| Sub-Chapter 6.4 - Habitability of the Control Room.pdf |
Pre-Construction Safety Report (PCSR) |
20-06-2023 |
82.42 KB |
| Sub-chapter 6.4 covers the habitability of the main control room during all events that might result in a radioactive release. This includes all the equipment, supplies and procedures necessary to enable the operators to remain in the main control room and take actions required to operate the plant safely in normal conditions, and to maintain it in a safe condition following an accident. The safety requirements, design criteria, system design description and operating parameters are given.
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| Sub-Chapter 6.3 - Safety Injection System.pdf |
Pre-Construction Safety Report (PCSR) |
20-06-2023 |
1.4 MB |
| Sub-chapter 6.3 provides a description of the safety injection/residual heat removal system and the in-containment refuelling water storage tank, including the safety and functional requirements; system description; system parameters and operating conditions; design criteria and assumptions; preliminary safety evaluation; testing, inspection and maintenance details.
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| Sub-Chapter 6.2 - Containment Systems.pdf |
Pre-Construction Safety Report (PCSR) |
20-06-2023 |
1.95 MB |
| Sub-chapter 6.2 provides a description of the containment systems and components, including as appropriate: safety and functional requirements; system description; relevant operating conditions; design criteria to be applied; materials and material properties; design details and calculations; preliminary safety analysis; and testing, inspection and maintenance details. The systems covered include: the annulus ventilation system, containment isolation system, combustible gas control system, leak rate control and testing system, core melt stabilisation system and containment heat removal system.
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| Sub-Chapter 6.1 – Materials.pdf |
Pre-Construction Safety Report (PCSR) |
20-06-2023 |
80.32 KB |
| Sub-chapter 6.1 defines the standard requirements for metallic and non-metallic materials used for the equipment in the nuclear island of the EPR.
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| Appendix 6 - MER Calculations - BDR Results.pdf |
Pre-Construction Safety Report (PCSR) |
20-06-2023 |
377.44 KB |
| Appendix 6 contains the sections from the generic EPR™ Basic Design Report 99 which describe calculations of mass and energy release into the containment which have not been specifically analysed for the UK EPR™ Pre-Construction Safety Report. The cases reported are: rupture of the pressuriser surge line; double-ended guillotine break at the cold leg of main coolant line; and double-ended guillotine break of a main steam line inside the containment building.
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| Sub-Chapter 5.5 - Reactor Chemistry.pdf |
Pre-Construction Safety Report (PCSR) |
20-06-2023 |
743.38 KB |
| Sub-chapter 5.5 provides a description of the primary and secondary side chemistry, including the chemistry of auxiliary systems. The choice of materials for primary and secondary systems is a key parameter to ensure the safe operation of the unit. Taking this into account, the chemistry is optimised to ensure the integrity of materials and to reduce radiation fields. The main chemistry parameters are described and justified in this chapter, including the design optimisation which provides the means to achieve the objectives of nuclear safety, radiation protection, material and equipment integrity, minimisation of environmental impact, hazard protection (explosion risk) and operational performance.
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| Sub-Chapter 5.4 - Components and Systems Sizing.pdf |
Pre-Construction Safety Report (PCSR) |
20-06-2023 |
2.22 MB |
| Sub-chapter 5.4 provides a description of the main reactor coolant systems and components, including as appropriate: the relevant operating conditions and interfaces; the design criteria to be applied; materials and material properties; design details and calculations; safety evaluation to demonstrate that primary components and piping meet High Integrity Component requirements, including assessment of mechanical integrity in accident conditions, the qualified manufacturing inspections proposed as well as fracture toughness proposals. The systems and components covered include: the reactor coolant pumps, the steam generators, the reactor coolant piping, the pressuriser and pressuriser relief line, valves associated with the reactor coolant pressure boundary, pressuriser pressure safety relief valves and severe accident depressurisation valves, and the primary component supports.
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| Sub-Chapter 5.3 - Reactor Vessel.pdf |
Pre-Construction Safety Report (PCSR) |
20-06-2023 |
839.83 KB |
| Sub-chapter 5.3 describes the reactor pressure vessel, including the design operating conditions, design requirements, materials used, and applicable mechanical design rules. A safety evaluation is provided to demonstrate RPV meets High Integrity Component requirements, including a description of the fracture mechanics analyses performed to assess the margins to fast fracture, qualified manufacturing inspections and fracture toughness proposed for RPV. Materials ageing mechanisms and materials irradiation monitoring are described.
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| Sub-Chapter 5.2 - Integrity of the Reactor Coolant Pressure Boundary (RCPB).pdf |
Pre-Construction Safety Report (PCSR) |
20-06-2023 |
296.15 KB |
| Sub-chapter 5.2 describes how the integrity of the reactor coolant pressure boundary is ensured. The design rules and material specifications applicable to the reactor coolant pressure boundary are summarised. A description of the requirements applied to High Integrity Components is given for 'non-breakable'; components based on specific requirements applied to the design, manufacture, inspection and in service surveillance. In addition, the design principles applied to non-breakable components are compared with the requirements conventionally applied to Incredibility of Failure Components in UK power reactors. The design criteria for the over-pressure protection system are given and an outline of the in-service inspection requirements is presented.
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| Sub-Chapter 5.1 - Description of the Reactor Coolant System.pdf |
Pre-Construction Safety Report (PCSR) |
20-06-2023 |
654.44 KB |
| Sub-chapter 5.1 describes the functional role of the reactor coolant system, together with the design assumptions, fluid characteristics and design description of the key components (reactor vessel, pressuriser, reactor coolant pumps and steam generators). System parameters are given for both normal operating conditions and standard shutdown states. The main control functions are outlined: reactor coolant system pressure control, pressuriser level control, reactor coolant system loop level control, steam generator level control, and reactor coolant pump standstill seal system actuation.
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| Sub-Chapter 5.0 - Safety Requirements.pdf |
Pre-Construction Safety Report (PCSR) |
20-06-2023 |
96.57 KB |
| Sub-chapter 5.0 describes the safety functional requirements and design criteria used in the functional design of the reactor coolant system and its auxiliary systems, together with a brief outline of testing requirements.
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| Sub-Chapter 4.5 - Functional design of reactivity control.pdf |
Pre-Construction Safety Report (PCSR) |
20-06-2023 |
91.84 KB |
| Sub-chapter 4.5 describes the safety functional requirements and design criteria used in the functional design of the reactivity control systems, including the control rod drive system, the chemical and volume control system, the extra boration system and the safety injection system.
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| Sub-Chapter 4.4 - Thermal and hydraulic design.pdf |
Pre-Construction Safety Report (PCSR) |
20-06-2023 |
794.52 KB |
| Sub-chapter 4.4 describes the safety functional requirements and design criteria used in the thermal hydraulic design of the reactor core. Details are provided of the various limiting physical phenomena, such as departure from nucleate boiling, and flow instability. The thermal hydraulic characteristics of the reactor core are given, together with a description of the transient analysis methodology, tools and design data. A description of the core instrumentation requirements is also provided.
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| Sub-Chapter 4.3 - Nuclear Design.pdf |
Pre-Construction Safety Report (PCSR) |
20-06-2023 |
718.93 KB |
| Sub-chapter 4.3 describes the safety functional requirements and design criteria used in the nuclear design of the fuel and reactivity control system. It provides an overall description of the core, together with a definition of the calculated power distributions, the fuel and moderator reactivity coefficients, the core control requirements and principles, and means by which these are achieved, the calculation of shutdown margins, the preliminary criticality design criteria and assumptions, and the residual heat characteristics. A brief review of the methods and tools used to determine neutron and gamma ray flux attenuation between the core and the pressure vessel is also given.
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| Sub-Chapter 4.2 - Fuel System Design.pdf |
Pre-Construction Safety Report (PCSR) |
20-06-2023 |
163.41 KB |
| Sub-chapter 4.2 lists the safety requirements to be met in the design of the fuel and control rod assemblies, and includes a design description and evaluation of both the fuel and control rod assemblies.
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| Sub-Chapter 4.1 - Summary description.pdf |
Pre-Construction Safety Report (PCSR) |
20-06-2023 |
154.18 KB |
| Sub-chapter 4.1 provides a summary description of the core, fuel and reactivity control, and presents the main parameters used in the core design, the assumptions considered at the present stage of the UK EPR™ design concerning the mechanical characteristics of the fuel assemblies, and the objectives of the nuclear and thermal-hydraulic design.
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| Appendix 4 - Computer codes used in Chapter 4.pdf |
Pre-Construction Safety Report (PCSR) |
20-06-2023 |
84.77 KB |
| Appendix 4 provides an outline description of the computer codes used in the analyses presented in Chapter 4.
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| Sub-Chapter 3.8 - Codes and standards used in the EPR design.pdf |
Pre-Construction Safety Report (PCSR) |
20-06-2023 |
464.66 KB |
| Sub-chapter 3.8 gives an overview of the codes and standards used in the EPR™ design. The set of codes and standards is part of the French regulatory regime, therefore a general description of the structure of French safety regulation is provided. This is followed by a description of the main content of the codes: the technical code for mechanical equipment (RCC-M), the technical code for critical defect size calculations for mechanical components (RSE-M Appendix 5.4), the technical code for electrical equipment (RCC-E), the EPR™ Technical Code for Civil Works (ETC-C), and the EPR™ Technical Code for Fire Protection (ETC-F), including comparisons with international practice e.g. ASME and IAEA standards.
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| Sub-Chapter 3.7 - Conventional Risks of Non-Nuclear Origin.pdf |
Pre-Construction Safety Report (PCSR) |
20-06-2023 |
125.54 KB |
| Sub-chapter 3.7 covers conventional risks of non-nuclear origin on the site, i.e. risks potentially induced by the presence of non-nuclear facilities and human activities on the site. The proposed methodology, outlined in this sub-chapter, aims to demonstrate that all potential 'conventional' risks have been identified and dealt with and that their consequences are acceptable for the environment, and especially for members of the public off-site. The methodology also considers the consequences of non-nuclear origin on safety-related installations located on the site.
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| Sub-Chapter 3.6 - Qualification of electrical and mechanical equipment for accident conditions.pdf |
Pre-Construction Safety Report (PCSR) |
20-06-2023 |
328.13 KB |
| Sub-chapter 3.6 outlines the qualification of equipment for accident conditions, including severe accidents. The purpose of qualification is to demonstrate that the equipment can fulfil its required function during accident conditions. The functions to be qualified (based on analyses of both functional requirements and requirements relating to post-accident operation) and their associated requirements are described. The design data required for qualification, primarily environmental conditions (pressure, temperature and irradiation) are described, together with the qualification methods, standards and practices used, and the arrangements made to maintain qualification during manufacturing and operation (qualification maintenance).
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