NOT PROTECTIVELY MARKED. REDACTED PUBLIC VERSION HPC PCSR3 Sub-chapter 12.3 Nuclear Island Structures NNB GENERATION COMPANY (HPC) LTD

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1 Page No.: i / vii NNB GENERATION COMPANY (HPC) LTD HPC PCSR3: CHAPTER 12 CIVIL STRUCTURES STRUCTURES {PI Removed } uncontrolled. SUB-CHAPTER 12.3 NUCLEAR ISLAND 2017 Published in the United Kingdom by NNB Generation Company (HPC) Limited, 40 Grosvenor Place, Victoria, London SW1X 7EN. All rights reserved. No part of this publication may be reproduced or transmitted in any form or by any means, including photocopying and recording, without the written permission of the copyright holder NNB Generation Company (HPC) Limited, application for which should be addressed to the publisher. Such written permission must also be obtained before any part of this publication is stored in a retrieval system of any nature. Requests for copies of this document should be referred to NNB Generation Company (HPC) Limited, 40 Grosvenor Place, Victoria, London SW1X 7EN. The electronic copy is the current issue and printing renders this document

2 Page No.: ii / vii APPROVAL SIGN-OFF: DOCUMENT CONTROL REVISION HISTORY { PI Removed } { PI Removed } { PI Removed } Text within this document that is enclosed within brackets { } is Sensitive Nuclear Information, Sensitive Commercial Information or Personal Information and has been removed.

3 Page No.: iii / vii TABLE OF CONTENTS 1. FOUNDATIONS (INCLUDING COMMON RAFT) SAFETY REQUIREMENTS ROLE OF THE STRUCTURE DESIGN BASIS DESCRIPTION OF THE STRUCTURE, LAYOUT AND INTERFACES DESIGN SUBSTANTIATION CONSTRUCTION PRE-STRESSING GALLERY SAFETY REQUIREMENTS ROLE OF THE STRUCTURE DESIGN BASIS DESCRIPTION OF THE STRUCTURE, LAYOUT AND INTERFACES DESIGN SUBSTANTIATION CONSTRUCTION REACTOR BUILDING (HR [RB]) SAFETY REQUIREMENTS ROLE OF THE STRUCTURE DESIGN BASIS DESCRIPTION OF THE STRUCTURE, LAYOUT AND INTERFACES DESIGN SUBSTANTIATION CONSTRUCTION INNER CONTAINMENT WALL WITH STEEL LINER SAFETY REQUIREMENTS ROLE OF THE STRUCTURE DESIGN BASIS DESCRIPTION OF THE STRUCTURE, LAYOUT AND INTERFACES DESIGN SUBSTANTIATION... 38

4 Page No.: iv / vii 4.5. CONSTRUCTION CONTAINMENT PENETRATIONS SAFETY REQUIREMENTS ROLE OF THE STRUCTURE DESIGN BASIS DESCRIPTION OF THE STRUCTURE, LAYOUT AND INTERFACES DESIGN SUBSTANTIATION CONSTRUCTION REACTOR BUILDING INTERNAL STRUCTURES SAFETY REQUIREMENTS ROLE OF THE STRUCTURE DESIGN BASIS DESCRIPTION OF THE STRUCTURE, LAYOUT AND INTERFACES DESIGN SUBSTANTIATION CONSTRUCTION AIR PLANE CRASH SHELL AND OUTER CONTAINMENT SAFETY REQUIREMENTS ROLE OF THE STRUCTURE DESIGN BASIS DESCRIPTION OF STRUCTURE, LAYOUT AND INTERFACES DESIGN SUBSTANTIATION CONSTRUCTION STAINLESS STEEL-LINED POOLS, TANKS AND RPE [NVDS] SUMPS SAFETY REQUIREMENTS ROLE OF THE STRUCTURES DESIGN BASIS DESCRIPTION OF THE STRUCTURE, LAYOUT AND INTERFACES DESIGN SUBSTANTIATION CONSTRUCTION... 97

5 Page No.: v / vii 9. FUEL BUILDING (HK [FB]) SAFETY REQUIREMENTS ROLE OF THE STRUCTURE DESIGN BASIS DESCRIPTION OF STRUCTURE, LAYOUT AND INTERFACES DESIGN SUBSTANTIATION CONSTRUCTION SAFEGUARD ELECTRICAL AND MECHANICAL BUILDINGS (HL [SB (E)/SB (M)]) SAFETY REQUIREMENTS ROLE OF THE STRUCTURE DESIGN BASIS DESCRIPTION OF THE STRUCTURE, LAYOUT AND INTERFACES DESIGN SUBSTANTIATION CONSTRUCTION EMERGENCY DIESEL BUILDINGS (HD [DB]) SAFETY REQUIREMENTS ROLE OF THE STRUCTURE DESIGN BASIS DESCRIPTION OF THE BUILDINGS, LAYOUT AND INTERFACES DESIGN SUBSTANTIATION CONSTRUCTION NUCLEAR AUXILIARY BUILDING (HN [NAB]) SAFETY REQUIREMENTS ROLE OF THE STRUCTURE DESIGN BASIS DESCRIPTION OF THE STRUCTURE, LAYOUT AND INTERFACES DESIGN SUBSTANTIATION CONSTRUCTION

6 Page No.: vi / vii 13. RADIOACTIVE WASTE STORAGE AND PROCESS BUILDINGS (HQA & HQB) SAFETY REQUIREMENTS ROLE OF THE STRUCTURE DESIGN BASIS DESCRIPTION OF THE STRUCTURE, LAYOUT AND INTERFACES DESIGN SUBSTANTIATION CONSTRUCTION EXTENSION OF NUCLEAR AUXILIARY BUILDING (HN [NAB]) FOR UNIT 2 (HQC) SAFETY REQUIREMENTS ROLE OF THE STRUCTURE DESIGN BASIS DESCRIPTION OF THE STRUCTURE, LAYOUT AND INTERFACES DESIGN SUBSTANTIATION CONSTRUCTION ACCESS BUILDING (HW) SAFETY REQUIREMENTS ROLE OF THE STRUCTURE DESIGN BASIS DESCRIPTION OF THE STRUCTURE, LAYOUT AND INTERFACES DESIGN SUBSTANTIATION CONSTRUCTION INTERMEDIATE LEVEL WASTES BUILDING (HHI) SAFETY REQUIREMENTS ROLE OF THE STRUCTURE DESIGN BASIS DESCRIPTION OF THE STRUCTURE, LAYOUT AND INTERFACES DESIGN SUBSTANTIATION

7 Page No.: vii / vii CONSTRUCTION HOT LAUNDRY (HVL) AND KER/TER/SEK/PTR TANKS (HXA) SAFETY REQUIREMENTS ROLE OF THE STRUCTURE DESIGN BASIS DESCRIPTION OF THE STRUCTURE, LAYOUT AND INTERFACES DESIGN SUBSTANTIATION CONSTRUCTION HOT WORKSHOP, HOT WAREHOUSE AND FACILITIES FOR DECONTAMINATION BUILDING (HVD) SAFETY REQUIREMENTS ROLE OF THE STRUCTURE DESIGN BASIS DESCRIPTION OF THE STRUCTURE, LAYOUT AND INTERFACES DESIGN SUBSTANTIATION CONSTRUCTION REFERENCES

8 Page No.: 1 / 207 SUB-CHAPTER 12.3 NUCLEAR ISLAND STRUCTURES 1. FOUNDATIONS (INCLUDING COMMON RAFT) This section outlines the Hinkley Point C (HPC) safety justification for the Nuclear Island (NI) Buildings Foundations, including the Common Raft. The overview of the safety justification for civil structures and links to the overall safety case are presented in Sub-chapter HPC site geotechnical and earthworks aspects are described in Sub-chapter The content of this section follows the outline described in Sub-chapter 12.1, section 2. The safety justification for the Common Raft is supplemented in Sub-chapter 3.3 which contains generic justifications applicable to all structures. Furthermore, the Common Raft s safety justification is supplemented by the structures supported by the Common Raft, i.e. the Reactor Building (HR [RB]) (including the internal and outer containments), the Fuel Building (HK [FB]), the Safeguard Electrical & Mechanical Buildings (HL [SB (E)] / [SB (M)]) and the Air Plane Crash (APC) Shell: Sub-chapter 12.3 section 4, for the Inner Containment Wall with Steel Liner, Sub-chapter 12.3 section 6, for the HR Building Internal Structures, Sub-chapter 12.3 section 3, for the HR Building, Sub-chapter 12.3 section 9, for the HK Building, Sub-chapter 12.3 section 10, for the HL Buildings, Sub-chapter 12.3 section 7, for the APC Shell and Outer Containment, Sub-chapter 12.3 section 2, for the Pre-Stressing Gallery. The Pre-stressing Gallery is noted for information only (the structures are not directly linked) SAFETY REQUIREMENTS The Common Raft s safety requirements are requirements that relate to nuclear safety and that shall be fulfilled in order to meet the HPC Main Safety Functions (MSFs) as defined in Sub-chapter 3.2, section Main Safety Functions The MSF (see Sub-chapter 3.2, section 3) relevant to the Common Raft is confinement of radioactive material. The Common Raft protects the general public, workers and the environment in all normal and accident situations by providing a barrier to the release of radioactivity. The Common Raft also provides protection to systems and components housed within the Common Raft buildings.

9 Page No.: 2 / 207 The Common Raft protects multiple systems and components responsible for the MSFs of confinement of radioactivity, fuel heat removal and control of reactivity of fuel; in particular those systems housed within the HR, HK and HL Buildings. Alongside the aforementioned systems, the Common Raft itself provides a barrier to the release of radioactivity into the environment. The Safety Functional Requirements (SFRs) referenced in section shall be met in order to ensure fulfilment of the aforementioned MSFs. As part of the overall safety approach implemented at the design stage for the HPC EPR, the Common Raft must fulfil a dual role. Firstly, it must safeguard the safety classified Structures, Systems and Components (SSCs) which it supports against all the fault and hazard conditions to which they might be exposed, in particular external hazards. Secondly, it must protect the environment against all the fault conditions whose frequency of occurrence cannot be reduced to insignificant levels. It must limit any release of radioactive materials Safety Classification of the Structure The Common Raft is a Safety Class 1 civil structure as it contributes directly to the main MSF of confinement of radioactive material and protects Safety Class 1 equipment and systems. The justification for the classification is given in Sub-chapter 3.2, section 10, using the classification approach defined in the Sub-chapter 3.2, section 8. The Common Raft must therefore meet C1 requirements and meet SC1 seismic requirements (see Sub-chapter 3.2, section 8) Safety Functional Requirements The SFRs applicable to the Common Raft are determined by taking load cases originating from the faults and hazards defined in Chapters 13, 14 and 15 and associating these with the operating situations whilst considering the required functions of the Common Raft. SFRs are applied in order to fulfil the MSFs defined in section The SFRs relate to the required behaviour of the structures, which will be either reversible or irreversible after application of sustained, variable or accidental loads. The structural performance and behavioural criteria are chosen with respect to operating and hazard conditions (normal, exceptional and accidental) (see Sub-chapter 3.3, section 1 for further detail). The general operating and hazard conditions are defined in the civil nuclear code ETC-C and UK Companion Document (UK CD) (see Sub-chapter 3.8, section 4). The expected functions of the Common Raft, after application of the loads, are summarised in a SFR schedule [Ref. 1]. The SFR schedule is developed as a set of detailed safety requirements on individual SSCs which comprise the Common Raft. The SFRs to be applied depend on how the respective SSCs contribute to plant and structural safety, their functional role and the physical arrangement of the SSC. Each SFR is relative to a plant or hazard condition. A set of engineering requirements are listed for each SFR to ensure that the SFR is met through application of civil engineering requirements for the hazard or condition in question.

10 Page No.: 3 / Beyond Design Basis Considerations In addition to the design conditions identified, it must be demonstrated that adequate margins exist to cover hypothetical scenarios beyond the design basis for the Common Raft so that cliff-edge failures do not occur. The Level 3 Design Substantiation Report (DSR) [Ref. 4] will demonstrate deterministically that adequate margins exist and are incorporated into the design of the Common Raft. The following aspects must be considered: description of alternative load paths, ductile design, evaluation of margins in loadings and factors, and first mode of failure. In addition to the DSR, a probabilistic study will be undertaken later in the design process that aims to confirm a Beyond Design Basis (BDB) capacity exists for the Common Raft Operation; Examination, Maintenance, Inspection and Testing (EMIT); and Decommissioning Requirements Sub-chapter 18.1 describes how Human Factors (HF) principles, methods and standards are integrated into the HPC EPR design. The civil engineering design contributes significantly to achieving the HF principles stipulated in Sub-chapter 18.1, section Operation Requirements A description of how operational requirements are considered within the civil engineering design can be found in Sub-chapter 3.3, section EMIT Requirements A description of how EMIT is considered within the civil engineering design can be found in Subchapter 3.3, section Decommissioning Requirements A description of how decommissioning requirements are considered within the civil engineering design can be found in Sub-chapter 3.3, section 1. Depending on the site specific decommissioning strategy chosen and the fuel storage options available after end of generation, it is anticipated that the HK Building will continue to hold spent fuel for a period of time after the end of generation, refer to Sub-chapter 12.3 section 9. It will be ensured that the Common Raft will be able to perform its confinement function and also provide support to SSCs linked to the fuel storage and decay heat removal function during this period.

11 Page No.: 4 / 207 Following dismantling of the SSCs supported by the Common Raft, the HR, HK and HL Buildings, including their respective APC Shells, and clearance monitoring of the facilities, demolition of the Common Raft will be undertaken. The detail of the sequencing will take account of the structural integrity of the buildings during access for demolition of internal walls and waste material removal as necessary, and the maintenance of safe access and egress routes ROLE OF THE STRUCTURE Main Roles of the Structure The main roles of the Common Raft are as follows: To provide stability and prevent differential displacement between the following structures: o o o o HK Building, HL Buildings, HR Building, and APC Shell. To act as a barrier for environmental protection by preventing soil contamination in case of installation failure: o o protect the water table from any risk of contamination, and protect the structures from groundwater Main Components and Systems Supported by the Structure The following is a non-exhaustive list of the main components and systems supported on the Common Raft: Nuclear Vent and Drain System (RPE [NVDS]) (including RPE [NVDS] pipes and sumps). For the equipment contained in the buildings supported by the Common Raft see the following sections: HR Building components and systems (see Sub-chapter 12.3, sections 3, 4 and 6 ), HK Building components and systems (see Sub-chapter 12.3, section 9), and HL Building components and systems (see Sub-chapter 12.3 section 10).

12 Page No.: 5 / DESIGN BASIS The general design basis including performance and behavioural criteria is described in Sub-chapter 3.3, section 1. The Common Raft specific performance criteria and design basis information, which ensure that the specific SFRs (described in section 1.0.3) and design requirements are met, are defined below Design Life The civil engineering design must ensure that the structures perform their safety function throughout the planned life of the plant. UK EPR structures are generally designed with an operating life time of 60 years. In the case of the Common Raft and to take into account the use of the Spent Fuel Pool (SFP) after the end of generation of the plant, this period is extended to 70 years. Additional periods of five years for construction and 15 years for decommissioning must also be taken into account. The detailed design of the reinforced concrete and steel structures against the load cases defined in the civil nuclear code ETC-C and its UK CD (see Sub-chapter 3.8, section 4) must be carried out for a construction and station operating life time of 75 years for the Common Raft. To allow for a decommissioning period of 15 years plus contingency, the durability of reinforced concrete is considered at the design stage with a 100-year lifespan adopted in the definition of the structural class in accordance with Eurocode 2 (i.e. structural class S5) Design Process The general design process is described in Sub-chapter 22.3 and further detail on civil specific aspects of the design process are given in Sub-chapter 3.3, section Design Basis Documents For Safety Class 1 structures, such as the Common Raft the design basis is provided mainly by the civil nuclear code ETC-C and the UK CD as described in Sub-chapter 3.8, section 4, and complemented by the generic design basis documentation presented in Sub-chapter 3.3, section 1. In addition, some specific documentation also forms the design basis for Common Raft. Note that information in a more detailed document, such as the Structural Design Method Statement (SDMS), supersedes information in more high level documents, such as the hypothesis notes. Hypothesis notes provide the Civil Works Designer(s) (CWD) with the initial input data and design criteria to enable the production of a SDMS and launch the global studies. The SDMSs are the CWD interpretation of the input data from the Responsible Designer written as methods and criteria to be used in the design. The SDMSs also include any structural analysis methodologies to be adopted during the detailed design of the Common Raft. The specific documents are, in order of increasing level of detail: General Hypothesis Note for Civil Engineering design of Nuclear Island Building [Ref. 2], Common Raft Hypotheses Note [Ref. 3], and Common Raft SDMS [Ref. 4].

13 Page No.: 6 / 207 Safety requirements for the HPC EPR are implemented in the civil engineering design through the application of these documents and the Engineering Requirements (ERs) in the SFR schedule referenced in section above. The design of the Common Raft must be assessed in accordance with the design principles and guidance provided in ETC-F, the EPR technical code for Fire Protection. This code is presented in Sub-chapter 3.8, section Design Actions and Combinations The safety and design requirements described in section 1.0 are translated in the civil nuclear code ETC-C and UK CD on the basis of Eurocode design principles in order to arrive at the different design load cases (actions), combinations and structural design criteria. The civil nuclear code ETC-C and UK CD differentiates actions into permanent, variable and accidental (see Sub-chapter 3.3, section 1). Combinations of actions with their corresponding coefficients are set out in the civil nuclear code ETC-C paragraph for different situations (construction, normal operation and accidental). Depending on the situation under consideration, the following are checked: o o o Static Equilibrium Limit State (EQU), Ultimate Limit State (ULS), and Serviceability Limit State (SLS). The design basis documents described in section above define the load cases that are to be considered in the design of the Common Raft: Construction, normal operation, inspection earthquake, level of the groundwater table, test (containment), Loss Of Coolant Accident (LOCA) 2A (applicable to the HR structures only), Severe Accidents (SA) (applicable to the HR structures only), High Energy Line Break (HELB) (diameter of pipes less than { SCI removed } which contain fluid at temperatures above { SCI removed }) or Rupture of High Energy Pipe (RHEP), locally (only for the parts in the affected rooms), ejection of internal projectile or dropped load (locally), Design Basis Earthquake (DBE) including induced vibrations, aircraft crash including induced vibrations (vibrations and intersections with APC shell), flooding, and

14 Page No.: 7 / 207 Steam Line Break LOCA (SLB-LOCA) plus DBE (applicable to the HR structures only). The following ETC-C load combinations are disregarded in the Common Raft Hypothesis Note: normal operation + climatic conditions, test (polar crane), test (other cranes), external explosion, extreme climatic conditions (wind, snow), exceptional temperature (water), exceptional situation temperatures (air), accidental situation temperatures (water), and DBE + Extreme Climatic Conditions (applicable to steelwork structures exposed to climatic conditions). The Common Raft SFR schedule referenced in section above lists and justifies all load combinations to be considered in the design, including those not listed in civil nuclear code ETC-C and UK CD Structural Analysis and Design A description of the structural analysis and design methods can be found in Sub-chapter 3.3, section 1. The Common Raft SDMS [Ref. 4] provides a description of the specific methods used in the analysis and design of the Common Raft DESCRIPTION OF THE STRUCTURE, LAYOUT AND INTERFACES All levels quoted in this section are given with respect to the Building Datum, which equals m Above Ordnance Datum (AOD) General Description of the Structure The Common Raft is a reinforced concrete structure which supports the HR, the HL, the HK Buildings and the APC Shell, there is one per Unit (2 per site) (see plot plan, Sub-chapter 2.2). Loads are transferred from the slabs to the walls at each level through the aforementioned buildings all the way down to the foundation level and into the Common Raft. The prestressing gallery is situated underneath the common raft; there is no structural connection between the two structures. A compressible joint is present at the interface between the prestressing gallery walls and the common raft. The raft takes the shape of a cross within a square measuring approximately { SCI removed } as shown in Section Figure 1; its surface area is approximately { SCI removed }.

15 Page No.: 8 / 207 The thickness of the Common Raft varies: from { SCI removed } to { SCI removed } under the HK and HL Buildings { SCI removed } under the majority of the HR Building, from { SCI removed } to { SCI removed } under the gusset area (junction between the Common Raft, Internal and Outer Containments). There are three distinct zones of the Common Raft: { SCI removed } under the HK Building and controlled areas of the HL Buildings, { SCI removed } or { SCI removed } under the non-controlled areas of the HL Buildings (and periphery of the HK Building), { SCI removed } under the HR Building. Protection against airplane crash is provided by the APC Shell, a reinforced concrete structure that is located over the buildings which it houses, and which can be considered as a separate structure connected monolithically to the Common Raft. The Common Raft supports controlled areas and rooms that contain radioactive materials (hot zones). There are pits in the Common Raft in all the buildings for housing mechanical and electrical systems, in particular in the HK Building; these pits are mainly located along the Outer Containment. They are designed to provide the ultimate protection against spillage and as such must be leak tight. This is achieved by: Applying specific protective coatings in areas where there is potential for spillage of radioactive liquids, and a double layered stainless steel liner in sumps which collect spillages of radioactive liquids (similar to the double layer pipes used for the RPE [NVDS] system). Limiting crack widths in accordance with the Common Raft structural class and Eurocode 2. Additional information on the verification of crack widths for concrete structures can be found in the Verification of crack width for EPR concrete structures for durability requirement [Ref. 5] Layout This section is not applicable to the Common Raft Interfaces with Surrounding Structures The HR [RB] Internal Structures have their own support slab that rests on the Common Raft at { SCI removed }. The periphery of the HR Internal Structures slab is shaped by the gusset of the Inner Containment Wall. There is no connection or anchor between the HR Internal Structures and the NI Common Raft and/or gusset. Moreover, the HR Internal Structures are separated from the Common Raft and the gusset by the Internal Containment steel liner.

16 Page No.: 9 / 207 The APC Shell is connected to the Common Raft and the Outer Containment wall but has no connection to the other structures. Although different structures, they are connected monolithically. The lowest walls of the HK and HL Buildings are structurally connected to the Common Raft. The prestressing gallery has no structural connection to the Common Raft Internal Civil Engineering Interfaces There are no Common Raft specific interfaces between the main structure and any secondary (other) structures. General interfaces between main and other structures connected to the Common Raft, including a description of how other structures are considered within the civil engineering design, are described in Sub-chapter 3.3, section 3. Requirements for safety related interfaces between civil engineering and mechanical equipment are defined in Sub-chapter 3.5. The way these interfaces are handled in the engineering sequence is described in the NI civil engineering design process note [Ref. 6] and in Sub-chapter Facilities must also be protected from natural (lightning strikes) and industrial electro-magnetic interference. An earthing system composed of earthing grids, Faraday cages, interconnection with structural reinforcement, lightning arrestors and lightning rods is designed such that the residual risk of overvoltage in each EPR room is consistent with the equipment s overvoltage immunity DESIGN SUBSTANTIATION Design Substantiation Process A description of the Design Substantiation Process can be found in Sub-chapter 3.3, section 1. Design substantiation for civil structures is provided through a number of deliverables produced at different stages throughout the engineering sequence. These deliverables will be completed before the lifting of the relevant concrete hold point. A list of these documents can be found in Sub-chapter 3.3, section Output of the Design An explanation of design outputs and how they are used to demonstrate that safety requirements have been met can be found in Sub-chapter 3.3, section CONSTRUCTION The purpose of this section is to examine if there are any construction sequencing or constructability issues which need to be addressed for the design of the Common Raft, which could in turn impact upon the safety justification to commence construction.

17 Page No.: 10 / 207 The Civil Work Contractor (CWC) will provide a detailed document (Civil Works Method Statement [Ref. 7]) concerning their construction methods. The relevant sections of this document will be considered in order to ensure consistency between those methods and the safety requirements. SECTION FIGURE 1 : { SCI REMOVED } { This figure contains SCI and has been removed }

18 Page No.: 11 / PRE-STRESSING GALLERY This section outlines the safety justification for the Hinkley Point C (HPC) Pre-Stressing Gallery (PST). The overview of the safety justification for civil structures and links to the overall safety case are presented in Sub-chapter The HPC site geotechnical and earthworks aspects are described in Sub-chapter The content of the section follows the outline structure described in Sub-chapter 12.1 section 2. The PST Gallery structure safety justification is supplemented by Sub-chapter 3.3, which contains generic justifications applicable to all structures. Furthermore, the PST Gallery structure safety justification is supplemented by the safety justifications for the common raft, Reactor Building (HR [RB]) and HR [RB] Building Inner Containment, with the relevant information contained in the sections below: Sub-chapter 12.3 section 1, for the Foundations (including the Common Raft). Sub-chapter 12.3 section 3, for the Reactor Building (HR [RB]). Sub-chapter 12.3 section 4, for the Inner Containment Wall with Steel Liner SAFETY REQUIREMENTS The PST Gallery structure safety requirements relate to nuclear safety and that shall be fulfilled in order to meet the HPC Main Safety Functions (MSFs) as defined in Sub-chapter 3.2 section Main Safety Functions The PST Gallery does not itself perform any of the three MSFs described in Sub-chapter 3.2 section 3. The PST Gallery structure does however house and protect the vertical and gamma pre-stressing cable bottom anchorages (which are critical for the Inner Containment function and therefore the MSF of confinement of radioactivity ). The Safety Functional Requirements (SFRs) discussed in section below are met in order ensure fulfilment of the aforementioned MSFs Safety Classification of the Structure The PST Gallery structure is a sub-structure of the HR Building, which is a Safety Class 1 civil structure. The PST Gallery structure also houses the vertical and gamma pre-stressing cable bottom anchorages, therefore the PST Gallery structure is a Safety Class 1 structure. The justification for the classification is given in Sub-chapter 3.2 section 10 using the classification approach defined in Sub-chapter 3.2 section 8. The PST Gallery structure must therefore meet C1 requirements and is subject to SC1 seismic requirements (see Sub-chapter 3.2 section 8).

19 Page No.: 12 / Safety Functional Requirements The SFRs applicable to the PST Gallery structure are determined by taking load cases originating from the faults and hazards defined in Chapters 13, 14 and 15, and associating these with the operating situations whilst considering the required functions of the structure. SFRs are applied in order to fulfil the MSFs defined in section The SFRs relate to the required behaviour of the structures, which will be either reversible or irreversible after application of sustained, variable or accidental loads. The structural performance and behavioural criteria are chosen with respect to operating and hazard conditions (normal, exceptional and accidental) (see Sub-chapter 3.3 for further detail). The general operating and hazard conditions are defined in the civil nuclear code ETC-C and the UK Companion Document (UK CD), see Sub-chapter 3.8 section 4. The expected function of the PST Gallery structure, after application of the loads, are summarised in an SFR schedule [Ref. 8]. This SFR schedule is developed as a set of detailed safety requirements on individual Structures, Systems and Components (SSCs) which comprise the PST Gallery. The SFRs to be applied are dependent on how the respective SSCs contribute to plant and structural safety, their functional role and the physical arrangement of the SSC. Each SFR is relative to a hazard or plant condition. A set of engineering requirements are listed for each SFR to ensure that the SFR is met through applied civil engineering for the hazard or condition in question Beyond Design Basis (BDB) Considerations In addition to the design conditions identified, it must be demonstrated that adequate margins exist to cover hypothetical scenarios beyond the design basis for the PST Gallery structure so that cliff edge failures do not occur. The Level 3 Design Substantiation Report (DSR) [Ref. 4] will demonstrate deterministically that adequate margins exist and are incorporated into the design of the PST Gallery structure. The following aspects must be considered: description of alternative load paths, ductile design, evaluation of margins in loadings and factors, first mode of failure. In addition to the DSR, a probabilistic study will be undertaken that aims to confirm a BDB capacity exists for the PST Gallery structure Operation; Examination, Maintenance, Inspection and Testing (EMIT); and Decommissioning Requirements Sub-chapter 18.1 describes how Human Factors (HF) principles, methods and standards are integrated into the HPC EPR design. The civil engineering design contributes significantly to achieving the HF principles stipulated in Sub-chapter 18.1 section 5.

20 Page No.: 13 / Operation Requirements A description of how operational requirements are considered within the civil engineering design can be found in Sub-chapter 3.3 section EMIT Requirements A description of how EMIT is considered within the civil engineering design can be found in Sub chapter 3.3 section Decommissioning Requirements A description of how decommissioning requirements are considered within the civil engineering design can be found in Sub-chapter 3.3 section ROLE OF THE STRUCTURE Main Roles of the Structure The main role of the PST Gallery structure is to provide an area from which vertical and gamma pre-stressing cable tendons can be installed, stressed and grouted (or greased) during construction. The PST Gallery structure houses and protects the vertical and gamma pre-stressing cable bottom anchorages from external hazards throughout the design life of the plant. It also permits access for EMIT to part of the Containment Instrumentation System (EAU) Main Components and Systems Housed in the Structure The following is a non-exhaustive list of the main components and systems housed in the PST Gallery structure: part of the EAU system, prestressing tendon heads DESIGN BASIS The general design basis for structures including general performance and behavioural criteria are described in Sub-chapter 3.3 section 1. The PST Gallery structure specific performance criteria and design basis information, which ensure that the specific SFRs (as described in section 2.0.3) and design requirements are met, are defined below Design Life The civil engineering design must ensure that the structures perform their safety function throughout the planned life of the plant. UK EPR structures are generally designed with an operating life time of 60 years. Additional periods of 5five years for construction and 15 years for decommissioning must also be taken into account.

21 Page No.: 14 / 207 The detailed design of the reinforced concrete and steel structures against the load cases defined in the civils nuclear code AFCEN ETC-C 2010 and UK companion document CD (see section 4 of the Sub-chapter 3.8 section 4) must shall be carried out for a construction and station operating life time of at least 65 years for the PST Gallery structure., To allow for a decommissioning period of 15 years plus contingency, the durability of reinforced concrete is considered at the design stage with a 100 year lifespan adopted in the definition of the structural class in accordance with Eurocode 2 (i.e. Structural Class S5) Design Process The general design process is described in Sub-chapter 22.3 and further detail on civil specific aspects of the design process are given in Sub-chapter 3.3 section Design Basis Documents For Safety Class 1 structures such as the PST Gallery structure, the design basis is provided mainly by the civil nuclear code ETC-C and UK CD as described in Sub-chapter 3.8 section 4 and complemented by the generic design basis documentation presented in Sub-chapter 3.3. In addition, some specific documentation also forms the design basis for the PST Gallery structure. Note that information in a more detailed document, such as the Structural Design Method Statement (SDMS), supersedes information in more high level documents, such as the hypothesis notes. Hypothesis notes provide the Civil Works Designer(s) (CWD) with the initial input data and design criteria to enable the production of a SDMS and launch the global studies. The SDMSs are the CWD interpretation of the input data from the Responsible Designer written as methods and criteria to be used in the design. The SDMSs also include any structural analysis methodologies to be adopted during the detailed design of the PST Gallery structure. The specific documents are, in order of increasing level of detail: General Hypothesis note for Civil Engineering design of Nuclear Island Buildings [Ref. 2]; Pre-stressing Gallery hypotheses note [Ref. 9]; Common raft and Pre-stressing Gallery SDMS [Ref. 4]. Safety requirements for the HPC EPR are implemented in the civil engineering design through the application of these documents and the Engineering Requirements in the SFR schedule referenced in section above. The design of the PST Gallery structure must be assessed in accordance with the design principles and guidance provided in ETC-F, the EPR technical code for Fire Protection. This code is presented in Sub-chapter 3.8 section Design Actions and Combinations The safety requirements described in section 2.0 are translated in the civil nuclear code ETC-C on the basis of Eurocode design principles to arrive at the different design load cases (actions), combinations and structural design criteria. The civil nuclear code ETC-C and UK CD differentiates actions into permanent, variable and accidental (see Sub-chapter 3.3).

22 Page No.: 15 / 207 Combinations of actions with their corresponding coefficients are set out in the civil nuclear code ETC-C and UK CD Table bis (see Sub-chapter 3.8 section 4) for different situations (construction, normal operation and accidental). Depending on the situation under consideration, the following are checked: Static Equilibrium Limit State (EQU); Ultimate Limit State (ULS); Serviceability Limit State (SLS). The design basis documents described in section above define the load cases that are to be considered in the design of the PST Gallery structure: construction, normal operation, Inspection Earthquake, level of ground water table, containment test, Severe Accidents, Design Basis Earthquake (DBE) including induced vibrations, and flooding. The following civil nuclear code ETC-C load combinations are disregarded in the PST Gallery structure hypotheses note: Normal operation with bounding climatic conditions (PST Gallery structure is a buried structure); Double ended break in the reactor coolant system loop pipe (referred to as a 2A -Loss of Coolant Accident (LOCA) (Pre-stressing Gallery is not directly subjected to this load case); High-Energy Line Break (HELB) (PST Gallery structure contains no high energy pipe work); Ejection of internal projectile or dropped load (no risk of internal missiles, risk and consequences of a dropped load negligible); External explosion (Pre-stressing Gallery is not directly subjected to this load case); Climatic accident (wind, snow) (PST Gallery structure is a buried structure); Steam Line Break LOCA (SLB-LOCA) and DBE (applicable to the HR Building and the HR Building containment structures);

23 Page No.: 16 / 207 Exceptional temperature (water / air) (no pools and tanks in the PST Gallery structure, the PST Gallery structure is a buried structure); Accidental temperature (water) (no pools or tanks present in the PST Gallery structure). The PST Gallery structure SFR schedule referenced in section above lists and justifies all load combinations to be considered in the design, including those not listed in the civil nuclear code ETC-C Structural Analysis and Design A description of the structural analysis and design methods can be found in Sub-chapter 3.3 section 1. The PST Gallery structure SDMS [Ref. 4] provides a description of the specific methods used in the analysis and design of this structure DESCRIPTION OF THE STRUCTURE, LAYOUT AND INTERFACES All levels quoted in this section are given with respect to the Building Datum, which equals m Above Ordnance Datum (AOD) General Description of the Structure The PST Gallery structure sits under the common raft which supports the main Nuclear Island buildings so there are two on the HPC site, one per unit (see Sub-chapter 2.2 for the location of the PST Gallery structure on the plot plan). The PST Gallery structure is used to pre-stress the vertical tendons of the HR Building inner containment. Once the tendons are pre-stressed and grouted, the PST Gallery structure is used during the lifetime of the power plant for the inspection of the tendon heads and monitoring and testing of instrumented non-grouted tendons. No equipment is located in the PST Gallery structure other than an access ladder leading to the Access Building (HW), cable trays (for normal and emergency lighting, electrical power sockets and a fire detection system), electrical cabinets, EAU system levelling bolts and cover guards (fixed on the roof slabs) and sumps covered by grating for water collection. The reinforced concrete PST Gallery structure can be separated into three main elements that have different structural requirements. These are the U-section gallery, the roof slabs over the gallery and the external access shaft. The U-shaped gallery section and the two antennae form a monolithic structure. These elements are illustrated in Section Figure 1 and Section Figure 2. The junction between the common foundation raft and the PST Gallery is located at a depth of { SCI removed }. The top of the gallery raft slab is at { SCI removed }. Pre-cast roof slabs with tendon cable anchorages embedded within them cover the entirety of the circular part of the PST Gallery structure. The slabs are used as formwork for the concreting of the common raft above. The slabs have a thickness of { SCI removed }. There are { SCI removed } tendon anchorages in total embedded in the slabs. The pre-cast roof slabs are structurally decoupled from the U-section gallery by a horizontal and vertical spacing ({ SCI removed }). A sealant will be included between the pre-cast roof slabs and the U-section gallery. Once the pre-stressing is carried out the slabs are effectively joined to the common raft and held in position by the containment tendons and reinforcement.

24 Page No.: 17 / 207 { This paragraph contains SCI only text and has been removed }. The circumference of the circular part of the PST Gallery structure is approximately { SCI removed }. The antennae (as shown in Section Figure 1) have a length of approximately { SCI removed } each. The gallery is founded at { SCI removed } below the platform level (at m AOD). The thickness of the walls and slabs ranges from { SCI removed }. The structural behaviour of the inner ring wall of the PST Gallery structure is similar to a circular reservoir, with lateral loads from the ground and water table placing it in tension, thus requiring high levels of reinforcement. The ground inside the circle formed by the inner gallery wall is confined so loads that arise due to concrete shrinkage can be significant. The outer wall in contrast to the inner walls benefits from a vaulting effect. The external access shaft sits on one of the radial galleries and transmits vertical loads to it. Unlike the main U-shaped gallery it is not subject to loads originating from the common raft and can be designed independently. Unlike for the PST Gallery structure under the common raft, the backfill around the external access shaft is soil backfill. A removable slab covers the external access shaft at ground level. A horizontal watertight movement joint separates the vertical external access shaft from the antennae (and consequently from the rest of the PST Gallery structure. The reinforcement between the vertical external access shaft and the antennae is not connected. Further detail on joints is given in Sub-chapter 3.3 section 3. The short access shaft linking the PST Gallery and the HW Building is part of the HW Building. As it is placed under the common raft which is a Safety Class 1 structure, the backfill between the PST Gallery structure walls and the unexcavated ground beneath the common raft is mass concrete backfill Layout The layout of the PST Gallery structure is shown in the plot plan in Sub-chapter 2.2. Due to the twin reactor design and the different position of the HW Building on both units, the PST Gallery structure accesses are inversed in Unit 1 with respect to Unit 2. The design has however been optimised so that under the footprint of the common raft the PST Gallery arrangement is nearly identical for both units. The internal dimensions of the PST Gallery are as follows: { SCI removed } from the bottom of the roof slab to the top of the base slab, { SCI removed } wide Interfaces with Surrounding Structures The PST Gallery structure interfaces with the common foundation raft which transmits loads directly to the gallery walls and roof slabs as well as indirectly though the ground / mass concrete surrounding the PST Gallery structure. The reinforcement between the PST Gallery structure and the common foundation raft is not connected.

25 Page No.: 18 / 207 The main structure of the gallery (U-shaped section and antennae see Section Figure 1) is decoupled from the common raft to prevent the concentration of stresses at the gallery-raft interface due to potential horizontal differential displacements from dynamic, thermal or long term static loading. A horizontal hard (mortar) joint between the top of the gallery walls and the common raft is nonetheless required as the water table hydrostatic pressures are high enough in the vicinity of the PST Gallery structure to create buoyancy and push the gallery against the raft. The joint detail also ensures leak-tightness at this interface through a waterstop. As discussed above, the pre-cast roof slabs have the vertical and gamma tendon anchorages of the HR Building Inner Containment embedded within them. The PST Gallery structure has an interface with the HW Building through an access shaft at the end of the radial gallery antennae. A horizontal joint separates the shaft from the PST Gallery structure. The HW Building does not transmit vertical loads to the PST Gallery structure however indirect loads are transmitted through the ground / mass concrete surrounding the PST Gallery structure Internal Civil Engineering Interfaces There are no PST Gallery structure specific interfaces between the main structure and any secondary (other) structures. General interfaces between main and other structures in the PST Gallery structure, including a description of how other structures are considered within the civil engineering design, are described in Sub-chapter 3.3 section 3. All the safety related interfaces between civil engineering and mechanical equipment are designed according to the requirements outlined in Sub-chapter 3.5. The way these interfaces are handled in the engineering sequence is described in the NI civil engineering design process note [Ref. 6] and in Sub-chapter The PST Gallery structure will be protected from natural (lightning strikes) and industrial electromagnetic interference. An earthing system composed of earthing grids and Faraday cages, (made by means of interconnecting structural reinforcement, lightning arrestors and lightning rods) is designed such that the residual risk of overvoltage in each EPR room is consistent with the equipment overvoltage immunity DESIGN SUBSTANTIATION Design Substantiation Process A description of the Design Substantiation Process can be found in Sub-chapter 3.3 section 1. Design substantiation for civil structures is provided through a number of deliverables produced at different stages throughout the engineering sequence. These deliverables will be completed before the lifting of the relevant concrete hold point. A list of these documents can be found in Sub-chapter 3.3 section 1.

26 Page No.: 19 / Output of the Design An explanation of design outputs and how they are used to demonstrate that safety requirements have been met can be found in Sub-chapter 3.3 section CONSTRUCTION The purpose of this section is to examine if there are any construction sequencing or constructability issues which need to be addressed for the design of the PST Gallery, which could in turn impact upon the safety justification to commence construction. The Civil Work Contractor (CWC) has provided detailed documents (Civil Work Method Statements [Ref. 7] and [Ref. 10]) concerning their construction methods. The relevant sections of this document will be considered in order to ensure consistency between those methods and the safety requirements. SECTION FIGURE 1 : { SCI REMOVED } { This figure contains SCI and has been removed }

27 Page No.: 20 / REACTOR BUILDING (HR [RB]) This section outlines the Hinkley Point C (HPC) safety justification for the Reactor Building (HR [RB]) structure. The overview of the safety justification for civil structures and links to the overall safety case are presented in Sub-chapter HPC site geotechnical and earthworks aspects are described in Sub-chapter The content of the section follows the outline described in Sub-chapter 12.1 section 2. The safety justification for the HR Building is supplemented in Sub-chapter 3.3 which contains generic justifications applicable to all structures. Furthermore, the different sub-structures of the HR Building are covered by other sections of Sub-chapter The HR Building safety justification is thus supplemented by the Sub-chapter sections below: Sub-chapter 12.3 section 1, for the Foundations (Including Common Raft); Sub-chapter 12.3 section 2, for the Pre-stressing Gallery; Sub-chapter 12.3 section 4, for the Inner Containment Wall with Steel Liner; Sub-chapter 12.3 section 5, for the Containment Penetrations; Sub-chapter 12.3 section 6, for the Reactor Building Internal Structures; Sub-chapter 12.3 section 7, for the Air Plane Crash Shell and Outer Containment; Sub-chapter 12.3 section 8, for the Stainless Steel Lined Pools, Tanks and RPE Sumps SAFETY REQUIREMENTS The HR Building safety requirements are requirements that relate to nuclear safety and that shall be fulfilled in order to meet the HPC Main Safety Functions (MSFs) as defined in Subchapter 3.2 section Main Safety Functions (MSFs) The MSF (from Sub-chapter 3.2 section 3) relative to the HR Building is "confinement of radioactive material". The HR Building protects the general public, workers and the environment from normal and accident situations by providing a barrier to the release of radioactivity. The HR Building also provides protection to systems and components housed within it. The HR Building houses systems and components responsible for the main level safety functions of: fuel heat removal, control of fuel reactivity, and confinement of radioactive material, in particular the Reactor Coolant System (RCP) and the Safety Injection System (RIS [SIS]). Alongside the aforementioned systems, the HR Building structure itself provides a barrier to the control of release of radioactivity into the environment. The safety functional requirements (SFRs) referenced in section shall be met in order to ensure fulfilment of the aforementioned MSFs.

28 Page No.: 21 / 207 In addition, the specificities of each sub-structure in terms of MSFs are dealt with in the different sub-chapter sections mentioned in section 3 above Safety Classification of the Structure The HR Building is a Safety Class 1 civil structure as it contributes directly to the main MSF of confinement of radioactive material and houses Safety Class 1 equipment and systems. The justification for the classification is given in Sub-chapter 3.2 Section 10 using the classification approach defined in the Sub-chapter 3.2 section 8. The HR Building must therefore meet C1 requirements and meet SC1 seismic requirements (see Sub-chapter 3.2 section 8) Safety Functional Requirements The SFRs applicable to the HR Building are determined by taking loads cases originating from the faults and hazards defined in Chapter 13, Chapter 14 and Chapter 15, and associating these with the operating situations whilst considering the required functions of the structure. SFRs are applied in order to fulfil the MSFs defined in section The SFRs relate to the required behaviour of the structures, which will be either reversible or irreversible after application of sustained, variable or accidental loads. The structural performance and behavioural criteria are chosen with respect to operating and hazard conditions (normal, exceptional and accidental) (see Sub-chapter 3.3 section 1 for further detail). The general operating and hazard conditions are defined in the civil nuclear code ETC-C and the UK Companion Document (see Sub-chapter 3.8 section 4). As the HR Building is divided in several sub-structures, the expected functions of the HR Building structure, after application of the loads, are those of its sub-structures and are listed within the associated sections of this Sub-chapter (see Sub-chapter 12.3 section 1.0.3, section 2.0.3, section 4.0.3, section 5.0.3, section 6.0.3, section 7.0.3, and section 8.0.3). The expected functions of the HR Building, after application of the loads, are summarised in the SFR schedules. As the HR Building is divided in several main sub-structures, its safety functional requirements are those of its sub-structures. They are presented within dedicated safety functional requirements schedules [Ref. 1] [Ref. 8] [Ref. 11] [Ref. 12] These SFRs schedules are developed as a set of detailed safety requirements on individual Structures, Systems and Components (SSCs). The SFRs to be applied depend on how the respective SSCs contribute to plant and structural safety, their functional role and the physical arrangement of the SSC. Each SFR is related to associated hazard or plant conditions. A set of engineering requirements are listed for each SFR to ensure that the SFR is met through applied civil engineering for the hazard or condition in question Beyond Design Basis Considerations Beyond design basis considerations relative to HR Building are presented for each of its substructures within the sections mentioned in section 3 above.

29 Page No.: 22 / Operation; Examination, Maintenance, Inspection and Testing (EMIT); and Decommissioning Requirements Operation, Examination, Maintenance Inspection and Testing and Decommissioning requirements relative to the HR Building are presented for each of its sub-structures within the sections mentioned in section 3 above ROLE OF THE STRUCTURE Main Roles of the Building The main roles of the HR Building are the following (as described in the HR Building Technical Specification [Ref. 13]): To provide the third barrier to radiation release from the Reactor Core; To house and protect the Primary Circuit components; To withstand the loading due to pressure, thermal and mechanical effects resulting from the situations described in section 3.0.3; To be capable of undergoing leak-tightness and resistance tests; To allow access and egress of the personnel and equipment to the primary circuit components under normal operating conditions Main Components and Systems Housed by the HR Building The following is a non-exhaustive list of the main components and systems housed in the HR Building: the Reactor Pressure Vessel (RPV), the Steam Generators (SG), the Pressuriser, the Reactor Coolant Pumps (RCP), the Equipment Hatch, the Personnel Airlocks, the Fuel Transfer Tube (FTT), the Polar Crane, the H2 dampers, the H2 recombiners,

30 Page No.: 23 / 207 tooling structure for the Core Melt Stabilization System (RSC [CMSS]), the Main Feed Water System (ARE [MFWS]) for steam generators, the Containment Sweep Ventilation System (EBA [CSVS]), the Annulus Ventilation System (EDE [AVS]), the Containment Leak-off and Seal Monitoring System (EPP), the HR Building Internal Filtration System (EVF), the Containment Cooling Ventilation System (EVR [CCVS]), the Containment Heat Removal System (EVU [CHRS]), the Fuel Pool Cooling (and Purification) System (PTR [FPCS/FPPS]), the RCP system, the Chemical and Volume Control System (RCV [CVCS]), the RIS system, the Nuclear Vent and Drain System (RPE [NVDS]), the Component Cooling Water System (RRI [CCWS]), the Main Steam Supply System (VVP [MSSS]) DESIGN BASIS Design basis relative to the HR Building is presented for each one of its sub-structures within the sections mentioned in section 3 above DESCRIPTION OF THE STRUCTURE, LAYOUT AND INTERFACES All levels quoted in this section are given with respect to Building Datum, which equals m Above Ordnance Datum (AOD) General Description of the Structure There is one HR Building per unit. The HR Building location on the plot plan is discussed in Subchapter 2.2. The HR Building is comprised of a double-walled containment located in the centre of the Common Raft. This raft is shared with the Safeguard Mechanical and Electrical Buildings (HL [SB]) and HK [FB] Building, which are located around the central HR Building.

31 Page No.: 24 / 207 The HR Building comprises (see Section Figure 1): A prestressed concrete Inner Containment wall, the inner surface of which is covered with a steel liner which is anchored to the inner face of the containment wall and simply set on the Common Raft (without anchorage). A Pre-stressing Gallery is located below the raft to provide access to, and permit tensioning of, the vertical and gamma prestressing tendons. The Inner Containment wall comprises electrical and mechanical penetrations, the largest of which is the Equipment Hatch through which heavy-duty RCP system components are brought into the HR Building. The key role of the pre-stressed concrete structure of the Inner Containment is to withstand the potential over pressure which could occur as the result of an accident. The steel liner is designed to help maintain leak-tightness in the event of an accident. Internal Structures, which separate the containment into two zones, (i.e. the two-room design concept) helping to provide radiological protection for personnel when in the upper section (outer zone) of the HR Building. The inner zone contains the reactor vessel, the primary coolant system and the In-containment Refuelling Water Storage Tank (IRWST). The IRWST contains { SCI removed } of water (dependent upon operating conditions) and is lined with a stainless steel leak-tight membrane (liner). The empty volume within the Inner Containment zone adjacent to the IRWST is designed to facilitate spreading and cool-down of corium melt that could potentially be released from the reactor vessel in the event of a Severe Accident (SA) (see Sub-chapter 6.1 for a description of the spreading area and the reactor pit). An Outer Containment wall, designed both to protect the Inner Containment from specified externally generated hazards and to contain gas leakages from the Inner Containment by means of an inter-containment EDE [AVS] system. An inter-containment annulus space located between the Inner and Outer Containments. This inter-containment space is occupied by electrical and mechanical installation linking the HR Building and the surrounding buildings, with steelwork platforms. It houses in particular the EDE [AVS] system mentioned above and the anchors of the horizontal prestressing tendons, embedded in the three ribs. A negative pressure is maintained in the inter-containment annulus thanks to the EDE system, to ensure the collection of leaks from inside the containment. Two compartments along the FTT path, one inside the HR [RB] Internal Structures and the other in the inter-containment annulus. Both of these compartments are water-tight because they must ensure there is no drainage path for the water in the HR Building and HK Building pools in the case of failure of the FTT. This water-tightness requirement also applies to the flexible seals used to plug the gaps between the different structures and the doors used to access these compartments in limit of these two rooms. Six double-walled pipes which provide a connection between the IRWST and the inlet of the RIS and EVU [CHRS] system pumps located in the HL Buildings. This pipework is embedded into the containment concrete over a length of more than ten meters. The following in-containment water tanks and pools (in addition to the IRWST) which are used during fuel discharge and reload, are the: o o o reactor pool, storage pool for reactor internals, instrumentation lances storage,

32 Page No.: 25 / 207 o transfer pool (connected to the spent fuel pool via the FTT). Except for the inter-containment annulus space, all main sub-structures of the HR Building are described in more details within their associated sections (see section 3 above) Layout The layout of the HR Building is presented for each of its sub-structures within the sections mentioned in section 3 above Interfaces with Surrounding Structures The HR Building is surrounded by the HK Building and HL Buildings. These buildings all have the same foundation: the Common Raft. The Outer Containment is the outermost structure of the HR Building, the connection of the surrounding buildings with the HR Building is described in detail in Sub-chapter 12.3 section 7 dealing with the Outer Containment. Other than the Outer Containment the HR Building interfaces the Common Raft only through a monolithic structural connection in the gusset area. The HR Internal Structures are structurally separated from the Common Raft and HR Inner Containment by the steel liner however sliding of the HR Internal Structures is precluded by socketing of the HR Internal Structures at the gusset level thus horizontal forces will be transmitted between the structures. The interfaces between the HR, HK and HL Buildings are identified in Section Figure 2. Mechanical and electrical connections between the HR Building and peripheral buildings are routed through Containment Penetrations, as well as access to the HR Building (see the list of penetrations [Ref. 14]). Access to the inside of the HR Building can be gained: from the HK Building thanks to: o o the Personnel Airlock located at { SCI removed } on the north-west part of the HK Building, the Equipment Hatch located at { SCI removed } on the north-east part of the HK Building, from Division 2 (HLB/HLG) of the HL Buildings thanks to the Personnel Airlock located at { SCI removed } on the south-east part of Division 2 of the HL Buildings. The inter-containment annulus can be reached thanks to doors located inside the Outer Containment wall in the HK Building and in the HL Buildings (see the list of penetrations [Ref. 14]) Internal Civil Engineering Interfaces The internal civil engineering interfaces for the HR Building are described for each of its substructures within the sections mentioned in section 3 above DESIGN SUBSTANTIATION Design substantiation relative to the HR Building is presented for each of its sub-structures within the sections mentioned in section 3 above.

33 Page No.: 26 / CONSTRUCTION Construction relative to the HR Building is presented for each of its sub-structures within the sections mentioned in section 3 above. SECTION FIGURE 1 : { SCI REMOVED } { This figure contains SCI and has been removed } SECTION FIGURE 2 : { SCI REMOVED } { This figure contains SCI and has been removed }

34 Page No.: 27 / INNER CONTAINMENT WALL WITH STEEL LINER This section outlines the Hinkley Point C (HPC) safety justification for the Reactor Building (HR [RB]) Inner Containment wall with steel liner (HR [RB] Inner Containment) structure. The overview of the safety justification for civil structures and links to the overall safety case are presented in Sub-chapter HPC site geotechnical and earthworks aspects are described in Sub-chapter The content of the section follows the outline described in Sub-chapter 12.1 section 2. The safety justification for the HR Inner Containment is supplemented in Sub-chapter 3.3 which contains generic justifications applicable to all structures and in Sub-chapter 12.3 section 3 which contains the global HR Building safety justification. Furthermore, the HR Building structures that link with HR Inner Containment (Common Raft, Air Plane Crash (APC) shell, Containment Penetrations and Prestressing Gallery) are covered by other sections of Subchapter 12.3, and the safety justification is thus supplemented by the sub-chapter sections below: Sub-chapter 12.3 section 1, for Foundations (Including Common Raft); Sub-chapter 12.3 section 2, for Prestressing Gallery; Sub-chapter 12.3 section 5, for Containment Penetrations; Sub-chapter 12.3 section 6, for the HR Building Internal Structures; Sub-chapter 12.3 section 7, for Air Plane Crash Shell and Outer Containment SAFETY REQUIREMENTS The Inner Containment wall with steel liner safety requirements are requirements that relate to nuclear safety and that shall be fulfilled in order to meet the HPC Main Safety Functions (MSFs) as defined in Sub-chapter 3.2 section Main Safety Functions (MSFs) The MSF (from Sub-chapter 3.2, section 3) relative to the HR Building is "confinement of radioactive material". The HR Building and more particularly the HR Inner Containment protects the general public, workers and the environment in all normal and accident situations. The HR Inner Containment constitutes the third containment barrier and thus directly contributes to the confinement MSF. The HR Building also provides protection to systems and components housed within it. The HR Building houses multiple systems and components responsible for the main level safety functions of: fuel heat removal, control of fuel reactivity, and confinement of radioactive material, in particular the Reactor Coolant System (RCP [RCS]) and the Safety Injection System (RIS [SIS]). Alongside the aforementioned systems, the HR Building structure itself provides a barrier to the control of release of radioactivity into the environment.

35 Page No.: 28 / 207 The HR Inner Containment provides a physical, resistant and leak tight barrier that ensures, in combination with associated circuits, the confinement of radioactive material that could be released in all normal, exceptional and accidental situations considered in the HPC EPR design (i.e. Plant Condition Categories (PCC) taken from the reference plant design, additional Design Basis Initiating Faults (DBIFs) considered as part of the HPC Fault and Protection schedule, Design Extension Conditions (DEC) including Severe Accidents, and other hazards). The Safety Functional Requirements (SFRs) referenced in section shall be met in order to ensure fulfilment of the aforementioned MSFs Safety Classification of the Structure The HR Inner Containment is a sub-structure of the HR Building. Both the HR Building and the HR Inner Containment are Safety Class 1 civil structures as they contribute directly to the main MSF of confinement of radioactive material and they house Safety Class 1 equipment and systems. The justification for the classification is given in Sub-chapter 3.2 section 10 using the classification approach defined in the Sub-chapter 3.2 section 8. The Inner Containment must therefore meet C1 requirements and SC1 seismic requirements (see Sub-chapter 3.2 section 8) Safety Functional Requirements The SFRs applicable to the HR Inner Containment are determined by taking loads cases originating from the faults and hazards defined in Chapter 13, Chapter 14 and Chapter 15 and associating these with the operating situations whilst considering the required functions of the structure. SFRs are applied in order to fulfil the main safety functions defined in section The SFRs relate to the required behaviour of the structures, which will be either reversible or irreversible after the application of sustained, variable or accidental loads. The structural performance and behavioural criteria are chosen with respect to operating and hazard conditions (normal, exceptional and accidental) (see Sub-chapter 3.3 section 1 for further detail). The general operating and hazard conditions are defined in the civil nuclear code ETC-C and UK Companion Document, (see Sub-chapter 3.8 section 4). The expected functions of the HR Inner Containment, after application of the loads, are summarised in a SFR schedule [Ref. 11]. This SFR schedule is developed as a set of detailed safety requirements on individual Structures, Systems and Components (SSC) which comprise the HR Inner Containment. The SFRs to be applied depend on how the respective SSCs contribute to plant and structural safety, their functional role and the physical arrangement of the SSC. Each SFR is related to associated hazard or plant conditions. A set of engineering requirements are listed for each SFR to ensure that the SFR is met through applied civil engineering for the hazard or condition in question Beyond Design Basis Considerations In addition to the design conditions identified, it must be demonstrated that adequate margins exist to cover hypothetical scenarios beyond the design basis for the HR Inner Containment structure, so that cliff edge failures do not occur.

36 Page No.: 29 / 207 The Level 3 Design Substantiation Report (DSR) [Ref. 4] will demonstrate deterministically that adequate margins exist and are incorporated into the design of the HR Inner Containment. For Nuclear Island buildings the following aspects shall be considered: description of alternative load paths, ductile design, evaluation of margins in loadings and factors, first mode of failure. In addition to the DSR, a probabilistic study will be undertaken later in the design process that confirms a Beyond Design Basis (BDB) capacity exists for the HR Inner Containment Operation; Examination, Maintenance, Inspection and Testing (EMIT); and Decommissioning Requirements Sub-chapter 18.1 describes how Human Factors (HF) principles, methods and standards are integrated into the HPC EPR design. The civil engineering design contributes significantly to achieving the HF principles stipulated in Sub-chapter 18.1 section Operation Requirements A description of how operational requirements are considered within the civil engineering design can be found in Sub-chapter 3.3 section EMIT Requirements A description of how EMIT is considered within the civil engineering design, can be found in Sub-chapter 3.3 section Decommissioning Requirements A description of how decommissioning requirements are considered within the civil engineering design can be found in Sub-chapter 3.3 section 1. The demolition of the HR Inner Containment will be undertaken after dismantling and demolition of the other buildings founded on the Common Foundation Raft ROLE OF THE STRUCTURE Main Roles of the Structure The main roles of the HR Inner Containment are the following: material. Be able to withstand pressure, thermal and mechanical loads resulting from the normal operation and accidental conditions and to ensure the confinement of radioactive

37 Page No.: 30 / 207 Be capable of undergoing leak tightness and resistance tests (See Sub-chapter 6.1 section 5 for information on pre-operational and periodic tests). Allow access and egress of personnel and equipment to the primary circuit components under normal operating conditions Main Components and Systems Housed by the HR Inner Containment Containment Penetrations are dealt with in Sub-chapter 12.3 section 5. Concerning the components and systems housed in the HR Building, see Sub-chapter 12.3 section 3 relative to the HR Building DESIGN BASIS The general design basis including performance and behavioural criteria are described in Sub-chapter 3.3 section 1. The HR Inner Containment specific performance criteria and design basis information, which ensure that the specific safety functional requirements (as described in section above) and design requirements are met, are defined in the following paragraphs Design Life The civil engineering design must ensure that the structures generally perform their safety function throughout the planned life of the plant. UK EPR structures are designed with an operating life time of 60 years. Additional periods of 5 years for construction and 15 years for decommissioning must also be taken into account. The detailed design of the reinforced concrete and steel structures against the load cases defined in the civil nuclear code ETC-C and its UK Companion Document (see Sub-chapter 3.8 section 4) must be carried out for a construction and station operating life time of 65 years for the HR Inner Containment. To allow for a decommissioning period of 15 years plus contingency, the durability of reinforced concrete is considered at the design stage with a 100-year lifespan adopted in the definition of the structural class in accordance with Eurocode 2 (i.e. structural class S5) Design Process The general design process is described in Sub-chapter 22.3 and further detail on civil specific aspects of the design process are given in Sub-chapter 3.3 section Design Basis Documents For Safety class 1 structures such as the HR Inner Containment, the design basis is provided mainly by the civil nuclear code ETC-C and the UK Companion Document as described in Subchapter 3.8 section 4 and is complemented by the generic design basis documentation presented in Sub-chapter 3.3 section 1.

38 Page No.: 31 / 207 In addition, some specific documentation also forms the design basis for HR Inner Containment. Note that information in a more detailed document, such as the Structural Design Method Statement (SDMS), supersedes information in more high level documents, such as the hypothesis notes. Hypothesis notes provide the Civil Works Designer(s) (CWD) with the initial input data and design criteria to enable the production of a SDMS and launch the global studies. The SDMSs are the CWD interpretation of the input data from the Responsible Designer written as methods and criteria to be used in the design. The SDMSs also include any structural analysis methodologies to be adopted during the detailed design of the HR Inner Containment. The specific documents are, in order of increasing level of detail: General Hypothesis Note for Civil Engineering Design of Nuclear Island Buildings [Ref. 2]. Inner Containment wall with steel liner hypothesis note [Ref. 15]. Inner Containment SDMS [Ref. 4]. Safety requirements for the HPC EPR are implemented in the civil engineering design through the application of these documents and the engineering requirements in the SFR schedule referenced in section above. The design of the HR Inner Containment must be assessed in accordance with the design principles and guidance provided in ETC-F, the EPR technical code for Fire Protection. This code is presented in Sub-chapter 3.8 section Design Actions and Combinations The safety requirements described in section 4.0 are translated in the ETC-C code on the basis of Eurocode design principles in order to arrive at the different design load cases (actions), combinations and structural design criteria. The ETC-C differentiates actions into permanent, variable and accidental (see Sub-chapter 3.3 section 1) Inner Containment Concrete Wall Combinations of actions with their corresponding coefficients are set out in the ETC-C paragraph (see Sub-chapter 3.8 section 4) for different situations (construction, normal operation and accidental). Depending on the situation under consideration, the following are checked: Static Equilibrium Limit State (EQU); Ultimate Limit State (ULS); Serviceability Limit State (SLS). The design basis documents described in section above define the load cases that are to be considered in the design of the HR Inner Containment concrete structure: construction; normal operation; inspection earthquake;

39 Page No.: 32 / 207 test polar crane; containment test; Loss Of Coolant Accident (LOCA) - 2A; Severe Accident (SA); High Energy Line Break (HELB); ejection of internal projectile; Design Basis Earthquake (DBE) including induced vibrations; aircraft crash including induced vibrations; Loss of Coolant Accident Steam Line Break (LOCA-SLB) plus DBE; flooding; exceptional temperature (air). The following ETC-C load combinations are disregarded for the HR Inner Containment concrete structure: normal operation with bounding climatic conditions (shielded by the Outer Containment), level of ground water table (shielded by the Outer Containment and Common foundation Raft); external explosion (shielded by the Outer Containment); climatic accident (wind, snow) (shielded by the Outer Containment); exceptional temperature (water) (concerns Inner Structures due to Pools); accidental temperature (concerns Inner Structures due to Pools). Moreover, section of the ETC-C which is specific to the HR Inner Containment concrete wall gathers some of the above load combinations under three groups to which specific design criteria are associated: Group 1: Construction, Normal operation, Containment test and Inspection Earthquake; Group 2: LOCA-2A, HELB, SA with containment pressure at { SCI removed } and DBE; Group 3: SA with containment pressure at { SCI removed } and LOCA-SLB plus DBE. Specific hazards are also considered in the design of the HR Inner Containment, which are not listed within the ETC-C (see Sub-chapter 13.1 and Sub-chapter 13.2): fire; pipework leaks and breaks;

40 Page No.: 33 / 207 failure of tanks, pumps and valves; internal explosions. The HR Inner Containment SFR schedule referenced in section above lists and justifies all load combinations to be considered in the design, including those not listed in ETC-C Steel Liner The load combinations considered for the design of the steel liner of the HR Inner Containment are given in section of ETC-C. These load combinations are divided into four groups (applying the same principles as for the HR Inner Containment concrete wall): Group 1: Construction, Normal operation, Containment Test and Inspection earthquake; Group 2: LOCA-2A, HELB, SA at { SCI removed }, Aircraft crash including induced vibrations and DBE; Group 3: SA at { SCI removed } and LOCA-SLB plus DBE; Group 3 bis: SA with hydrogen deflagration. The HR Inner Containment SFR schedule referenced in section above also covers the steel liner and lists and justifies all load combinations to be considered in the design of the steel liner, including those not listed in ETC-C Structural Analysis and Design A description of the general structural analysis and design methods can be found in Sub-chapter 3.3 section 1. The HR Inner Containment SDMS [Ref. 4] provides a description of the specific methods used in the analysis and design of this structure. The software used to proceed to the design of the HR Inner Containment is presented in Appendix 3 of Chapter DESCRIPTION OF THE STRUCTURE, LAYOUT AND INTERFACES All levels quoted in this section are given with respect to Building Datum, which equals m Above Ordnance Datum (AOD) General Description of the Structure The HR Inner Containment is a sub-structure of the HR Building (there is one per unit). Its position on the plot plan is discussed in Sub-chapter Geometry The prestressed reinforced concrete HR Inner Containment is comprised, from bottom to top of (see Section Figure 1): cylindrical gusset (from level { SCI removed } to { SCI removed });

41 Page No.: 34 / 207 cylindrical section (from { SCI removed } to { SCI removed }); truncated section (from level { SCI removed } to { SCI removed }); cylindrical section (from { SCI removed } to { SCI removed }), with the following characteristics: o internal diameter: { SCI removed }; o typical thickness: { SCI removed }; o height: { SCI removed }. torispherical dome connected to the top of the inner containment wall (typical thickness, { SCI removed }). The concrete intersection between the cylindrical wall and dome is called the dome ring. It includes (see Section Figures 1, 2 and 3): a leak-tight steel liner on the inner face, anchored to the concrete; support brackets for the polar crane girder; three vertical ribs on the outer face, for anchoring the horizontal prestressing tendons; penetrations such as Equipment Hatch, Personnel Airlocks or Fuel Transfer Tube (discussed in section 5 of Sub-chapter 12.3); local wall thickness increases and strengtheners around the Fuel Transfer Tube sleeve and Equipment Hatch. The thickness of the Inner Containment is compatible with the installation of the prestressing system, especially ducts receiving tendons. The foundation raft of the HR Inner Containment is part of the Common Raft (see Sub-chapter 12.3 section 1). The Prestressing Gallery (see Sub-chapter 12.3 section 2) for anchoring vertical and gamma tendons is located underneath the Common Raft. This gallery is not structurally linked to the Common Raft, there is a movement joint. Protection against airplane crash is provided by the APC Shell (see Sub-chapter 12.3 section 7), which is a monolithic reinforced concrete structure that goes over the building and can be considered as a separate structure. The HR Inner Containment is physically separated from this shell to protect it from vibrations and displacements produced in the event of an impact, the only connection being through the HR Inner Containment gusset and the Common Raft Prestressing The inner containment wall and the dome are prestressed concrete structures. Prestressing is provided by an arrangement of fully cement grouted bonded steel tendons. The tendon characteristics are given in the civil nuclear code ETC-C (see Sub-chapter 3.8 section 4):

42 Page No.: 35 / 207 Each horizontal tendon makes a complete loop of the containment and is anchored within one of the inner containment wall ribs. Each horizontal tendon is tensioned at both ends, at the same rib. The vertical tendons are divided into two families: o o Short vertical tendons: tensioned at their lower end located inside the prestressing gallery which is located underneath the Common foundation Raft, and passively anchored at { SCI removed } at their upper end. This upper end is embedded within the HR Inner Containment wall. Long vertical tendons: tensioned at their lower end located in the prestressing gallery and are passively anchored in the dome ring. The gamma tendons are vertical tendons which extend into the dome and are tensioned at both ends. The lower end is anchored in the prestressing gallery and the upper end is anchored in the dome ring at the opposite side of the HR Inner Containment wall. The route taken by the tendons is globally straight with some possible deviations in the vicinity of large penetrations. The tendons quantities are summarised below: horizontal tendons: { SCI removed }, short vertical tendons: { SCI removed }, long vertical tendons: { SCI removed } (of which 4 are instrumented and un-bonded), gamma tendons: { SCI removed }. As mentioned above, tendons are fully injected with a cement grout after tensioning. This grouted bonded prestressing technology ensures a long term protection of the tendons against corrosion and prevents the ingress of water or other aggressive agents. Moreover, in the event of failure of a tendon, part of the tensioning force will continue to be transmitted to the structure through bond between the tendon and the concrete wall. Specific tests on prestressing mock-ups are to be performed to confirm the effectiveness of the grouting and the correct filling of the prestressing ducts. These mock-ups are at 1:1 scale. Full length mock-ups are used to test the horizontal tendon ducts. For gamma tendon ducts, the vertical and dome part of the duct are represented by separate mock-ups each of which is at full length for the section of the duct simulated. A safety justification for the EPR prestressing system [Ref. 16] shows that the design will provide adequate reliability of prestressing through the life of the EPR containment structure and that the proposed design meets the ALARP principle. The arguments presented are as follows: Design and construction methods are robust enabling construction to high standards of quality and reliability. The tendons are adequately protected from corrosion during installation and subsequent operation of the plant. Feedback experience from Nuclear Power Plant (NPP) containments constructed using grouted tendons confirms that no evidence of corrosion has been found.

43 Page No.: 36 / 207 Results of Finite Element models show that the containment structure is tolerant to multiple failures of entire tendons even it tendon failures occur in close proximity to one another. Monitoring of concrete strains during operation and periodic pressure testing provides a high level of confidence that any significant degradation of the prestressing tendons during the plant lifetime, endangering the containment structural integrity would be detectable. A specific optimum layout of strain gauges is defined in a specific report [Ref. 17] and implemented in the containment wall to improve the detectability of hypothetical failures of prestressing tendons Steel Liner The steel liner comprises steel plate sections of { SCI removed }-thickness, welded together, covering the entire internal surface of the HR Inner Containment walls, dome and Common Foundation Raft (see Section Figure 4). To ensure continuity the liner base is located between the top of the Common Foundation Raft and the underside of the support slab for the Internal Structures. A continuous anchorage system is welded to the steel liner plates and is integrated into the concrete (see Section Figure 5). It comprises continuous vertical and horizontal steel anchors. Inside the areas enclosed by the crossing continuous anchors are meshes of smaller stud anchors (spaced { SCI removed } apart both horizontally and vertically). The layout of the anchorage system is justified in the civil nuclear code ETC-C (see Sub-chapter 3.8 section 4). The role of the anchoring system is to stiffen the liner and to ensure its strength during construction, operation and accident scenarios. The continuous anchors transmit concrete deformation to the steel liner. They limit the movement of the liner relative to the concrete due to differences of thickness, temperature conditions or elasto-plastic conditions. In addition, the continuous steel anchors provide the liner with sufficient rigidity during its assembly and during the construction phase. The spacing of the stud anchorages is such that local buckling, which can occur (due to geometrical manufacturing defects) during prestressing or when heated, remains within acceptable limits. Numerous anchor plates for equipment support and the inner containment penetrations (see Sub-chapter 12.3 section 5) are incorporated into the liner and its anchorage system. The liner base is not anchored to the Common Raft. It is set directly onto the Common Raft and covered with concrete at the pouring of the HR Internal Structures, without any mechanical connection to either of these structures Layout Layout is not relevant for the HR Inner Containment (see section for general arrangement of the structure).

44 Page No.: 37 / Interfaces with Surrounding Structures The following structures have structural interfaces with the HR Inner Containment: Common Raft, Outer Containment, Prestressing Gallery, Reactor Building Internal Structures, Fuel Building (HK [FB]), Safeguard Electrical and Mechanical Buildings (HL [SB]). The HR Inner Containment is linked to the HR Outer Containment through the gusset that ensures the embedment of these two structures and their connection to the Common Raft. The gusset provides a monolithic connection between these three structures. Elsewhere, the HR Inner Containment is separated from the Outer Containment by a { SCI removed } wide annulus. The HR Inner Containment is connected to the Prestressing Gallery through the vertical and gamma prestressing tendons anchored into the soffit of the gallery, which ensure the prestressing of the HR Inner Containment. The Prestressing Gallery and Common Raft interface is described in Sub-chapter 12.3 section 1 relating to the Common Raft. The HR Inner Containment gusset is in direct contact with the HR Internal Structures (which are cast directly on top of the liner) along the cylindrical part of the gusset. In case of severe elevation of temperature inside the HR Building, forces are transmitted from the Internal Structures to the Inner Containment (more particularly to the gusset) due to the differential dilatation of the Internal Structures. The HR Inner Containment liner base is set on the Common Raft without a specific anchoring system and the HR Internal Structures are poured directly onto the liner base. The HR Inner Containment is crossed by Containment Penetrations (See Sub-Chapter section 5), including the personnel and equipment accesses, that provide connection to the HL and HK Buildings Internal Civil Engineering Interfaces The list of penetration sleeves embedded inside the HR Inner Containment is given [Ref. 14]. The connection between the penetration sleeves and the HR Inner Containment are discussed in Sub-chapter 12.3 section 5. General interfaces between main and other structures in the HR Internal Structures, including a description of how other structures are considered within the civil engineering design, are described in Sub-chapter 3.3 section 3. Requirements for safety related interfaces between civil engineering and mechanical equipment are defined in Sub-chapter 3.5. The way these interfaces are handled in the engineering sequence is described in the design process note [Ref. 18] and in Sub-chapter 22.3.

45 Page No.: 38 / 207 The HR Internal Structures will be protected from natural (lightning strikes) and industrial electro-magnetic interference. An earthing system composed of earthing grids, Faraday cages, (made by means of interconnecting structural reinforcement), lightning arrestors and lightning rods is designed such that the residual risk of overvoltage in each EPR room is consistent with the equipment s overvoltage immunity DESIGN SUBSTANTIATION Design Substantiation Process A description of the Design Substantiation Process can be found in Sub-chapter 3.3 section 1. Design substantiation for civil structures is provided through a number of deliverables produced at different stages throughout the engineering sequence. These deliverables will be completed before the lifting of the relevant concrete hold point. A list of these documents can be found in Sub-chapter 3.3 section Output of the Design An explanation of design outputs and how they are used to demonstrate that safety requirements have been met can be found in Sub-chapter 3.3 section CONSTRUCTION The purpose of this section is to examine if there are any construction sequencing or constructability issues which need to be addressed for the design of the HR Inner Containment, which could in turn impact upon the safety justification to commence construction. The Civil Work Contractor (CWC) has provided detailed documents (Civil Works Method Statements [Ref. 19] and [Ref. 20]) concerning its construction methods. The relevant sections of these documents have been considered in order to ensure consistency between those methods and the safety requirements. An example of a construction method impacting the design is discussed below. At this stage no dedicated method statement from the CWC concerning anticipated prestressing is available. Nevertheless, a description of the main principles of the anticipated prestressing are provided hereafter: The Inner Containment wall will be built up to the upper level of short vertical tendons. For each vertical short tendon, the bearing device of the upper anchorage will be embedded in the last concrete lift poured. The short vertical tendons will be tensioned according to the detailed prestressing sequence to be defined during detailed design. Around { SCI removed } horizontal tendons will be tensioned according to the detailed prestressing sequence to be defined during detailed design. The Inner Containment wall concreting will be continued and the upper anchorage of each short vertical tendon will be embedded within the wall.

46 Page No.: 39 / 207 The long vertical tendons will be tensioned according to the detailed prestressing sequence to be defined during detailed design. A first batch of gamma tendons will be tensioned according to the detailed prestressing sequence to be defined during detailed design. The rest of the horizontal tendons will be tensioned according to the detailed prestressing sequence to be defined during detailed design. Finally, the instrumented long vertical tendons will be tensioned according to the detailed prestressing sequence to be defined during detailed design. SECTION FIGURE 1 : { SCI REMOVED } { This figure contains SCI and has been removed } SECTION FIGURE 2 : { SCI REMOVED } { This figure contains SCI and has been removed } SECTION FIGURE 3 : { SCI REMOVED } { This figure contains SCI and has been removed } SECTION FIGURE 4 : { SCI REMOVED } { This figure contains SCI and has been removed }

47 Page No.: 40 / 207 SECTION FIGURE 5: STEEL LINER ANCHORAGE SYSTEM PRINCIPLE (FOR INFORMATION ONLY)