form, scale and duration of any intervention shall be optimized so that the net benefit is maximized. The objectives of emergency planning are most likely to be achieved in accordance with the principles for intervention by having a sound programme for emergency preparedness in place as part of the infrastructure for protection and safety. Emergency preparedness also helps to build confidence that an emergency response would be managed, controlled and co-ordinated effectively. A detailed description of the general, functional and infrastructure requirements for effective emergency planning and response as listed in this document are available in Appendix B. 5.2 Existing Emergency Plans The Province of Ontario Nuclear Emergency Plan (PNEP) ( Province of Ontario Nuclear Emergency Plan: Part I Provincial Master Plan, Interim Plan, 2nd Edition, Ontario, March 1999) provides the basis upon which nuclear emergency planning, preparedness and response shall be undertaken to safeguard the health, safety, welfare and property of the inhabitants of the province, and to protect the environment in the event of a nuclear emergency. The plan also includes a Darlington specific section ( Province of Ontario Nuclear Emergency Plan: Part IV Darlington Nuclear Emergency Plan, Interim Plan, 2nd Edition, Ontario, December 1998). The PNEP Part I specifies the overall principles, policies, basic concepts, organizational structures and responsibilities. PNEP Part I has been revised and is now called the Provincial Nuclear Emergency Response Plan (PNERP). The PNERP received Cabinet approval at the end of January 2009 and was issued by an Order of Council on February 11th, Subsequently, OPG is now working with the province on implementation of the PNERP. The information used in this report is not affected by the revision. The PNEP divides the area surrounding the plant into zones and sectors, with the level of planning and preparedness being the highest in the regions nearest to the plant. In the event of an accident, projected doses are determined for the most exposed individual in the most critical group. The response to the accident is determined by comparing the projected dose to a set of Protective Action Levels (PALs), for which protective measures have been identified. The type of accident also determines the timing of radioactive releases from the plant to the public, which in turn affects the emergency response to the accident. The area around the boundary of a nuclear installation for which a nuclear emergency plan is made is divided into the following zones: Contiguous Zone (~3 km radius): The zone immediately surrounding the nuclear installation. An increased level of emergency planning and preparedness shall be P1093/RP/001 R05 Page 108 of 132

110 undertaken within this area because of its proximity to the source of the potential hazard. Primary Zone (~10 km radius): The zone around the nuclear installation within which detailed planning and preparedness shall be carried out for measures against exposure to a radioactive plume, including evacuation. (The Primary Zone includes the Contiguous Zone). Secondary Zone (~50 km radius): A larger zone within which it is necessary to plan and prepare measures against exposure from ingestion of radioactive material. (The Secondary Zone includes both the Primary and Contiguous Zones). The specific sector boundary map for the Primary Zone around the Darlington Nuclear Generating Station is shown in Figure The Primary Zone is itself subdivided into 3 sub-zones called rings as follows: Inner Ring (Contiguous Zone) - Sector D1 Middle Ring - Sectors D2, D3, D4, D5, and lake sectors D14 and D15. (D14 and D15 are not shown in the Figure). Outer Ring - Sectors D6, D7, D8, D9, D10, D11, D12, D13, and lake sectors D16 and D17. (D16 and D17 are not shown in the Figure). Sectors D6 and D8 are each divided into subsectors A and B for evacuation planning. Figure 5.2-1: Darlington Specific Response Sectors in Primary Zone P1093/RP/001 R05 Page 109 of 132