Public Version of UK EPR™ GDA Submission
 EDF and AREVA have obtained for generic design acceptance for the EPR™ reactor from the UK regulatory authorities. The UK EPR™ design is based on the Flamanville EPR™ (FA3) plant being built in France but incorporates a number of design changes agreed through the Generic Design Assessment with the UK Regulators.  The UK Office for Nuclear Regulation has issued guidance defining the process by which they will confirm their acceptance of new designs of nuclear power stations for construction in the UK. Similarly the UK Environment Agency has proposed a process by which they will assess generic design issues associated with discharges and wastes produced by new nuclear power stations. |
EPR Submission
PCSRThe Pre-Construction Safety Report (PCSR) presents a detailed description of the architecture of the UK EPR™ systems and components, their safety functions and reliability and availability requirements, and an explanation of the codes and standards that have been used in the design. The PCSR aims to demonstrate that sufficient analysis and engineering substantiation has been performed to give high confidence that the UK EPR™ design meets its declared safety objectives. The PCSR has undergone detailed assessment by HSE.- -- Chapter 1 - Introduction and General DescriptionChapter 1 describes the overall purpose and scope of the Pre-Construction Safety Report (PCSR), within the scope of the Generic Design Assessment (GDA) process. It gives a general overview of the UK EPR™ buildings, structures and main power plant systems. A comparison is provided of the data for the main reactor systems against the equivalent data for the latest generations of reactors built in France (N4 Reactors) and Germany (KONVOI Reactors). Additionally, there is an overview of the structure of the UK regulations, the associated regulatory framework and key relevant UK regulations, together with an outline of the design and safety assessment process for the EPR™ within France, Finland and the USA, and an overview of comparisons of the EPR™ design against international safety standards.- -- Sub-Chapter 1.1 - IntroductionSub-chapter 1.1 describes the overall purpose and scope of the Pre-Construction Safety Report (PCSR), within the scope of the Generic Design Assessment (GDA) process. It also describes the structure and layout of the PCSR, and provides an outline Table of Contents.- file size 97 Kb Sub-Chapter 1.2 - General Description of the UnitSub-chapter 1.2 provides a general overview of the UK EPR™. It gives a brief description of the buildings and structures of UK EPR™ unit, and describes the criteria for location and installation of these. It also provides a description of the main power plant systems, together with a brief overview of the general operating principles for the UK EPR™.- file size 1,426 Kb Sub-Chapter 1.3 - Comparison with reactors of similar designSub-chapter 1.3 consists of a comparison table of the data for the main reactor systems of the UK EPR™ against the equivalent data for the latest generations of reactors built in France (N4 Reactors) and Germany (KONVOI Reactors), from which the EPR™ design is derived.- file size 146 Kb Sub-Chapter 1.4 - Compliance with regulationsSub-chapter 1.4 provides an overview of the UK regulations to be complied with by the UK EPR™ design. An overview of the structure of the UK regulatory regulations and the associated regulatory framework is provided, followed by an outline of the key relevant UK regulations. The list is not exhaustive but gives a general overview of the basic regulatory requirements.- file size 155 Kb Sub-Chapter 1.5 - Safety assessment and international practiceSub-chapter 1.5 provides an overview of the design and safety assessment process for the EPR™ within France, Finland and the USA, together with an overview of comparisons of the EPR™ design against international safety standards (the Western European Nuclear Regulators' Association (WENRA) reference levels, International Atomic Energy Agency (IAEA) Safety Standards, and the European Utility Requirements for LWR nuclear power plants).- file size 131 Kb Chapter 2 - Generic Site Envelope and DataChapter 2 presents a summary of the site data used in the safety analysis presented in the PCSR. The data are considered to be typical of UK coastal sites in England and Wales on which nuclear power stations have been sited. An outline of the site data which need to be considered in the environmental impact assessment of a reactor design, together with the methodologies to be used for the baseline studies during the site specific phase, is also presented. In order to carry out environmental impact studies for Generic Design Assessment (GDA), a set of site parameters representative for the UK is determined. It will be necessary to further update this chapter to include site specific details and to utilise the available site-specific data.- -- Sub-Chapter 2.1 - Site Data used in the Safety AnalysesSub-chapter 2.1 presents a summary of the site data used in the safety analysis presented in the PCSR. The data are considered to be typical of UK coastal sites in England and Wales on which nuclear power stations have been sited. The data include: the external heat sink and the power supply (grid connections), the bounding site data used for the off-site consequence risk assessment and external hazards assessment, and the assumptions used in evaluation of the accidental radiological consequences of accidents. It will be necessary to further update this chapter to include site specific details and to utilise the available site-specific data.- file size 138 Kb Sub-Chapter 2.2 - Site environmental characteristicsSub-chapter 2.2 presents the site data which need to be considered when assessing the environmental impacts of the UK EPR™ design (e.g. terrestrial and aquatic characteristics, habitats data). The sub chapter outlines the information and methodologies to be used for the baseline studies during the site specific phase. The sub chapter also includes a set of site parameters appropriate for the UK, i.e. representative of a range of potential coastal and estuarine locations, in order to carry out environmental impact studies for Generic Design Assessment (GDA). It will be necessary to further update this sub-chapter to include site specific details and to utilise available site-specific data.- file size 205 Kb Chapter 3 - General Design and Safety AspectsChapter 3 covers the overall safety approach implemented in the UK EPR™ plant design. It provides a summary of the main EPR™ design requirements, including the concept of defence in depth, and a description of the technical approach adopted to meet these requirements. The classification approach adopted for the EPR™ design, and the link between the safety classification and the design requirements for structures, systems and components are explained. The safety requirements applicable to the design of the safety-classified civil structures are set out, together with the safety related interfaces between those and other systems or buildings. The mechanical components, rods mechanisms, vessel internal and other components are presented. The qualification of equipment for accident conditions is outlined, including the functions to be qualified, their associated requirements and design data, and the qualification methods, standards and practices used. An overview of the codes and standards used in the EPR™ design for mechanical equipment, electrical equipment, civil works and fire protection is provided. Finally, a brief description of the computer codes used in the EPR™ design is given.- -- Sub-Chapter 3.1 - General Safety PrinciplesSub-chapter 3.1 describes the safety approach implemented in the UK EPR™ plant design, which is based on the concept of defence in depth. It provides a summary of the main EPR™ design requirements and a description of the technical approach adopted to meet these requirements. It refers to other chapters in the PCSR in which the applicable safety requirements are set out in more detail. Although the sub chapter is concerned mainly with design principles adopted to meet requirements of the French Safety Authority, reference is also made to supplementary "UK EPR™ safety design objectives" that would be demonstrated for an EPR™ operating in the UK to address UK specific safety principles.- file size 459 Kb Sub-Chapter 3.2 - Classification of structures, equipment and systemsSub-chapter 3.2 presents the classification approach adopted for the EPR™ design, the aim of which is to utilise a deterministic approach for ensuring that the plant is designed, manufactured, constructed commissioned and operated so that the appropriate level of reliability and integrity is achieved for all structures, systems and components (SSCs). The classification process involves the systematic assessment of the importance to nuclear safety of each SSC, its allocation to a Safety Feature Group (SFG) and Safety Feature (SF) and to a safety class on the basis of this safety significance. The safety class allocated to an SSC defines the design, testing and maintenance measures to be applied in its design, construction, commissioning, and operation. The link between the safety classification and the SSC design requirements is explained (Components, Structures); the different types of requirements are defined and an explanation is given of how they are applied. The methodology for utilising a Probabilistic Safety Assessment (PSA) approach for performing verification of the deterministically derived classification of SSCs is explained and supported by example applications of the UK classification approach.- file size 714 Kb Sub-Chapter 3.3 - Design of Category 1 Civil StructuresSub-chapter 3.3 sets out the requirements applicable to the design of the safety-classified civil structures in the UK EPR™. The safety requirements have been established as follows: a review was carried out of the relevant structures, identifying the various components of the structures to which particular requirements would apply; the various load cases applicable to the structures were specified, covering both normal reactor operation and fault conditions (including internal and external hazards), to which the structures may be subjected; the load cases applicable to particular components of the civil structures, and the associated safety requirements, were identified. The safety requirements were then incorporated into the design of the civil structures and into the definition of their associated behavioural requirements. The C1 and C2 requirement levels, as defined in Sub-chapter 3.2, apply to the safety classified civil structures, which include: the steel lined inner containment, the containment penetrations, the concrete and steel structure internal to the reactor building, seismically classified buildings in the nuclear island, and the foundation raft.- file size 1,680 Kb Sub-Chapter 3.4 - Mechanical systems and componentsSub-chapter 3.4 describes the safety requirements applied to the design of the mechanical systems and components. It defines the operating conditions upon which the equipment design is based; the nature of the loads to be considered for all the pressurised equipment, in particular the primary and secondary pressure system boundaries; the rules for combining the loads and the criteria to be used; and the analytical methods used. A structural analysis of the reactor coolant pressure boundary is presented, together with analysis of hydraulic loading on the main steam system. Analyses are presented demonstrating that the criteria relative to overpressure risks for the primary and secondary pressure boundaries are met. A discussion of vibration testing and analyses is included. The specific mechanical systems considered in detail here are: the control rod drive mechanism and the reactor pressure vessel upper core support structures and lower internals.- file size 3,760 Kb Sub-Chapter 3.5 - Safety Related InterfacesSub-chapter 3.5 describes safety related interfaces in three areas: mechanical equipment and civil structures in the nuclear island; electrical equipment and civil engineering; and the interfaces between nuclear and non nuclear plant. In each area, the Interfaces, their functions and design, and the specific requirements for the systems and structures are given.- file size 126 Kb Sub-Chapter 3.6 - Qualification of electrical and mechanical equipment for accident conditionsSub-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).- file size 324 Kb Sub-Chapter 3.7 - Conventional Risks of Non-Nuclear OriginSub-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.- file size 126 Kb Sub-Chapter 3.8 - Codes and standards used in the EPR designSub-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.- file size 461 Kb Appendix 3 - Computer codes used in Chapter 3Appendix 3 contains brief descriptions of the computer codes used in the analyses presented in Chapter 3, including the structural analysis of the main civil structures (Sub chapter 3.3) and the mechanical analysis of the reactor coolant pressure boundary (Sub chapter 3.4)- file size 325 Kb Chapter 4 - Reactor and Core DesignChapter 4 summarises the nuclear and thermal-hydraulic characteristics of the reactor core, 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. The safety functional requirements and design criteria used in the design of the fuel and control rod assemblies, the nuclear design and the thermal hydraulic design of the reactor core, and the functional design of the reactivity control systems are defined. Details are given of the design characteristics of the fuel and control rod assemblies and the reactor core, together with a description of the methods, tools and design data used.- -- Sub-Chapter 4.1 - Summary descriptionSub-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.- file size 150 Kb Sub-Chapter 4.2 - Fuel System DesignSub-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.- file size 159 Kb Sub-Chapter 4.3 - Nuclear DesignSub-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.- file size 1,279 Kb Sub-Chapter 4.4 - Thermal and hydraulic designSub-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.- file size 809 Kb Sub-Chapter 4.5 - Functional design of reactivity controlSub-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.- file size 92 Kb Appendix 4 - Computer codes used in Chapter 4Appendix 4 provides an outline description of the computer codes used in the analyses presented in Chapter 4.- file size 81 Kb Chapter 5 - Reactor Coolant System and Associated SystemsChapter 5 provides a description of the main reactor coolant systems and components, including as appropriate: the functional role of each system; the relevant operating conditions, system parameters, and interfaces; the design criteria to be applied; materials and material properties; design details and calculations; safety evaluation to demonstrate primary components and piping meet High Integrity Component (HIC) requirements, including assessment of mechanical integrity in accident conditions; the qualified manufacturing inspection proposed as well as a fracture toughness proposal. The systems and components covered include: the reactor pressure vessel, reactor coolant pumps, steam generators, reactor coolant piping, pressuriser and pressuriser relief line, valves, primary component supports, and the main control functions. A description of the requirements applied to High Integrity Components and "non-breakable" components is also detailed here. A description of the primary and secondary side chemistry, including the chemistry of auxiliary systems is also given.- -- Sub-Chapter 5.0 - Safety RequirementsSub-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.- file size 93 Kb Sub-Chapter 5.1 - Description of the Reactor Coolant SystemSub-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.- file size 651 Kb Sub-Chapter 5.2 - Integrity of the Reactor Coolant Pressure Boundary (RCPB)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.- file size 292 Kb Sub-Chapter 5.3 - Reactor VesselSub-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.- file size 853 Kb Sub-Chapter 5.4 - Components and Systems SizingSub-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.- file size 2,265 Kb Sub-Chapter 5.5 - Reactor ChemistrySub-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.- file size 734 Kb Chapter 6 - Containment and Safeguard SystemsChapter 6 provides a description of the containment and safeguard systems, including as appropriate: safety and functional requirements; system description; system parameters and operating conditions; design criteria and assumptions; materials and material properties; design details and calculations; safety assessment; and testing, inspection and maintenance details. The containment 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. The safeguard systems covered include: the safety injection/residual heat removal system and the in containment refuelling water storage tank; the emergency feed water system; the extra boration system; and the main steam relief train system. This chapter also describes the principles covering in-service inspection on the nuclear island pressurised equipment and addresses the habitability of the main control room during events resulting in a radioactive release.- -- Sub-Chapter 6.1 - MaterialsSub-chapter 6.1 defines the standard requirements for metallic and non-metallic materials used for the equipment in the nuclear island of the EPR™.- file size 81 Kb Sub-Chapter 6.2 - Containment SystemsSub-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.- file size 1,993 Kb Sub-Chapter 6.3 - Safety Injection SystemSub-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.- file size 1,432 Kb Sub-Chapter 6.4 - Habitability of the Control RoomSub-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.- file size 79 Kb Sub-Chapter 6.5 - In-Service Inspection Principles (excluding main primary and secondary systems)Sub-chapter 6.5 presents the principles covering in-service inspection on the nuclear island pressurised equipment, including pressurised accessories, safety accessories and the parts welded to the pressurised containments. In-service inspection is preventive maintenance, consisting of non-destructive examinations and checks on equipment, which is systematically scheduled and implemented during planned outages. Included here are the safety requirements, areas of applicability, and scope, nature and frequency of inspections.- file size 122 Kb Sub-Chapter 6.6 - Emergency Feedwater SystemSub-chapter 6.6 presents the safety and functional requirements, design criteria, system design description, operating parameters, preliminary safety analysis, and testing, inspection and maintenance details, for the emergency feed water system. The emergency feed water system is a safety system which, in the event that normal operational feed water systems become unavailable, supplies the necessary feed water to the steam generators.- file size 299 Kb Sub-Chapter 6.7 - Extra Boration SystemSub-chapter 6.7 presents the safety and functional requirements, design criteria, system design description, system parameters and operating conditions, preliminary safety analysis, and testing, inspection and maintenance details, for the extra boration system. The extra boration system is a safety system which ensures boration of the reactor coolant, under fault conditions, to enable the core to be brought to a controlled or safe shutdown state.- file size 515 Kb Sub-Chapter 6.8 - Main steam relief train systemSub-chapter 6.8 presents the safety and functional requirements, design criteria, system design description, system parameters and operating conditions, preliminary safety analysis, and testing, inspection and maintenance details, for the main steam relief train system. The main steam relief train system is part of the main secondary cooling system and is used for dumping heat from the steam generators into the atmosphere when the condenser is unavailable.- file size 203 Kb Appendix 6 - MER Calculations - BDR ResultsAppendix 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.- file size 374 Kb Chapter 7 - Instrumentation and ControlChapter 7 presents the design principles of the Instrumentation and Control (I&C) system and the general architecture of the system. System descriptions are provided for the control and protection systems, and the instrumentation, including safety and functional requirements; design criteria; architecture, technology, and power supply, and testing, and maintenance details. Information is provided on the I&C procedures and tools used for the design of three I&C platforms, including when relevant the reliability, availability and functional requirements; the software and hardware programming design and configuration; system integration, installation, and commissioning; verification and validation.- -- Sub-Chapter 7.1 - Design principles of the Instrumentation and Control systemsSub-chapter 7.1 presents the design principles of the overall I&C system, including the safety, functional and design requirements; classification of the I&C systems and equipment; and design criteria for the I&C system architecture, including the concept of defence in depth and general priority rules. An outline of the management and quality procedures covering the design and development of the I&C systems is provided as well as a summary of Claims, Arguments and Evidence for I&C systems.- file size 168 Kb Sub-Chapter 7.2 - General architecture of the Instrumentation & Control systemsSub-chapter 7.2 describes the general architecture of the overall I&C system, including the interfaces between the various system levels and operating conditions and explains how the I&C architecture fulfils the design requirements such as reliability, independence and diversity. The layout of the I&C equipment in different rooms (mainly in the safeguard and electrical buildings) is discussed. The qualification process and principles are defined, in order to demonstrate that the various instrumentation & control components and systems are capable of responding to design requirements for safety performance.- file size 639 Kb Sub-Chapter 7.3 - Class 1 Instrumentation & Control systemsSub-chapter 7.3 describes the protection system and the safety information and control system. For each system, the safety and functional requirements; system description; design criteria; architecture, technology, and power supply, and testing, and maintenance details are provided. These class 1 systems manage category A I&C functions and are required to attain a controlled state after a fault, or failure during normal power plant operation that would lead to a fault.- file size 995 Kb Sub-Chapter 7.4 - Class 2 Instrumentation & Control systemsSub-chapter 7.4 describes the safety automation system, the reactor control surveillance and limitation system, and the non-computerised safety system. For each system, the safety and functional requirements; system description; design criteria; architecture, technology, and power supply, and testing, and maintenance details are provided.- file size 483 Kb Sub-Chapter 7.5 - Class 3 Instrumentation & Control systemsSub-chapter 7.5 describes the RRC-B safety automation system, the process information and control system, the process automation system, and the severe accident I&C system. For each system, the safety and functional requirements; system description; design criteria; architecture, technology, and power supply, and testing, and maintenance details are provided.- file size 382 Kb Sub-Chapter 7.6 - InstrumentationSub-chapter 7.6 describes the instrumentation used in the UK EPR™. It covers: conventional process instrumentation; accident and severe accident instrumentation; the process instrumentation pre-processing system; in core and ex core instrumentation; rod position measurement; reactor pressure vessel water level measurement; loose parts monitoring and vibration monitoring; radiation monitoring; and boron instrumentation. The overall safety, functional and design requirements are given, and the description of each system includes where appropriate: measurement principles; system interfaces: design rules and structure; and installation, calibration, maintenance and testing details.- file size 456 Kb Sub-Chapter 7.7 - IandC Tools Development Process and SubstantiationSub-chapter 7.7 provides information on the I&C procedures and tools used for the design of the three I&C platforms: the Teleperm XS platform for the protection system, the reactor control surveillance and limitation system, and the severe accident I&C; the SPPA-T2000 platform for the process information and control system, process automation system safety automation system; and the RRC-B safety automation system and the UNICORN platform for the non computerised safety system. For each platform, details are provided of: the reliability, availability and functional requirements; the software and hardware programming design and configuration; system integration, installation, and commissioning; verification and validation; and an overview of arrangements for operation, maintenance and modifications. Details are also given of the substantiation approach for software based systems using the Teleperm XS, SPPA-T2000 and UNICORN platforms and for the SMART devices and programmable complex electronic component, including production excellence activities and confidence building.- file size 381 Kb Chapter 8 - Electrical Supply and LayoutChapter 8 describes the role and main elements of the external power supplies for both the nuclear island, and the conventional island and balance of plant. The emergency power supplies and power distribution systems for the nuclear island, which restore power in the event of loss of off-site supplies, are also described, including the overall safety, functional and design requirements, and system architecture, and the operating role, design criteria, system parameters and operating requirements for each of the systems. Cabling design principles, including the separation and routing requirements, are presented, together with a description of the earthing and lightning protection systems and other electrical protection measures. A summary is given of the main features of the civil installations where the electrical distribution and instrumentation and control systems are installed. The main features of the prevention and protection against Common Cause Failure are given including protection against component failure, hazards, over-voltage and human errors.- -- Sub-Chapter 8.1 - External Power SupplySub-chapter 8.1 describes the role and main elements of the external power supply for the nuclear island: the main and auxiliary connections, and transformers, including their operating role and design criteria, and system description.- file size 96 Kb Sub-Chapter 8.2 - Power Supply to the Conventional Island and Balance of PlantSub-chapter 8.2 describes the role and main elements of the power supply to the conventional island and balance of plant, both the emergency and non-emergency power supplies.- file size 96 Kb Sub-Chapter 8.3 - Nuclear Island Power SupplySub-chapter 8.3 describes power distribution systems and the emergency power supplies for the nuclear island. In the event of loss of off-site power, 10 kV main diesel generators restore the power supply or, in the event of loss of both off-site supplies and on-site normal emergency supplies, the 690 V ultimate emergency diesel generators restore power to two of the four divisions. The overall safety, functional and design requirements, and system architecture, are described. For each of the systems, the operating role, design criteria, system parameters and operating requirements are given. Details are also given of the safety requirements specified for equipment qualification.- file size 603 Kb Sub-Chapter 8.4 - Specific Design PrinciplesSub-chapter 8.4 provides a general description of cabling design principles, including the requirements for separation between cableways and the various cable categories, and a description of the design principles applied to cable routing. The earthing and lightning protection systems, whose main function is to limit the risks of overvoltage on the electrical systems, and other electrical protection measures, are also described here.- file size 139 Kb Sub-Chapter 8.5 - InstallationSub-chapter 8.5 summarises the main features of the civil installations in the nuclear island, conventional island and the balance of plant where the electrical distribution and instrumentation and control systems are installed, which contribute to the safety functionality and availability of the systems.- file size 68 Kb Sub-Chapter 8.6 - Prevention and protection against common cause failureSub-chapter 8.6 summarises the main features of the prevention and protection against Common Cause Failure (CCF). CCFs are failures on demand or during a system mission period that could affect and make several pieces of equipment or components within the same system or with the same function simultaneously unavailable, where the failures are due to the same cause. The sub-chapter sets out the arguments and evidence for the safety requirements which have been specified for CCF, including component failure, both external and internal hazards, and human error within the design of the electrical systems. The availability of an internal network and associated equipment is considered, including protection against over-voltage and high frequency range phenomena.- file size 195 Kb Chapter 9 - Auxiliary SystemsChapter 9 provides details of the auxiliary systems including: safety and functional requirements; system description; system parameters and operating conditions; design criteria and assumptions; materials and material properties; design details and calculations; safety assessment; and testing, inspection and maintenance details. The auxiliary systems include the fuel handling and storage systems, including the fuel storage racks, the spent fuel cooling and purification system, and the fuel handling systems; the water systems, including the essential service water system, the component cooling water system, the demineralised water systems, the service water filtering system, and the ultimate cooling water system; the nuclear island physical and chemical sampling system, the chemical and volume control system and the reactor boron and water makeup system, and the coolant storage and treatment system; the ventilation and cooling systems; the fire protection systems; the diesel generators; the compressed air systems; the plant communication and lighting systems; and the gas distribution and storage systems.- -- Sub-Chapter 9.1 - Fuel Handling and StorageSub-chapter 9.1 describes the fuel handling and storage systems: the new fuel dry storage rack, the underwater fuel storage rack, the spent fuel cooling and purification system, fuel and other handling systems, and the liners for the fuel handling and storage pools. For each system, the safety and functional requirements, design criteria, system design description, system parameters and operating conditions, preliminary safety analysis, and testing, inspection and maintenance details, are provided.- file size 2,213 Kb Sub-Chapter 9.2 - Water SystemsSub-chapter 9.2 describes the water systems: the essential service water system, the component cooling water system, the demineralised water systems (for production, storage and distribution of demineralised water), the service water filtering system, the potable water system, and the ultimate cooling water system. For each system, the safety and functional requirements, design criteria, system description, system parameters and operating conditions, preliminary safety analysis, and testing, inspection and maintenance details, are provided.- file size 1,019 Kb Sub-Chapter 9.3 - Primary System AuxiliariesSub-chapter 9.3 describes the following systems: the nuclear island physical and chemical sampling system, which comprises three separate sampling systems, one for the primary system, a second for the secondary side of the steam generators and a third system for the treatment systems located in the effluent treatment building; the chemical and volume control system and the reactor boron and water makeup system, both of which regulate the boron concentration in the reactor coolant; and the coolant storage and treatment system, which includes the coolant storage and supply, treatment, purification and degasification systems. For each system, the safety and functional requirements, design criteria, system design description, system parameters and operating conditions, preliminary safety analysis, and testing, inspection and maintenance details, are provided.- file size 1,220 Kb Sub-Chapter 9.4 - Heating, Ventilation and Air-Conditioning SystemsSub-chapter 9.4 describes the main ventilation and cooling systems. Those systems which contribute to reducing radioactive discharges comprise: ventilation of the nuclear auxiliary building, the fuel building, and the waste treatment building; purging of the reactor building; and ventilation of the controlled areas of the safeguard buildings, the operational service centre and the access building. The systems that maintain the conditions required for the safety and habitability of the main control room include: air-conditioning of the main control room; ventilation of the containment, diesel buildings and the pumping station; and ventilation of the non-controlled area in the safeguard buildings. Also described here are the chilled water systems which produce and distribute chilled water to the ventilation systems. For each system, the safety and functional requirements, design criteria, system design description, system parameters and operating conditions, preliminary safety analysis, and testing, inspection and maintenance details, are provided.- file size 3,381 Kb Sub-Chapter 9.5 - Other Supporting SystemsSub-chapter 9.5 describes the following systems: the fire protection systems and equipment, including the active methods of fire control, detection and extinguishing, and the design principles for the door monitoring system; the main and ultimate back-up/emergency diesel generators; the compressed air systems, including the compressed air production system and the instrument and service air distribution systems; the plant communication systems, including the alarm systems, paging and internal and external telephone communication systems; lighting systems including both normal and emergency lighting arrangements; gas distribution and storage systems, for distribution of nitrogen, oxygen, hydrogen and carbon dioxide. For each system, the safety and functional requirements, design criteria, system design description, system parameters and operating conditions, preliminary safety analysis, and testing, inspection and maintenance details, are provided.- file size 972 Kb Chapter 10 - Main Steam and Feedwater LinesChapter 10 provides a general design description of the secondary cooling system, which converts thermal energy produced by the nuclear steam supply system into electrical power. The following systems are described in detail: the main steam supply system; the turbine bypass system; the steam generator blowdown system; and the main feedwater system. The High Integrity Components (HIC) principles applied to the main steam lines inside and outside the reactor building containment is also detailed here.- -- Sub-Chapter 10.1 - General DescriptionSub-chapter 10.1 provides a general design description of the secondary cooling system, which converts thermal energy produced by the nuclear steam supply system into electrical power.- file size 207 Kb Sub-Chapter 10.2 - Turbo-generator SetInformation relating to the turbo-generator is both site and operator specific, and is therefore not included in the PCSR for generic design assessment.- file size 83 Kb Sub-Chapter 10.3 - Main Steam System (safety classified part)Sub-chapter 10.3 describes the safety, functional and design requirements, the design criteria, system design description, system parameters and operating conditions, instrumentation and control details, safety assessment, details of testing, and maintenance requirements for the main steam system. The main steam supply system is designed to ensure, a supply of steam to the turbine during normal power operation, and, in certain conditions, to protect the steam generators against overpressure. The main steam lines from the steam generators up to the first terminal fixed points downstream of the Main Steam Isolation Valves (MSIV) and including the MSIV are High Integrity Components (HIC) and details of the fast fracture analysis, manufacturing inspections and fracture toughness requirements are described.- file size 752 Kb Sub-Chapter 10.4 - Other Features of Steam and Power Conversion SystemsSub-chapter 10.4 describes the turbine bypass system, which discharges the steam flow to the condenser when the turbine is unavailable; the main feedwater pump, the start-up and shutdown feedwater system, the cooling water system (for a coastal site) and the steam generator blowdown system which is used to maintain the necessary quality of the water/steam cycle in conjunction with the nuclear sampling system. Information relating to the condenser and condensate extraction systems, the remainder of the feedwater plant, and turbine gland system, is operator and/or site specific, and is not included here.- file size 385 Kb Sub-Chapter 10.5 - Integrity of the main steam lines inside and outside the containmentSub-chapter 10.5 describes how the High Integrity Components (HIC) principle is applied to the main steam lines inside and outside the reactor building containment. The main steam lines from the steam generators up to the first terminal fixed points downstream of the Main Steam Isolation Valves (MSIV) and including the MSIV are HIC. The HIC requirements are set out in this sub-chapter together with further requirements related to levels of defence-in-depth.- file size 125 Kb Sub-Chapter 10.6 - Main Feedwater SystemSub-chapter 10.6 describes the safety, functional and design requirements, the design criteria, system design description, system parameters and operating conditions, preliminary safety assessment, and specific testing arrangements for the main feedwater system. During normal power operation, the main feedwater system is used to provide the required water flow rate to the steam generators, and in certain conditions, to help isolate the steam generators.- file size 199 Kb Chapter 11 - Discharges and Waste - Chemical and RadiologicalChapter 11 provides details of the waste and effluent management systems, which include the liquid and gaseous waste treatment systems, and the solid waste treatment systems. Also given in this chapter are: details of the activity levels of the main primary system for normal operation; details of the effluent management processes (collection, processing, storage and disposal) including the sources, nature, treatment and storage of liquid and gaseous effluent discharges; information regarding the UK EPR™ radioactive solid waste and spent nuclear fuel strategy, including waste treatment and interim storage facilities and disposability; a quantitative estimate of the effluents discharged and the waste produced by the operation of the UK EPR™ reactor, including spent fuel and solid waste arisings, liquid and gaseous radioactive effluent discharges, and chemical effluent discharges; and a discussion of the UK EPR™ designs for interim storage facilities, for the radioactive Intermediate Level Waste (ILW) operational waste, and the spent fuel.- -- Sub-Chapter 11.0 - Safety RequirementsSub-chapter 11.0 describes the safety functions, and functional and design requirements, of the radioactive waste management systems, which include the liquid and gaseous waste processing systems, and the solid waste treatment system.- file size 63 Kb Sub-Chapter 11.1 - Sources of radioactive materialsSub-chapter 11.1 provides details of the activity levels of the main primary system when the reactor is in normal operation (both steady-state power operation and shutdown transient). The concentrations of fission, corrosion and activation products provided are based on realistic operational data, therefore they may be used in a generic way as interface data for effluent discharges, radiological protection and assessment of realistic dose uptake, and radiological protection requirements. Source terms used for the design and sizing of the EPR™ structures, rooms, systems, shielding and the effluent treatment systems are also defined.- file size 159 Kb Sub-Chapter 11.2 - Details of the Effluent Management ProcessSub-chapter 11.2 details the effluent management process (collection, processing, storage and disposal) for liquid and gaseous radioactive discharges and solid radioactive waste. It identifies the sources, nature, treatment and storage of liquid and gaseous effluent discharges. It discusses the chemicals associated with liquid effluent that could be discharged into the marine environment, including the purpose of each substance, the main circuits involved, the path from injection to discharge, and the proposed maximum amounts discharged. Information is also given regarding the UK EPR™ radioactive waste and spent nuclear fuel strategy, including waste treatment and interim storage and disposability. The strategy takes into account the UK EPR™ solid radioactive waste and spent fuel characteristics, and addresses the government and NDA policies for the assessment of disposability of radioactive waste. It includes the description of processes and interim storage facilities for managing the waste and spent nuclear fuel.- file size 374 Kb Sub-Chapter 11.3 - Outputs for the operating installationSub-chapter 11.3 presents the quantitative estimate of the effluents discharged and the waste produced by the operation of the UK EPR™ reactor. The estimates cover: spent fuel arisings, solid waste arisings, liquid and gaseous radioactive effluent discharges, and chemical effluent discharges, whether associated with liquid radioactive effluent or arising from operation of non-nuclear parts of the plant. The expected performance discharges are derived from recent operational feedback, taking credit for design features and expected quantifiable gains. Maximum additional annual discharges are provided, together with estimates of monthly discharges of gaseous and liquid radioactive waste.- file size 2,760 Kb Sub-Chapter 11.4 - Effluent and waste treatment systems design architectureSub-chapter 11.4 describes the EPR™ effluent and waste treatment systems for liquid, gaseous and solid waste. The role of the effluent treatment systems, in normal operation, involves: collecting effluent produced in the unit and site facilities, which may come from draining, bleeding, venting or leaks from systems that are normally or accidentally contaminated; treating, if necessary, the effluents, i.e. reducing the level of contamination in order to meet regulatory discharge limits; and discharging effluents outside the site following monitoring and accounting of the levels of activity discharged under regulatory conditions. In some accident situations which could lead to significant transfer of contamination to the buildings adjacent to the reactor building, the effluent treatment systems operate to ensure any contamination is contained, in order to limit release of radioactivity to the environment. For each system, the safety functions, functional and design requirements, the design criteria, system design description, system parameters and operating conditions, safety assessment, and specific testing arrangements, are provided as appropriate.- file size 3,441 Kb Sub-Chapter 11.5 - Interim storage facilities and disposability for UK EPRSub-chapter 11.5 discusses the UK EPR™ designs for interim storage facilities, for the radioactive Intermediate Level Waste (ILW) operational waste, and the spent fuel, produced during the UK EPR™ 60 years of operation. The need for plans to demonstrate when facilities will be developed and constructed is also addressed. Additionally, the current status of the discussions with the NDA, with respect to the disposability of UK EPR™ solid radioactive waste, and spent fuel arising during the operational period, and the discussions with LLW Repository Ltd with regard to Low Level Waste disposability, are described. This sub-chapter and supporting documents provide a basis for a potential UK EPR™ operator to develop radioactive waste management cases for both ILW and spent fuel.- file size 281 Kb Chapter 12 - Radiation ProtectionChapter 12 covers radiation protection requirements for workers in normal plant operation. The regulatory framework relating to radiation protection is defined, and the basic radiation protection principles and strategy are outlined. Details are provided of both specific activities of the fission and corrosion products and concentrations and deposited activity of important nuclides in the main circuit. Details are given of the radiation protection measures implemented on the UK EPR™ including: radiation protection classification and zoning, design rules for equipment, rooms and ventilation systems; shielding provisions; monitoring of rooms and workers; and the plant radiation monitoring system which monitors the integrity of the containment barriers and initiates the actions needed to control the activity levels. The design measures adopted to optimise operator dose in normal plant operation are defined, and predictions are given of the expected level of collective doses. Radiation protection issues associated with systems and components to which access is required in long term post-accident situations are identified.Risks to workers due to accidents are considered in Chapter 15 (PSA) and Chapter 17 (ALARP) of this PCSR. Radiation doses to the public from normal operation are addressed separately in Chapter 11 of the Pre-Construction Environmental Report (PCER).- -- Sub-Chapter 12.0 - Radiation Protection RequirementsSub-chapter 12.0 aims to define the regulatory framework relating to radiation protection on which the design of the UK EPR™ is based. Legislative requirements, taken from international, European and national regulations, are given together with an overview of the regulatory bodies and the main radiation protection requirements, including the definition of radiation protection zoning and cleanliness / waste zoning.- file size 99 Kb Sub-Chapter 12.1 - Radiation Protection approachSub-chapter 12.1 outlines the basic principles applying to radiation protection, as specified by the International Commission on Radiological Protection (ICRP) and the radiation protection strategy implemented on the UK EPR™, including optimisation of the plant radioactive inventory and shielding provisions, and compliance with individual and collective dose limits.- file size 85 Kb Sub-Chapter 12.2 - Definition of radioactive sources in the primary circuitSub-chapter 12.2 provides details of the specific activities of the fission and corrosion products in the primary coolant, both in normal operation and shutdown conditions. Additionally, specific concentrations and deposited activity of important nuclides in the main loops are identified, based on measurements taken in French power stations. The system and chemistry measures, implemented on the UK EPR™ to control and optimise the primary circuit inventory levels, are also outlined.- file size 133 Kb Sub-Chapter 12.3 - Radiation protection measuresSub-chapter 12.3 provides details of the radiation protection measures implemented on the UK EPR™ including: radiation protection classification and zoning, and rules for the design (and installation) of equipment and rooms; shielding provisions, including materials and measures for the main buildings; design of the ventilation systems to minimise worker exposure to any contamination; monitoring, both of dose rates in rooms and of workers for potential contamination; and the plant radiation monitoring system which monitors the integrity of the containment barriers and initiates the actions needed to control the activity levels, including system functional flow diagrams.- file size 587 Kb Sub-Chapter 12.4 - Dose uptake optimisationSub-chapter 12.4 describes the design measures adopted in the UK EPR™ to optimise operator dose in normal plant operation, and gives predictions for the level of collective doses expected to be achieved. The dose received by individual workers is also considered with reference to the dose target for an individual worker that has been adopted as a safety design objective for the UK EPR™.- file size 278 Kb Sub-Chapter 12.5 - Post accident accessibilitySub-chapter 12.5 defines the systems and components to which access is required in long term post-accident situations, and specifies the associated accessibility conditions in terms of individual dose. Post-accident accessibility is defined for the systems used to maintain long-term cooling of the plant (over a year) and long term cooling of the fuel pool. The statutory requirements covering intervention personnel qualified to access rooms and operate systems are also outlined. This chapter only covers radiation protection issues; information on emergency arrangements is given in Sub-chapter 18.2 of the PCSR.- file size 94 Kb Chapter 13 - Hazards ProtectionChapter 13 presents the requirements, the design basis and the design verification methods for both the external and internal hazards taken into account in the UK EPR™ design. External hazards covered here include: earthquake, aircraft crash, external explosion, off-site fire, movement of toxic or corrosive gases, external flooding, extreme weather conditions (including wind generated missiles), and lightning and electromagnetic disturbances. Internal hazards covered include: pipework leaks and breaks; failure of tanks, pumps and valves; internal missiles; dropped loads; internal explosions; fire; and internal flooding.- -- Sub-Chapter 13.1 - External Hazards ProtectionSub-chapter 13.1 presents the requirements, the design basis and the design verification method for the external hazards taken into account in the UK EPR™ design: earthquake, aircraft crash, hazards associated with the industrial environment and transport routes (external explosion, off-site fire, movement of toxic or corrosive gases), external flooding, extreme weather conditions (snow and wind, wind generated missiles, low ambient temperatures, frazil and ice, high ambient temperatures, drought), and lightning and electromagnetic disturbances. The completeness of this list will need to be confirmed for each specific site, further site-specific assessments will be provided at that stage. For the external flooding hazard the design is fully site-dependent; in this case, and in the framework of the generic design assessment, the presentation is reduced to general principles. In addition the methodology and principles for development of the Hazard Fault Schedule are presented with reference to an example for external explosion.- file size 860 Kb Sub-Chapter 13.2 - Internal Hazards ProtectionSub-chapter 13.2 presents the requirements, the design basis and the design verification methods for the internal hazards taken into account in the UK EPR™ design: pipework leaks and breaks (including breaks of both high energy and moderate energy pipework); failure of tanks, pumps and valves; internal missiles; dropped loads; internal explosions; fire; and internal flooding. Internal hazard protection is provided to ensure that the safety related functions required to meet the safety objectives discussed in Chapter 3 of this PCSR are not unacceptably affected as the result of a hazard. In addition the methodology and principles for development of the Hazard Fault Schedule are presented with reference to an example for internal flooding.- file size 752 Kb Chapter 14 - Design Basis AnalysisChapter 14 presents the analyses of the plant response to postulated disturbances, malfunctions or failures of equipment, for the three plant condition categories (PCCs), PCC 2 to PCC-4 (design basis transients, incidents and accidents). These PCCs are events caused by either the failure of a single component or I&C function, or an operator error, or the loss of offsite power. The safety analysis of the PCC events defines the deterministic design of the safety systems. The purpose of the analyses is to demonstrate that the radiological consequences of abnormal events remain below the acceptance limits. The general principles, numerical targets, data, methods, and assumptions are given, together with details of the calculations performed and computer codes used, and a summary of the fault and protection schedule.- -- Sub-Chapter 14.0 - Assumptions and Requirements for the PCC Accident AnalysesSub-chapter 14.0 presents the assumptions and requirements for the Plant Condition Category (PCC) accident analyses. The standard reactor states, covering full power operation to cold shutdown, are defined, together with a list of the PCCs to be assessed. A set of "PCC accident analysis rules" are defined, which are used to demonstrate that the safety systems are designed in an appropriate manner. The rules include acceptance criteria, and the application of the single failure criterion.- file size 222 Kb Sub-Chapter 14.1 - Plant Characteristics taken into account in the Accident AnalysesSub-chapter 14.1 describes the generic plant characteristics that are applied to the accident analyses. Characteristics which are specific to a particular accident analysis are specified within the section describing that accident analysis. The plant characteristics assumed in the accident analyses cover: plant geometrical data, plant initial conditions, reactivity coefficients, residual (decay) heat, I&C signals related to reactor trip and safety systems operation, and safety system characteristics. The approach used in the accident analysis with regard to Departure from Nucleate Boiling (DNB) is also defined in this sub-chapter.- file size 468 Kb Sub-Chapter 14.2 - Analysis of the Passive Single FailureSub-chapter 14.2 presents the methodology and analysis of the consequences of passive single failures at the time of the initiating event for design basis (PCC-2 to PCC-4) events. The UK EPR™ is designed in compliance with the single failure criterion, which includes either an active single failure within the first 24 hours after the occurrence of an initiating event or a passive single failure at 24 hours after the event occurrence. This criterion is part of the assumptions made for the design basis fault studies of PCSR Sub-chapters 14.3 to 14.5. From a UK perspective, there is a need to review each PCC-2 to PCC-4 event, and to demonstrate that there are no passive single failures within 24 hours which could prevent the successful operation of the safety function that is provided to protect against the fault. The overall analysis shows that the UK EPR™ design is robust with respect to passive single failures.- file size 1,753 Kb Sub-Chapter 14.3 - Analyses of PCC-2 eventsSSub-chapter 14.3 describes the analysis of the PCC-2 events listed in Sub chapter 14.0 (i.e. design basis transients, events where the estimated frequency of occurrence is greater than 1 in a 100 reactor years). For each transient analysed, a description of the typical or assumed sequence of events is given together with identification of potential causes of the transient; any transient specific methods and assumptions are defined, as are any impact on the design of the claimed safety systems.- file size 1,543 Kb Sub-Chapter 14.4 - Analyses of the PCC-3 eventsSub-chapter 14.4 describes the analysis of the PCC-3 events listed in Sub chapter 14.0 (i.e. design basis incidents, or events where the estimated frequency of occurrence is between 1 in 100 and 1 in 10,000 reactor years). For each transient analysed, a description of the typical or assumed sequence of events is given together with identification of potential causes of the transient; any transient specific methods and assumptions are defined, as are any impact on the design of the claimed safety systems.- file size 2,494 Kb Sub-Chapter 14.5 - Analyses of the PCC-4 eventsSub-chapter 14.5 describes the analysis of the PCC-4 events listed in Sub chapter 14.0 (i.e. design basis accidents, or events where the estimated frequency of occurrence is between 1 in 10,000 and 1 in a million reactor years). For each transient analysed, a description of the typical or assumed sequence of events is given together with identification of potential causes of the transient; any transient specific methods and assumptions are defined, as are any impact on the design of the claimed safety systems.- file size 9,236 Kb Sub-Chapter 14.6 - Radiological consequences of design basis accidentsSub-chapter 14.6 presents the assessment of the radiological consequences of the various transients (PCC 2), incidents (PCC 3) and standard accidents (PCC 4) that are significant from the point of view of radiological releases. The aim of the assessment is to verify that the systems are properly designed and operated, such that the potential discharge of radioactive products outside the plant, which could have consequences for the public, remains within the prescribed limits. The general principles, numerical targets, data and assumptions are given, together with details of the calculations performed.- file size 626 Kb Sub-Chapter 14.7 - Fault and Protection ScheduleSub-chapter 14.7 provides a summary of the fault and protection schedule for the UK EPR™. The fault and protection schedule is a list of the initiating events that are considered in the EPR™ design due to their potentially unacceptable consequences, and it includes the initiating events and sequences, their frequencies, the classified safety systems that protect against them and the overall protection claims. The rationale and justification for the initiating events considered in the fault schedule is given, together with a justification of the comprehensiveness of the fault protection. This sub-chapter also describes the principles used to define the protection systems set-points and those applied to modify or deactivate settings during changes in the reactor state, and presents the diverse means available to reach a long-term safe shutdown state. In addition, information is provided on the ALARP discussion regarding the adequacy of the UK EPR™ design to protect the plant against faults.- file size 2,499 Kb Appendix 14A - Computer codes used in Chapter 14Appendix 14A contains brief descriptions of the computer codes used within the framework of the accident analysis of plant condition category events (Chapter 14), risk reduction analysis sequences (Sub-chapter 16.1) and overpressure protection analyses (Sub chapter 3.4). The computer codes used for each event are shown, including the use of coupled computer codes. Also included here are computer codes used in transient calculations taken from the generic EPR™ Basic Design Report 99, which are presented in Appendix 14B.- file size 605 Kb Appendix 14B - 4900MW Safety Analyses used in Chapter 14Appendix 14B contains sections from the generic EPR™ Basic Design Report 99 corresponding to accident analyses which are necessary to support some sections of the Pre-Construction Safety Report, but where all or some of the transient has not been specifically analysed for the UK EPR™. The cases reported include: small, intermediate and large break loss of coolant accidents from power operation; inadvertent closure of all/one main steam isolation valves; loss of non-emergency AC power to the plant auxiliaries; loss of condenser vacuum; full or partial loss of forced reactor coolant flow; uncontrolled rod cluster control assembly bank withdrawal from power operation; excessive increase in secondary steam flow, or steam system piping failure; and feedwater system pipe break; and steam generator tube rupture (1 or 2 tubes).- file size 476 Kb Appendix 14C - Analysis of single failure for main steam line breakAppendix 14C contains an analysis of a main steam line break during power operation with a break size greater than the area equivalent to a diameter of 50 mm (20 cm²), which is classified as a PCC-4 event. The objective of this analysis is to determine the most conservative single failure for this event in terms of departure from nucleate boiling. This single failure is used in the PCSR analysis of a main steam line break during power operation performed in section 2.1 of Sub-chapter 14.5- file size 1,709 Kb Chapter 15 - Probabilistic Safety AnalysisChapter 15 describes the probabilistic safety analysis performed in support of the UK EPR™ safety case. The safety objective of the reactor design is to ensure that the risk of potential release of radioactive products into the environment is reduced to as low as is reasonably practicable. In order to give confidence that the UK EPR™ design complies with this general safety objective, a probabilistic safety assessment (PSA) is performed, covering three levels of consequences. The Level 1 PSA analyses a number of initiating events together with total and partial failure of associated protection or mitigation measures. The Level 1 PSA failure states consider events that lead to core damage. Other, less onerous, endpoints are considered that do not result in a designated failure state but lead to potential releases. The Level 2 PSA takes the Level 1 PSA failure states, analyses the severe accident sequences and the corresponding containment response in those situations, and assigns a release category to sequences where the containment failure, leak or rupture is assumed. The consequence in terms of fission product releases (i.e., characteristics of activity release to the off-site environment) are also assessed for all sequence families. A probabilistic assessment of the radiological consequence of initiating events (Level 3 PSA) determines the off-site risk to the public.- -- Sub-Chapter 15.0 - Safety requirements and PSA objectivesSub-chapter 15.0 outlines the purpose and scope of the UK EPR™ probabilistic safety assessment (PSA), and presents the numerical targets that are used to evaluate and verify the UK EPR™ design. A number of Safety Design Objectives (SDOs) are adopted for the UK EPR™ and demonstration that these SDOs are met confirms that the EPR™ design complies with key ONR Safety Assessment Principles (SAPs). In addition to the SAPs numerical targets, the design also considers quantitative probabilistic safety objectives in accordance with the Technical Guidelines.- file size 129 Kb Sub-Chapter 15.1 - Level 1 PSASub-chapter 15.1 presents the level 1 probabilistic safety assessment (PSA), which allows identification of the sequences of internal events that can lead to core damage, and estimates the overall core damage frequency due to internal events in all reactor states. The scope of the probabilistic study for internal events is defined, together with the methodology used, including plant data assumptions, reliability data, treatment of common cause failures, I&C modelling, and human reliability assessment (both for pre-fault human errors and human errors during the course of the fault sequences) methodology. A discussion of the internal initiating events analysed is given, and the results of the Level 1 PSA analysis for internal initiating events are presented in terms of core damage frequency (per reactor per year).- file size 1,711 Kb Sub-Chapter 15.2 - PSA for internal and external hazardsSub-chapter 15.2 presents the results of an initial study to analyse the risk of core damage associated with internal and external hazards for the UK EPR™. The analysis of internal hazards covers all potential hazards and quantifies the risk associated with fire and flooding hazards. The analysis of external hazards is applied to a generic UK site (coastal or estuary) using the data presented in Chapter 2. The list of external hazards is reduced by a screening analysis that considers the expected impact on the plant safety; the expected frequency of core damage and the expected frequency of the external hazard. The results of the study are presented in terms of core damage frequency (per reactor per year), for both internal fire and flooding, and the 'screened in' external hazards. An assessment of the core damage frequency associated with the loss of ultimate heat sink (total loss of the water intakes) is also presented here.- file size 671 Kb Sub-Chapter 15.3 - PSA of accidents in the spent fuel poolSub-chapter 15.3 addresses the likelihood of damage to the fuel assemblies located in the spent fuel pool, or of boiling to the fuel building atmosphere. An assessment of the frequency of initiating events affecting the spent fuel pool cooling is presented, together with plant data assumptions and reliability data. The results of analysis, for both spent fuel pool non-draining and draining initiating events, are presented in terms of fuel damage frequency (per reactor per year).- file size 342 Kb Sub-Chapter 15.4 - Level 2 PSASub-chapter 15.4 presents the Level 2 probabilistic safety analysis (PSA), which assesses the probability of the severe accident phenomenon and the related response of the containment to potential loads. The Level 2 PSA assesses the characteristics of radiological releases from severe core damage sequences leading to containment failure, leakage or damage. The Level 2 PSA calculates the probability, composition, magnitude, and timing of fission product releases from the plant. The scope of the probabilistic study is described, together with a detailed description of the methodology used, including: the interface with the level 1 PSA (via the definition of core damage end states); the phenomenological evaluations performed for severe accidents; the development of containment event trees; the definition of fission product release categories; and the evaluation of the sensitivities and uncertainties. The results of the Level 2 PSA are presented in terms of the frequencies and release characteristics of the severe accident sequences, the binning of these sequences into release categories and the release risk associated with each release category.- file size 1,210 Kb Sub-Chapter 15.5 - Level 3 - PSA Assessment of off-site risk due to postulated accidentsSub-chapter 15.5 summarises the process followed to perform a probabilistic assessment of the radiological impact of the UK EPR™ design (Level 3 PSA). The results are presented in terms of: individual risk to any person off the site, presented as frequency versus consequences; and societal risk, that is the annual frequency of events which could potentially lead to more than 100 immediate or eventual fatalities in the wider UK population.- file size 295 Kb Sub-Chapter 15.6 - Seismic Margin AssessmentSub-chapter 15.6 presents the Seismic Margin Assessment (SMA) performed for the generic design assessment of the UK EPR™. The EPR™ design objective is that, following an earthquake, the safety functions needed to return the plant to a safe shutdown state should not be unacceptably affected. The seismic capacity of the UK EPR™ is assessed by a PSA-based SMA approach, which uses the PSA model to identify combinations of seismic equipment failures which could result in core damage, as well as combinations of seismic failures, random failures and human errors which contribute significantly to seismic risk. The seismic margin assessment demonstrates that the seismic capacity of the UK EPR™ is greater than 1.6 times the Design Basis Earthquake and that there are no cliff edge effects for seismic events which are more severe than the design basis event.- file size 924 Kb Sub-Chapter 15.7 - PSA Discussion and ConclusionsSub-chapter 15.7 presents the results and main findings of the Level 1 and Level 2 probabilistic safety analyses, together with the assessment of the radiological impact of the UK EPR™ design in terms of off-site risk to the public due to postulated accidents, and the use of the PSA in the design process. An iterative process to identify design improvements using PSA was implemented throughout the development of the EPR™ design; the main examples of design changes made resulting from PSA studies are presented here.. Several sensitivity analyses are also presented in this sub-chapter including the effects of preventative maintenance, the analysis of long term transients, sensitivity to initiating event frequencies; modelling of common cause failures; operator actions; reliability data and system design features. The probabilistic studies performed for the UK EPR™ design during the GDA give the assurance that the risk of potential release of radioactive products into the environment is reduced to as low as reasonably practicable.- file size 600 Kb Chapter 16 - Risk Reduction and Severe Accident AnalysesChapter 16 presents the risk reduction analysis and severe accident analyses, which supplement the design basis analysis presented in Chapter 14. In the risk reduction analysis, a set of design extension conditions (multiple-failure sequences) are identified and analysed in order to justify the design of additional safety features, which enable the prevention of severe accidents in these complex situations. The severe accident analysis considers very unlikely events which result in fuel failure, degradation of the structural integrity of the reactor core, and release of radioactive fission products into the reactor coolant system. It focuses on "practical elimination" and preservation of the long term integrity of the containment. Detailed analyses are presented, together with an assessment of radiological consequences to demonstrate that the release of radioactive products off-site remains within the prescribed limits. Sequences that are considered to be "practically eliminated", due to the implementation of specific design measures to reduce the risk of a large early off-site release as far as is reasonably practicable, are also presented, together with specific studies of highly unlikely events such as the 2A-LOCA. In addition, the functional diversity of the UK EPR™ design is addressed for all frequent faults and the Analysis of Extreme Beyond Design Basis Events carried out in Response to Fukushima is presented.- -- Sub-Chapter 16.1 - Risk Reduction Analysis (RRC-A)Sub-chapter 16.1 describes the risk reduction analysis, in which a set of design extension conditions (multiple-failure sequences) are analysed in order to justify the design of additional safety features ("RRC-A features"), which enable the prevention of severe accidents in these complex situations. The overall safety objectives, methodology, analysis principles, assumptions and data, are presented. An assessment of design extension condition sequences is detailed, including: failure of all control/shutdown rods on reactor trip demand (either mechanical failure or failure of the reactor trip signal); loss of off site power plus failure of all emergency diesel generators (station blackout); total loss of feedwater; total loss of cooling chain; small break loss of coolant accidents with additional safety injection failures; and boron dilution with additional operator error. An assessment of radiological consequences is presented, to demonstrate that, following these transients, the release of radioactive products outside the plant will have very low consequences for members of the public.- file size 2,844 Kb Sub-Chapter 16.2 - Severe Accident Analysis (RRC-B)Sub-chapter 16.2 describes the analysis of severe accidents, i.e. very unlikely events which result in fuel failure, degradation of the structural integrity of the reactor core, and release of radioactive fission products into the reactor coolant system. The overall safety objectives, and the approach to severe accident control, which focuses on preservation of the integrity of the containment under the associated loads, are presented. An assessment of core melt sequences is detailed, including identification and classification of the relevant scenarios from a phenomenological point of view. The areas assessed include: the primary depressurisation system, which prevents overpressurisation of the primary circuit; hydrogen control within the containment; stabilisation of core melt outside the reactor pressure vessel; and calculation of containment pressure and temperature in severe loss of coolant accidents. Radiological consequence calculations for severe core melt accidents are presented, to demonstrate that, taking account of the design provisions, the release of radioactive products outside the plant remains within the prescribed limits.- file size 8,418 Kb Sub-Chapter 16.3 - Practically eliminated situationsSub-chapter 16.3 presents sequences that are considered to be "practically eliminated". In the EPR™ context, "practical elimination" refers to the implementation of specific design measures to reduce the risk of a large early release of radioactive material to the environment to as low as reasonably practicable. To achieve practical elimination, each type of accident sequence that could lead to a large early release of radioactivity is examined and addressed by design measures. Demonstration of practical elimination of an accident sequence takes into account uncertainties due to the limited knowledge of physical phenomena involved in severe accident analysis. Conditions covered by specific treatment leading to their practical elimination are those which are liable to give rise to significant early releases; mainly high-pressure core meltdown sequences. The sequences considered in this approach are related to severe accidents and containment bypass situations.- file size 129 Kb Sub-Chapter 16.4 - Specific StudiesSub-chapter 16.4 presents the deterministic assessment of six specific faults: a double ended break of the main coolant line (2A-LOCA); a double-ended break of the main steam line outside the containment (2A-SLB); a steam generator tube rupture (1 tube) with main steam line break; a steam generator tube rupture (1 tube) with the affected steam generator main steam relief train stuck open; a multiple steam generator tube rupture (10 tubes in one steam generator at power) and spurious reactor protection system actuation. Assessment of the faults is presented, including description of the analysis methods, main assumptions and models used. In addition to these six specific faults the safety cases for the following types of events are also presented in this sub-chapter: heterogeneous boron dilution faults, faults linked with the Spent Fuel Pool (e.g.; connected pipework, cask loading pit faults or faults in fuel transfer tube) and faults associated with the failure of support systems (e.g.; ventilation, electrical and cooling chain).- file size 7,569 Kb Sub-Chapter 16.5 - Adequacy of the UK EPR™ design regarding functional diversitySub-chapter 16.5 provides a demonstration of the adequacy of the functional diversity of the UK EPR™ design. Functional diversity is addressed for all the frequent postulated initiating events (including front line systems and support systems), as these have higher requirements for mitigation. The safety functions are composed of plant level safety functions and the diversity is demonstrated within the plant level safety functions for all the frequent faults. Since some events are more bounding than others for a given plant level safety function, a comprehensive review of the transients is performed to select the limiting events before their examination by calculations. The sub-chapter is comprised of two parts: firstly, the methodology assessing the frequent events and the safety functions considered is presented and the selection of the most limiting events for each plant level safety function is demonstrated, and secondly, the limiting events are analysed by calculations or argumentations to demonstrate that the safety criteria are met.- file size 4,685 Kb Sub-Chapter 16.6 – Analysis of Extreme Beyond Design Basis Events Carried Out in Response to FukushimaSub-chapter 16.6 outlines the analysis of the robustness of the UK EPR™ design against extreme events, carried out in response to the Fukushima event that occurred in March 2011. The sub-chapter reviews the ability of the UK EPR™ to withstand severe external events, (notably major seismic and external flooding events of magnitudes beyond those considered in the current design). It also presents the time periods that would be available for operators to take action to prevent the release of radioactivity to the environment in hypothetical scenarios involving total loss of AC power and ultimate heat sink. Design modifications are described which are being implemented to enhance the robustness of UK EPR™ systems, structures and components against such extreme beyond design basis events.- file size 266 Kb Appendix 16A - Computer codes used in Chapter 16Appendix 16A briefly presents the computer codes used for severe accident (RRC B) analyses. The codes used are listed, alongside their current field of application and in which sub-section of Sub chapter 16.2 these codes are used. Each code is described, together with an outline of its validation basis, where required. Since the risk reduction (RRC A) analyses presented in Sub chapter 16.1 use the same codes as the design basis analyses these codes are described in Appendix 14A.- file size 230 Kb Appendix 16B - 4900 MW safety analyses used in Chapter 16Appendix 16B contains the generic EPR™ Basic Design Report 99 section corresponding to the analysis of risk reduction category A (RRC A) sequence "Small break loss of coolant accident without low head safety injection system". This sequence has not yet been specifically analysed for the UK EPR™ in the framework of the PCSR studies. The analysis presented in this appendix should be read in conjunction with section 3.8 of Sub chapter 16.1.- file size 648 Kb Chapter 17 - Compliance with ALARP PrincipleChapter 17 presents information to demonstrate that the UK EPR™ design meets the ALARP requirement. UK Health and Safety Legislation places a duty on all companies to conduct their operations such that the risk posed to their workers and members of the public is as low as reasonably practicable (ALARP). The required demonstration of relevant good practice includes reviews of: the experience of EPR™ designers; the design process, including the R&D underpinning the design; the design codes used in EPR™ design; and the use of operational feedback from French and German plants in optimising EPR™ design. A review of the design options considered in the EPR™ design optimisation process is presented, together with the rationale for the options chosen in terms of the fundamental requirement to minimise risks to workers and the public from EPR™ operation. The results of the probabilistic safety analysis demonstrating that the risk to the public due to accidents is 'broadly acceptable' are summarised. An assessment of additional modification options required by international regulators, and modifications implemented in the Sizewell B PWR, concludes that these are not considered to be reasonably practicable according to the quantitative ALARP assessment methodology.- -- Sub-Chapter 17.1 - Explanation of ALARP RequirementSub-chapter 17.1 provides an overview of the requirements of the ALARP principle. UK Health and Safety Legislation places a duty on all companies to conduct their operations such that the risk posed to their workers and members of the public is as low as reasonably practicable (ALARP). In the context of a nuclear power plant, this duty requires that all measures are taken during design and operation to minimise radiation doses to workers or members of the public, provided the cost of such measures is not disproportionately large compared with the benefits achieved. For generic design acceptance, it is proposed that supporting safety submissions should include the following information: a demonstration of relevant good practice; a review of design options considered in the design process; presentation of results of a probabilistic risk assessment; and arguments that no further reasonable practicable improvements could be implemented, and therefore the risk has been reduced to ALARP.- file size 88 Kb Sub-Chapter 17.2 - Demonstration of Relevant Good Practice in EPR DesignSub-chapter 17.2 provides the required demonstration of relevant good practice and includes: a review of both the experience of EPR™ designers and the review and assessment process applied to the design, including a summary of the R&D effort underpinning the EPR™ design; a review of the design codes used in EPR™ design, including reference to international and national codes; a summary of the use of operational feedback, in particular from French and German plants, in optimising EPR™ design; and a discussion of a comparison of the UK EPR™ design against the ONR safety assessment principles, to confirm that all the key nuclear safety requirements embodied in the SAPs are met by the UK EPR™ design.- file size 387 Kb Sub-Chapter 17.3 - EPR Design OptioneeringSub-chapter 17.3 describes the optioneering process carried out in France and Germany between 1987 and 2006 to develop the EPR™ design, and the design review carried out by independent safety experts on behalf of the French and German safety authorities. The outcome of the design optioneering process in terms of the principal design options that were selected and rejected to achieve a balanced design, in terms of the fundamental requirement to minimise risks to workers and the public, while achieving practical constructability and a cost-effective design, is presented. The rationale for the evolution of the design, and the improvements from predecessor designs, are explained along with the reasons why certain features were selected and others rejected. It is demonstrated that the process of design optimisation adopted to minimise risk due to accidents and to optimise operator dose in normal plant operation and accidents is closely analogous to the formal UK approach of ALARP.- file size 191 Kb Sub-Chapter 17.4 - Review of PSA Results: Comparison with Numerical Risk TargetsSub-chapter 17.4 summarises the results of the probabilistic safety analysis (PSA) for the UK EPR™ (presented in Sub-chapter 15.5 of the PCSR), which demonstrates that the risk to the public due to accidents meets the ONR targets for 'Broadly Acceptable' risk. It is confirmed that UK EPR™ safety design objectives for risk (presented in Sub-chapter 3.1 of the PCSR) are achieved, and therefore that the risk from UK EPR™ operation is likely to be in the broadly acceptable region.- file size 137 Kb Sub-Chapter 17.5 - Review of Possible Design Modifications to Confirm Design meets ALARP PrincipleUK regulatory guidance on the application of As Low As Reasonably Practicable (ALARP) for new civil reactors recommends that there should be a clear conclusion that no further reasonable practicable improvements could be implemented in the reactor design, and that therefore the risk has been reduced to ALARP. Sub-chapter 17.5 provides an ALARP assessment methodology, and an ALARP assessment of several UK EPR™ design alternatives, based on modifications required by international regulators in their assessment of the EPR™, or modifications implemented in the Sizewell B PWR. None of the modifications considered is indicated as reasonably practicable according to the quantitative ALARP assessment methodology.- file size 302 Kb Sub-Chapter 17.6 - Conclusions of EPR ALARP AssessmentSub-chapter 17.6 provides the conclusions of the UK EPR™ ALARP assessment, and concludes that the design of the UK EPR™ complies with the overall requirements of the ALARP principle, which is a key safety design objective for the UK EPR™.- file size 67 Kb Chapter 18 - Human Factors Operational AspectsChapter 18 describes how Human Factors and Operational Aspects support the overall nuclear safety demonstration for the plant. It describes how Human Factors (HF) principles, methods and standards are integrated into the plant design and operating concepts. Humans play an essential role in nuclear power plant operation. They play a key role in the safe operation of the plant, particularly in managing unexpected situations, reducing the potential for plant degradation in normal operation, and in the testing and maintenance of systems. However, humans may be fallible and it is therefore important to support both positive human contributions to overall safety and to minimise the risks associated with human error.- -- Sub-Chapter 18.1 - Human FactorsSub-chapter 18.1 describes how Human Factors (HF) principles, methods and standards are integrated into the UK EPR™ design and operating concepts. It explains the objectives and scope of HF integration including the use of operating experience from previous plants to improve the design from a HF viewpoint. It presents a set of safety arguments as to how human performance will contribute to the achievement of nuclear safety and provides evidence to support those arguments. The Human Based Safety Claims that are part of the overall Nuclear Safety demonstration for the plant are identified and the requisite evidence (e.g. task analyses) to support these claims is presented. The requirements of the human machine interface design are described. The assurance process that underlies the achievement of adequate human reliability to support the overall nuclear safety demonstration is described.- file size 1,205 Kb Sub-Chapter 18.2 - Normal OperationThe PCSR contains assumptions and requirements necessary to ensure the results claimed in the overall safety case. Operating documents will need to be defined in order to ensure that the plant is operated consistently with these assumptions and requirements. Sub chapter 18.2 outlines the methods that will provide design limits and conditions for the UK EPR™. It details the arrangements for moving the PCSR to an operating regime which will ensure that the requirements and assumptions contained in the PCSR are captured in operating documents. During normal operation the plant must be operated to manage normal scheduled operating transients and specific operations involving unplanned events, and in a manner consistent with the safety case assumptions. This is achieved using the principles and procedures for normal operation described here, and by establishing operating rules for normal operation.- file size 649 Kb Sub-Chapter 18.3 - Abnormal OperationSub-chapter 18.3 describes the operating principles to be applied to the UK EPR™ plant outside of normal operation. Emergency operation and severe accident conditions are defined. The operating principles for both of these conditions are defined, as far as they can influence the plant safety assessment. The emergency operating principles address plant operation following an accident, when operator actions are needed, and are based on a plant State Oriented Approach. Operating principles concerning severe accidents are outlined, covering the scenarios taken into account in the safety analysis (described in Sub chapter 16.2). Additionally the principles that would be applied in developing an Emergency Plan for the UK EPR™ are described.- file size 256 Kb Chapter 19 - CommissioningChapter 19 provides a brief outline of the commissioning requirements and programme. The commissioning stage and the associated test programme are required to demonstrate that the plant, as built, is suitable for safe operation and that the design intent, as claimed in the safety case, has been achieved. A comprehensive commissioning programme will be prepared for implementation on site, and the commissioning tests will be designed to demonstrate, to the extent practicable, that the plant components and systems are capable of safe and reliable operation in accordance with their design specification, their performance objectives and their safety requirements.- -- Sub-Chapter 19.0 - Commissioning Safety RequirementsSub-chapter 19.0 provides a brief outline of the commissioning safety, programme and organisational requirements, together with the associated regulatory framework.- file size 77 Kb Sub-Chapter 19.1 - Plant Commissioning ProgrammeSub-chapter 19.1 provides an overview of the objectives, principles and scope of the plant commissioning programme, which is carried out in three main phases. Phases I and II are part of the pre-operational test programme, which includes: the preliminary tests and controls, and the first start-up of equipment and functions, without any interaction between the primary and the secondary-side systems; followed by the cold and hot functional tests of the reactor coolant and secondary-side systems before fuel loading. Phase III of the commissioning programme corresponds to the initial start-up test programme, which includes core loading, cold and hot pre-critical tests and actual start-up, including a demonstration run, up to the start of commercial operation.- file size 103 Kb Chapter 20 - Design Principles related to DecommissioningChapter 20 provides an overview of the decommissioning and dismantling process, which is the final stage of operation of the nuclear power plant. After presenting the design principles associated with decommissioning, the chapter explains how these principles have been implemented for the EPR™ reactor. These measures facilitate the dismantling of the reactor, limit the dose uptake for the corresponding operations and limit the quantity and activity of the nuclear waste produced, particularly in comparison to existing PWRs, which did not consider decommissioning at the design stage.- -- Sub-Chapter 20.1 - General Decommissioning Principles - RegulationsSub-chapter 20.1 provides an overview of the dismantling process which results in the regulatory delicensing of the site, the approach to reduction of risk to the public during dismantling, the documentation required, and outlines the UK EPR™ design principles which ensure that both the radioactive dose received by workers, and the amount of radioactive waste and hazardous material produced are reduced to as low as reasonably practicable. The baseline principles and objectives that should be adopted during decommissioning, to enable adequate management of the decommissioning process and preparation of plans and proposals for decommissioning, are outlined, together with the need for records of information pertinent to decommissioning.- file size 72 Kb Sub-Chapter 20.2 - Decommissioning - Implementation for the EPR™Sub-chapter 20.2 describes how the principles and regulations discussed in PCSR Sub-chapter 20.1 have been followed at the early UK EPR™ design stage to ensure and facilitate future decommissioning in a safe and environmentally acceptable way. It describes the measures adopted at the EPR™ design stage to minimise the volume of radioactive structures, reduce the toxicity of the waste, lower the activity level of irradiated components, restrict the spread of contamination and permit easier decontamination, facilitate the access of personnel and machines and the removal of waste, and ensure the collection of building and operating data needed to prepare for decommissioning. An estimate of the volumes of low and intermediate level waste is also given, together with a discussion of the disposability thereof. Finally, the baseline decommissioning strategy currently envisaged is presented.- file size 122 Kb Chapter 21 - Quality and Project ManagementChapter 21 outlines the organisation and management system arrangements established for the generic design assessment (GDA) process of the UK EPR™. A detailed description of the whole UK EPR™ project organisation requires that the organisational structure, interfaces, and responsibilities be clearly defined for the different phases, from design licensing through to plant operation. For the present submission, the project has focussed on the organisation required for the generic design assessment process. The post-GDA organisation will depend on the partnerships established for the future project phases, in particular the plant owner.- -- Sub-Chapter 21.1 - Project OrganisationSub-chapter 21.1 outlines the organisation established for the project management of the generic design assessment (GDA) process, and the main organisational arrangements required to control the design and construction of a new plant in accordance with the applicable standards and regulations.- file size 84 Kb Sub-Chapter 21.2 - Management SystemSub-chapter 21.2 outlines the quality programme established for the UK EPR™ generic design assessment (GDA) project, and the management systems of the co-applicants (EDF and AREVA) and main subcontractor (AMEC), including documentation, control, records, design review and change management processes, and audits.- file size 122 Kb |
PCERThe Pre-Construction Environmental Report (PCER) provides the information requested by the Environment Agency in its guidance Process and Information (P&I) Document, to allow their detailed assessment of the UK EPR™ environmental impact.Contents The Pre-Construction Environmental Report is divided into 12 chapters. The full list of the chapters and associated sub-chapters is available in the PCER Table of Contents. - -- Chapter 1 - Introduction and General DescriptionChapter 1 describes the overall purpose and scope of the Pre-Construction Environmental Report (PCER), within the scope of the Generic Design Assessment (GDA) process. It gives a general overview of the UK EPR™ buildings, structures and main power plant systems. A comparison is provided of the data for the main reactor systems against the equivalent data for the latest generations of reactors built in France (N4 Reactors) and Germany (KONVOI Reactors). Additionally, there is an overview of the structure of the UK regulations, the associated regulatory framework and key relevant UK regulations, together with an outline of the design and safety assessment process for the EPR™ within France, Finland and the USA, and an overview of comparisons of the EPR™ design against international safety standards.- -- Sub-Chapter 1.1 - IntroductionSub-chapter 1.1 describes the overall purpose and scope of the Pre-Construction Environmental Report (PCER), within the scope of the Generic Design Assessment (GDA) process. It also describes the structure and layout of the PCER, and provides an outline description of the contents, and a road map of the PCER to the requirements of the EA guidance Process and Information (P&I) Document.- file size 129 Kb Sub-Chapter 1.2 – General Description of the UnitSub-chapter 1.2 provides a general overview of the UK EPR™. It gives a brief description of the buildings and structures of UK EPR™ unit, and describes the criteria for location and installation of these. It also provides a description of the main power plant systems, together with a brief overview of the general operating principles for the UK EPR™.- file size 2,659 Kb Sub-Chapter 1.3 - Comparison with reactors of similar designSub-chapter 1.3 consists of a comparison table of the data for the main reactor systems of the UK EPR™ against the equivalent data for the latest generations of reactors built in France (N4 Reactors) and Germany (KONVOI Reactors), from which the EPR™ design is derived.- file size 139 Kb Sub-Chapter 1.4 - Compliance with regulationsSub-chapter 1.4 provides an overview of the UK regulations to be complied with by the UK EPR™ design. An overview of the structure of the UK regulatory regulations and the associated regulatory framework is provided, followed by an outline of the key relevant UK regulations. The list is not exhaustive but gives a general overview of the basic regulatory requirements.- file size 164 Kb Sub-Chapter 1.5 - Safety assessment and international practiceSub-chapter 1.5 provides an overview of the design and safety assessment process for the EPR™ within France, Finland and the USA, together with an overview of comparisons of the EPR™ design against international safety standards (the Western European Nuclear Regulators Association (WENRA) reference levels, International Atomic Energy Agency (IAEA) Safety Standards, and the European Utility Requirements for LWR nuclear power plants).- file size 137 Kb Chapter 2 - Quality and Project ManagementChapter 2 outlines the organisation and quality assurance arrangements established for the project management of the generic design assessment (GDA) process. A detailed description of the whole UK EPR™ project organisation requires that the organisational structure, interfaces, and responsibilities be clearly defined for the different phases, from design licensing through to plant operation. For the present submission, the project has focussed on the organisation required for the generic design assessment process. The post-GDA organisation will depend on the partnerships established for the future project phases, in particular the plant owner.- -- ContentsChapter 2, Quality and Project Management, comprises 2 sub-chapters. Sub-chapter 2.1 outlines the organisation established for the project management of the generic design assessment (GDA) process, and the main organisational arrangements required to control the design and construction of a new plant in accordance with the applicable standards and regulations. Sub-chapter 2.2 outlines the quality programme established for the UK EPR™ generic design assessment (GDA) project, and the management systems of the co-applicants (EDF and AREVA), and main subcontractor (AMEC), including documentation, control, records, design review and change management processes, and audits.- file size 173 Kb Chapter 3 - Aspects having a Bearing on the Environment during Operation PhaseChapter 3 provides an overview of the interfaces of the UK EPR™ with the environment and outlines the operational phases that will potentially have an impact on the environment because they either use natural resources or generate wastes and discharges. This chapter describes the needs of the EPR™ design in terms of natural resources; how the wastes and discharges arise; and what they are, related to both the terrestrial and aquatic environments. The Control of Major Accidents Hazards linked to hazardous substances stored and / or manipulated on the site is addressed here. Elements related to the Pollution Prevention & Control (PPC) application, now incorporated into the Environmental Permitting Regulations, for standby diesel generators are given in this chapter.- -- ContentsChapter 3, Aspects having a Bearing on the Environment during Operation Phase, comprises 4 sub-chapters. Sub-chapter 3.1 provides a general introduction and gives an overview of the interfaces of the UK EPR™ with the environment. Sub-chapter 3.2 outlines the different phases and activities that will take place during the UK EPR™ operational phase, including outages and maintenance, emphasising those that will potentially have an impact on the environment because they either use natural resources or generate wastes and discharges. Sub-chapter 3.3 identifies the aspects of the operational phase, which have a potential impact on the terrestrial environment for an operating UK EPR™. The Control of Major Accidents Hazards linked to hazardous substances stored and / or manipulated on the site is addressed here. Sub-chapter 3.4 identifies the aspects of the operational phase, in terms of both requirements for natural resources and outputs, which have a potential impact on the aquatic environment, for a coastal or estuarine operating UK EPR™ Unit.- file size 421 Kb Chapter 4 - Aspects having a Bearing on the Environment during Construction PhaseChapter 4 describes the various activities that will take place during the UK EPR™ construction, from site preparation through to commissioning, and focuses on those aspects (i.e. needs and outputs) which have a potential impact on the terrestrial and aquatic environments. In particular water abstraction, discharge of contaminants and the non-radioactive waste strategy are highlighted.- -- ContentsChapter 4, Aspects having a Bearing on the Environment during Construction Phase, comprises 3 sub-chapters. Sub-chapter 4.1 provides a brief introduction. Sub chapter 4.2 outlines the activities that take place during the EPR™ construction phase, including site preparation activities, preliminary works, the EPR™ plant construction, and commissioning. Sub-chapter 4.3 focuses on the aspects of the construction phase that have a potential impact on the terrestrial and aquatic environments; i.e. the interface of the plant construction, erection and commissioning with the environment in terms of requirements for natural resources and outputs that might impact the existing environment. In particular water abstraction, discharge of contaminants and the non-radioactive waste strategy are highlighted.- file size 485 Kb Chapter 5 - Design principles related to decommissioningChapter 5 provides an overview of the decommissioning and dismantling process, which results in the regulatory delicensing of the site. After presenting the design principles associated with decommissioning, the chapter explains how these principles have been implemented for the EPR™ reactor. These measures facilitate the dismantling of the reactor, limit the dose uptake for the corresponding operations and limit the quantity and activity of the nuclear waste produced, particularly in comparison to existing PWRs, which did not consider decommissioning at the design stage.- -- ContentsChapter 5, Design Principles related to decommissioning, comprises 2 sub-chapters. Sub-chapter 5.1 provides an overview of the dismantling process which results in the regulatory delicensing of the site, the approach to reduction of risk to the public during dismantling, the documentation required, and outlines the UK EPR™ design principles which ensure that both radioactive doses and radioactive waste are reduced to as low as reasonably practicable. The baseline principles and objectives that should be adopted during decommissioning are outlined, together with the need for records of information pertinent to decommissioning. Sub-chapter 5.2 describes how the principles and regulations have been followed at the early UK EPR™ design stage to ensure future decommissioning in a safe and environmentally acceptable way. It describes the measures adopted at the EPR™ design stage to facilitate decommissioning, such as minimisation of radioactive waste and contamination. An estimate of the volumes of low and intermediate level waste is also given, together with a discussion of the disposability thereof. Finally, the baseline decommissioning strategy currently envisaged is presented.- file size 170 Kb Chapter 6 - Discharges and Waste - Chemical and RadiologicalChapter 6 provides details of the waste and effluent management systems, which include the liquid and gaseous waste treatment systems, and the solid waste treatment systems. Also given in this chapter are: details of the activity levels of the main primary system for normal operation; details of the effluent management processes (collection, processing, storage and disposal) including the sources, nature, treatment and storage of liquid and gaseous effluent discharges; information regarding the UK EPR™ radioactive solid waste and spent nuclear fuel strategy, including waste treatment and interim storage facilities and disposability; a quantitative estimate of the effluents discharged and the waste produced by the operation of the UK EPR™ reactor, including spent fuel and solid waste arisings, liquid and gaseous radioactive effluent discharges, and chemical effluent discharges; and a discussion of the UK EPR™ designs for interim storage facilities, for the radioactive Intermediate Level Waste (ILW) operational waste, and the spent fuel.- -- Sub-Chapter 6.0 – Safety RequirementsSub-chapter 6.0 describes the safety functions, and functional and design requirements, of the waste management systems, which include the liquid and gaseous waste processing systems, and the solid waste treatment system.- file size 64 Kb Sub-Chapter 6.1 – Sources of radioactive materialsSub-chapter 6.1 provides details of the activity levels of the main primary system when the reactor is in normal operation (both steady-state power operation and shutdown transient). The concentrations of fission, corrosion and activation products provided are based on realistic operational data, therefore they may be used in a generic way as interface data for effluent discharges, radiological protection and assessment of realistic dose uptake, and radiological protection requirements. Source terms used for the design and sizing of the EPR™ structures, rooms, systems, shielding and the effluent treatment systems are also defined.- file size 119 Kb Sub-Chapter 6.2 – Details of the Effluent Management ProcessSub-chapter 6.2 details the effluent management process (collection, processing, storage and disposal) for liquid and gaseous radioactive discharges and solid radioactive waste. It identifies the sources, nature, treatment and storage of liquid and gaseous effluent discharges. It discusses the chemicals associated with liquid effluent that could be discharged into the marine environment, including the purpose of each substance, the main circuits involved, the path from injection to discharge, and the proposed maximum amounts discharged. Information is also given regarding the UK EPR™ radioactive waste and spent nuclear fuel strategy, including waste treatment and interim storage and disposability. The strategy takes into account the UK EPR™ solid radioactive waste and spent fuel characteristics, and addresses the government and NDA policies for the assessment of disposability of radioactive waste. It includes the description of processes and interim storage facilities for managing the waste and spent nuclear fuel.- file size 420 Kb Sub-Chapter 6.3 – Outputs for the operating installationSub-chapter 6.3 presents the quantitative estimate of the effluents discharged and the waste produced by the operation of the UK EPR™ reactor. The estimates cover: spent fuel arisings, solid waste arisings, liquid and gaseous radioactive effluent discharges, and chemical effluent discharges, whether associated with liquid radioactive effluent or arising from operation of non-nuclear parts of the plant. The expected performance discharges are derived from recent operational feedback, taking credit for design features and expected quantifiable gains. Maximum additional annual discharges are provided, together with estimates of monthly discharges of gaseous and liquid radioactive waste.- file size 1,645 Kb Sub-Chapter 6.4 – Effluent and waste treatment systems design architectureSub-chapter 6.4 describes the UK EPR™ effluent and waste treatment systems for liquid, gaseous and solid waste. The role of the effluent treatment systems, in normal operation, involves: collecting effluent produced in the unit and site facilities, which may come from draining, bleeding, venting or leaks from systems that are normally or accidentally contaminated; treating, if necessary, the effluents, i.e. reducing the level of contamination in order to meet regulatory discharge limits; and discharging effluents outside the site following monitoring and accounting of the levels of activity discharged under regulatory conditions. In some accident situations which could lead to significant transfer of contamination to the buildings adjacent to the reactor building, the effluent treatment systems operate to ensure any contamination is contained, in order to limit release of radioactivity to the environment. For each system, the safety functions, functional and design requirements, the design criteria, system design description, system parameters and operating conditions, safety assessment, and specific testing arrangements, are provided as appropriate.- file size 3,473 Kb Sub-Chapter 6.5 – Interim storage facilities and disposability for UK EPR™Sub-chapter 6.5 discusses the UK EPR™ designs for interim storage facilities, for the radioactive Intermediate Level Waste (ILW) operational waste, and the spent fuel, produced during the UK EPR™ 60 years of operation. The need for plans to demonstrate when facilities will be developed and constructed is also addressed. Additionally, the current status of the discussions with the NDA, with respect to the disposability of UK EPR™ solid radioactive waste, and spent fuel arising during the operational period, and the discussions with LLW Repository Ltd with regard to Low Level Waste disposability, are described. This sub-chapter and supporting documents provide a basis for a potential UK EPR™ operator to develop radioactive waste management cases for both ILW and spent fuel.- file size 286 Kb Chapter 7 - Measures for Monitoring the DischargesChapter 7 presents the approach to the monitoring of radioactive releases and chemical discharges both to water and to atmosphere from the UK EPR™. It identifies the methods, including sampling arrangements, techniques and systems used to measure and assess discharges. In particular, it specifies methods used to determine discharges of groups of radionuclides and substances that are subjected to an authorisation and provides information on the efficiency of abatement plant.- -- ContentsChapter 7, Measures for Monitoring Discharges, comprises 4 sub-chapters. Sub-chapter 7.1 provides a general introduction and gives an overview of the discharge control and environmental monitoring programme to be implemented. Sub-chapter 7.2 describes the UK legislation covering both radioactive and non-radioactive discharges and the associated monitoring principles. Sub-chapter 7.3 describes the approach to the monitoring of radioactive releases both to water and to atmosphere from the UK EPR™. It identifies the methods, including sampling arrangements, techniques and systems/devices used to measure and assess discharges. In particular, it specifies methods used to determine discharges of groups of radionuclides and substances that are subjected to a permit and provides information on the efficiency of abatement plant. Sub-chapter 7.4 describes the approach to the monitoring of chemical discharges both to water and to atmosphere from the UK EPR™. It describes the types of discharges expected, and identifies the methods used to measure and assess discharges.- file size 380 Kb Chapter 8 - Best Available TechniquesChapter 8 provides an evaluation of environmental options considered, and demonstrates that the Best Available Techniques (BAT) have been used to minimise the production, discharge and disposal of waste. The demonstration is based on the following successive points: explanation of the design process of the EPR™ (historical background, environmental approach, description of optimisation measures assessed with regard to BAT management factors from the OECD decision-aiding strategy for effluent discharges and solid waste options; application of BAT standards and good practices for the design of the EPR™; implementation of BAT within monitoring procedures.- -- ContentsChapter 8, Best Available Techniques (BAT), comprises 4 sub-chapters. Sub-chapter 8.1 outlines the scope and objectives, and gives an overview of the concept of BAT, together with an outline of the OECD decision-aiding strategy for effluent discharges and solid waste options based on optimisation factors which demonstrate the application of BAT. Sub chapter 8.2 outlines the EPR™ design process, demonstrating how optimisation of effluent management has been an integral part of the design process from the earliest design stage. The main optimisation measures are assessed in the context of the OECD optimisation factors. The expected environmental performance of the UK EPR™ is evaluated, and compared with existing facilities. Sub-chapter 8.3 describes the application of BAT standards and good practices within the UK EPR™ design. Information is also provided on compliance of the effluent treatment systems with relevant safety classifications; minimisation of unplanned emissions to groundwater; and management and disposal of non-radioactive liquid waste streams over the lifetime of the facility. Sub-chapter 8.4 presents a justification that the monitoring methods implemented for the UK EPR™ radioactive discharges (including the discharge and environmental monitoring programme) represent the best available techniques.- file size 664 Kb Chapter 9 - Principles and Methods used for Environmental Approach at the Design StageChapter 9 provides an introduction to the generic impact study being undertaken for the UK EPR™ and presented in Chapters 10, 11 and 12 of the PCER. It presents the general regulatory framework, the scope of the impact study and a brief description of the methodologies used to assess the potential impacts of a UK EPR™. The impact study is divided into three parts: a description of the site environmental characteristics, the radiological impacts, and the non-radiological impacts on human health and non human species. The present impact study applies to a UK EPR™ unit located on a coastal or estuarine site, which is judged as being representative of the development of a potential nuclear power station in the UK.- -- ContentsChapter 9, Principles and Methods used for Environmental Approach at the Design Stage, comprises 4 sub-chapters. Sub-chapter 9.1 provides a general introduction to Chapter 9 and gives an overview of the concept and scope of the impact assessment carried out for the UK EPR™ in the context of GDA. Sub-chapter 9.2 describes the general UK legislative framework covering the requirement for, and scope and content of, an environmental impact assessment. Additionally, the requirement for a radiological dose assessment is outlined, together with the associated dose limits. Sub-chapter 9.3 outlines the radiological impact assessment carried out for the UK EPR™. It summarises the methodology covering the radiological impacts on human health. Sub-chapter 9.4 outlines the non-radiological (environmental) impact assessment to be carried out for the UK EPR™. It summarises the methodology covering impact on the impacts on the terrestrial and aquatic ecosystems and potential ecological impacts.- file size 180 Kb Chapter 10 - Site Environmental CharacteristicsChapter 10 presents the site data which need to be considered when assessing the environmental impacts of the UK EPR™ design (e.g. terrestrial and aquatic characteristics, habitats data). The chapter outlines the information and methodologies to be used for the baseline studies during the site specific phase. The chapter also includes a set of site parameters appropriate for the UK, i.e. representative of a range of potential coastal and estuarine locations, in order to carry out environmental impact studies for Generic Design Assessment (GDA). Following GDA it will be necessary to further update this Chapter to include site specific details and to utilise available site-specific data.- -- ContentsChapter 10, Site Environmental Characteristics, comprises 6 sub-chapters. It should be noted that most of the sub chapters are currently largely templates which will be completed for a specific site. Sub-chapter 10.1 provides a short introduction and outlines where information can be found within Chapter 10. Sub-chapters 10.2 and 10.3 outline the information required to characterise the terrestrial and aquatic ecosystems beyond the influence of the nuclear power station. Sub-chapter 10.4 outlines the information required to facilitate the identification of designated sites, sensitive habitats, and protected species. Sub-chapter 10.5 outlines the information required to characterise the human reference state, including regional demographics, agriculture, food intake, recreational fishing and leisure activities, noise, access and traffic. Sub-chapter 10.6 outlines the information required to characterise the radiological reference state, including: radio-ecological studies to provide the baseline occurrence of natural and artificial radionuclides in the terrestrial and marine environments.- file size 256 Kb Chapter 11 - Radiological Impact AssessmentChapter 11 describes the methodology used to carry out dose assessments for the public and for non human species. The doses from gaseous and liquid radioactive discharges at the proposed limits for the UK EPR™ station and from direct radiation are considered. A generic annual dose to the public is calculated using a set of site parameters representative of a potential UK EPR™ site and the conservative assumptions of the initial radiological assessment methodology recommended by the Environment Agency. Calculations of potential short term doses from gaseous discharges, together with doses associated with the potential build up of radionuclides in the local environment over the station lifetime, and collective doses to wider populations from atmospheric and liquid discharges are also presented, together with calculated doses to both terrestrial and marine organisms.- -- ContentsChapter 11, Radiological Impact Assessment, comprises 5 sub-chapters. Sub-chapter 11.1 describes the methodology and assumptions used for an initial conservative radiological assessment of the dose from radioactive waste discharges to the environment, and identifies those discharges requiring more detailed assessment . A generic annual dose to the public is calculated. Calculations of potential short term doses from gaseous discharges, doses associated with the potential build up of radionuclides in the local environment over the station lifetime, and collective doses to wider populations from atmospheric and liquid discharges are also presented. Sub-chapter 11.2 describes the methodology used to assess the impact from radioactive discharges on non-human species and presents calculated doses to both terrestrial animal and plant, and marine organisms. Sub-chapter 11.3 outlines the uncertainties associated with the methods used for the assessment of radiological impacts on human health and non human species. Sub-chapter 11.4 outlines the methods and options available for radiological monitoring and sampling, covering both the terrestrial and marine environments. Sub-chapter 11.5 summarises the results of the radiological assessments presented in this chapter.- file size 837 Kb Chapter 12 - Non-radiological Impact AssessmentChapter 12 describes the methodology used to assess the non radiological impacts due to a UK EPR™. It deals in particular with the impact of water abstractions, thermal and chemical substances discharges. Noise impact, impacts on landscape and socio-economics are also considered, together with the requirements for a site specific environmental protection and monitoring programme. The identification of potential impacts focuses on the construction and operational phases of the EPR™. The majority of the potential impacts will need to be assessed in detail during the site specific phase. However, initial assessments of the impact of liquid chemical discharges associated with both start up testing and operation are presented, together with an assessment of potential impacts on designated sites, and UK sensitive habitats and protected species, for a representative site.- -- ContentsChapter 12, Non radiological Impact Assessment, comprises 7 sub-chapters. Sub-chapters 12.1 and 12.2 identify the potential impacts of the UK EPR™ on the terrestrial and aquatic ecosystems, including the impact of chemical discharges. Sub-chapter 12.3 presents an ecological impact assessment for a representative site. Sub-chapter 12.4 identifies potential socio-economic impacts of the UK EPR™ and outlines the overall approach to stakeholder engagement. Sub-chapter 12.5 outlines the potential uncertainties associated with the methods used for the assessment of non-radiological impacts.Sub-chapter 12.6 outlines the objectives of the site specific environmental protection and monitoring programme which will be established. Sub-chapter 12.7 summarises the initial results of the generic non radiological impact assessments, and outlines the further work that will be required.- file size 816 Kb |
Introduction to the Safety, Security and Environmental Report (SSER)The Safety, Security and Environmental Report (SSER) for the proposed UK EPR™ design addresses the requirements of the ONR and EA Generic Design Assessment process. It provides design, safety and environmental information to support the detailed assessment of the UK EPR design against UK safety and environmental standards. The SSER comprises two documents: the Pre-Construction Safety Report (PCSR) and the Pre-Construction Environmental Report (PCER). Additionally, these are supplemented by a number of supporting documents, providing further information in a number of areas.- -- Introduction to the Safety, Security and Environmental Report (SSER)The Introduction to the Safety, Security and Environmental Report provides an outline of the contents of the Pre-Construction Safety Report (PCSR) and the Pre-Construction Environmental Report (PCER), together with a list of definitions and acronyms used in the PCSR and PCER.- file size 435 Kb UK EPR™ GDA Design Reference ConfigurationThe Design Reference Configuration defines the main technical requirements for the UK EPR™ design and the reference design configuration supporting the GDA submission including design changes and out of scope items. This design reference agreed at the end of GDA to be considered as a starting point for future licensees.- file size 222 Kb UK EPR™ GDA Submission Master ListThe Submission Master List is a list of documentation which defines the complete UK EPR™ GDA Submission configuration- file size 2 Mb |
Supporting DocumentsAs part of the UK EPR GDA submission, a number of supporting documents have been provided which supplement the information provided in the Safety, Security and Environmental Report (SSER). The topics covered are largely in the waste and decommissioning topic area and include: waste and spent fuel management options; interim storage facilities for spent fuel and Intermediate Level Waste; decommissioning waste inventory; arrangements for monitoring of liquid and gaseous discharges; the longer term proposed storage facilities for Intermediate Level Waste and spent fuel; the EPR integrated waste strategy; and the application of Best Available Techniques. A comparison of the EPR design against the HSE/NII Safety Assessment Principles has also been provided.- -- PPC Application – Generic information for UK EPR™ diesel generatorsThis document provides relevant information about UK EPR™ installations that could be subject to Pollution Prevention Control (PPC) Regulations 2000 (now incorporated into the Environmental Permitting Regulations). Within the Generic Design Acceptance phase, the only installation identified as being subject to a PPC permit are the standby emergency diesel generators, as the total thermal input of these combustion plants will exceed 50 MW. The standby emergency diesel generators only operate for around 100 hours per year; nevertheless, it is a requirement of the application for a PPC Permit that the installation and the process equipment, technology and management controls be assessed against indicative Best Available Techniques (BAT) defined in relevant Sector Guidance Note. Therefore this document provides an assessment of BAT, with justification of the processes and / or equipment used to explain how it is or will have to be answered on site by the Operator.- file size 811 Kb Comparison of EPR™ Design with HSE_NII SAPSThis document provides an assessment of the UK EPR™ design against the HSE_NII Safety Assessment Principles (SAPs), which express HSE_NII views on relevant good practice in reactor design and operation, and against which it assesses safety submissions for civil nuclear facilities in the UK. This comparison between the UK EPR™ design and the expectations of the SAPs is intended to be a contribution to the demonstration that the EPR™ design process has followed "relevant good practice", as required by the guidance from HSE_NII in application of the ALARP principle.The EPR™ design was developed within a French and German framework involving both national Safety Authorities, following a specific set of recommendations for the design of new PWRs, known as the "Technical Guidelines", which were the fundamental requirements applied to the EPR™ design. Subsequently, the EPR™ design was compared against international standards such as IAEA safety guidelines, EURs and WENRA reference levels. These guidelines and principles do not correspond in all respects to the recommended practices suggested in the SAPs. Nonetheless, it is considered that all the key nuclear safety requirements embodied in the SAPs are met by the UK EPR™ design, and in particular that the UK EPR™ achieves the fundamental objective that the radiological risk to workers and the public is as low as reasonably practicable, which is the basic legal requirement underpinning UK nuclear safety regulations.- file size 1,985 Kb Solid Radioactive Waste Strategy Report (SRWSR)This document presents waste and spent fuel management options to the Flamanville 3 reference case for the UK EPR™. The SRWSR describes how solid radioactive waste and spent fuel generated by the UK EPR™ over its complete lifecycle can be managed within the constraints of the current UK Government Policy and regulatory requirements. The SRWSR describes and characterises the predicted arisings of solid radioactive waste and spent fuel from the UK EPR™, and sets options for waste and spent fuel treatment, conditioning, packaging, record keeping, storing, transporting and disposal. It identifies the facilities required to manage waste and spent fuel arisings of a single EPR™ for up to 100 years from the start of its 60 year operational life. It also considers the impact of multi-reactor units as a variant to the base case configuration. A description of the Waste Treatment Building for operational waste and Interim Storage Facilities for Intermediate Level Waste and for spent fuel for the UK EPR™ are provided. The SRWSR describes how the generation of waste will be avoided or minimised where reasonably practicable.- file size 6,088 Kb Interim Storage Facility for Spent Fuel Assemblies coming from an EPR PlantThis document presents a clear and concise description of the design and operation of the proposed wet (pool – base) interim storage facility for spent fuel assemblies from a UK EPR™ plant. The nuclear safety principles are described and an outline of the safety provisions and functions is presented. The document lists the legislation and safety requirements that will be met by the plant, and describes how these will be implemented. None of the features are novel, they all employ proven technology. Cask, fuel handling, and pool storage have been successfully used around the world at a large number of sites with positive feedback. The document demonstrates that there is nothing in the proposed design that will not conform to UK legislation and standards, and concludes that the proposed interim storage will be suitable for licensing, construction and operation in the UK.- file size 1,074 Kb Dry Interim Storage facility for ILWThis document presents a clear and concise description of the design and operation of the proposed interim storage facility for intermediate level waste packages from a UK EPR™ plant for 60 years of operation. The document describes the UK legislation and safety requirements that will be met by the facility, and the methodology that will be followed for the detailed design studies of the facility. Although the detailed design is not complete, none of the features are novel, they all employ proven technology. The document demonstrates that the proposal is licensable in the UK.- file size 2,697 Kb EPR™ UK – Decommissioning waste inventoryThis document provides an evaluation of UK EPR™ decommissioning waste, based on the assumption that decommissioning begins immediately after 60 years of UK EPR™ operation, in order to maximise activated matters volumes. The analysis identifies the masses in the different level of waste categories, depending on the origin of the equipment and structure in the plant, and includes waste generated by decommissioning activities. The study also introduces types and volumes of packaged nuclear wastes.- file size 99 Kb Monitoring of Liquid and Gaseous Discharges: Prospective Arrangements for the UK EPR™This document provides additional details on the monitoring regime for both liquid and gaseous effluents prior to discharge that could be implemented on the UK EPR, by looking at the arrangements currently in place on existing stations, both on the 1300 MW French fleet and at a German KONVOI site. Monitoring arrangements for both liquid and gaseous effluents are presented, including both radiochemical and associated chemical substances, from their production through to collection, storage and discharge. The information given here provides additional detail to that presented in PCER Sub-chapter 7.3 (radioactive effluents), PCER Sub chapter 7.4 (chemical associated with radioactive effluents) and PCER Sub-chapter 6.2 (management of liquid effluents). The document does not deal with monitoring of substances in the environment, which has been dealt with in Chapters 11 and 12 of the PCER.- file size 388 Kb Longer Term ILW Interim Storage FacilityThis document demonstrates that the risks over the lifetime of an interim storage facility for Intermediate Level Waste (ILW) are as low as reasonably practicable. Information is provided covering the design safety principles and safety aims for the storage facility. A preliminary assessment of hazards and risks is provided to demonstrate that the risks over the lifetime of the facility will be controlled to levels that are as low as reasonably practicable (ALARP). The means by which the need for active safety management during longer-term storage has been minimised are described. Potential waste package damage mechanisms are outlined, together with options for dealing with damaged or degraded waste packages. The provisions and functions necessary for the retrieval and inspection of waste packages and potential inspection regimes are discussed, including consideration of changes to the retrieval and inspection regimes as materials age and their characteristics change. Finally, plans for the facilities and functions needed to retrieve the waste packages and prepare them for onward processing or disposal are outlined.- file size 352 Kb Longer Term Spent Fuel Interim Storage FacilityThis document demonstrates that the risks over the lifetime of an interim storage facility for spent fuel are as low as reasonably practicable. Information is provided covering the design safety principles and safety aims for the storage facility. A preliminary assessment of hazards and risks is provided to demonstrate that the risks over the lifetime of the facility will be controlled to levels that are as low as reasonably practicable (ALARP). The means by which the need for active safety management during longer-term storage has been minimised are described. The characteristics of the fuel and equipment whose integrity must be maintained over the storage period are identified, including their evolution over that period. Potential fuel damage and containment degradation mechanisms are described, together with options for dealing with damaged fuel. The provisions and functions necessary for the retrieval and inspection of spent fuel and potential inspection regimes are discussed, including consideration of changes to the retrieval and inspection regimes as materials age and their characteristics change. Finally, plans for final fuel retrieval are outlined.- file size 780 Kb Integrated Waste Strategy DocumentThis Integrated Waste Strategy (IWS) document outlines the current strategy for managing radioactive and non radioactive wastes arising from the construction, operation and decommissioning of the UK EPR™. It shows that there is a management strategy for all the waste streams produced by the UK EPR™ design and that they have been suitably planned for. The IWS refers out to a range of other UK EPR™ GDA submission documentation which provide more details on the how, why and when waste management strategies on the UK EPR™ are developed. The document shows that waste management strategies have been developed using a standardised approach which takes into account all relevant factors including: delivery of compliance with relevant regulatory obligations and Government policy; consideration of a full range of health, safety, environmental, security, economic and social issues; minimisation of waste via implementation of the waste hierarchy; and application of Best Available Techniques (BAT). The IWS identifies uncertainties and assumptions that may influence the implementation of an EPR™ waste management strategy in the UK and also recognises the risks and opportunities within the strategy.- file size 513 Kb GDA UK EPR™ - BAT DemonstrationThis document provides a key reference to the GDA UK EPR™ Integrated Waste Strategy (IWS) document. The information presented demonstrates that the Best Available Techniques (BAT) are being applied in the design of the EPR™ to minimise radioactive wastes at source and to minimise the impacts of the disposal of wastes into the environment. Much of the information presented has been drawn from the PCER, but other information has also been used, including international literature relating to waste management and detailed technical reports. Potential radionuclides are identified, and their significance assessed. For the small number of radionuclides that are assessed as significant, more detailed information has been provided in "BAT forms", in particular on the minimisation techniques applied and conformance with the IWS. The report covers the radioactive wastes arising from reactor operations, noting that the optimised EPR™ design will influence decommissioning waste management requirements.- file size 656 Kb GDA UK EPR™ - Mapping Document for Radioactive Waste Management CaseThis mapping document demonstrates how a Radioactive Waste Management Case (RWMC) can be achieved for the UK EPR™. The wastes considered are the higher activity radioactive wastes, including spent fuel, that are generated through operation, maintenance and decommissioning of a UK EPR™. Intermediate level waste and spent fuel are dealt with separately in this document to provide clear and concise references and to conform to the guidance provided to nuclear licensees by the UK Regulators. This mapping for the RWMC provides a demonstration that support for safe operation of plant, processes and activities associated with the UK EPR™ will comply with regulatory requirements. Furthermore, this document provides evidence that there is a commitment by the licensee to meet national and international standards of radioactive waste management for the waste streams produced by the UK EPR™ design. The conclusion is that the RWMC requirements are addressed to a level of detail that is appropriate to this stage of the GDA process.- file size 667 Kb UK EPR™ LLWR Disposability Assessment – Preliminary D1 Form InformationThis document provides the forms which were sent to LLW Repository Ltd to support their preliminary disposability assessment of the UK EPR™ Low Level Waste. These forms contain information related to the types of UK EPR™ waste for which acceptance is sought from LLWR. They include information on the nature, volume and activity of the waste.- file size 442 Kb Form D1 Application: UK EPR™ ProjectThis document provides the letter from LLW Repository Ltd giving agreement in principle for treatment / disposal through LLW Repository Ltd of the waste described in the "Preliminary D1 forms".- file size 58 Kb GDA: Summary of Disposability Assessment for Wastes and Spent FuelThis document provides a summary of the GDA Disposability Assessment produced by the Nuclear Decommissioning Authority (NDA) on the disposability of any proposed arisings of higher activity wastes or spent fuel in a Geological Disposal Facility. Details of the wastes and spent fuel, and their characteristics applied in the assessment are provided. The GDA Disposability Assessment process is summarised, together with the radionuclide assessment inventories for ILW and spent fuel derived for the purposes of the GDA Disposability Assessment. More comprehensive details are provided in the separate assessment reports (parts 1 and 2).- file size 999 Kb GDA: Disposability Assessment of Wastes and Spent Fuel - Part 1 Main ReportThis document presents the results of the Disposability Assessment undertaken by the Radioactive Waste Management Directorate (RWMD) of the Nuclear Decommissioning Authority (NDA) on the disposability of any proposed arisings of higher activity wastes or spent fuel in a Geological Disposal Facility. The report provides comprehensive details of the wastes and spent fuel, and their characteristics, and assesses the implications of the disposal of the proposed waste packages against the waste package standards and specifications developed by RWMD and the supporting safety assessments for a Geological Disposal Facility. The safety of transport operations, handling and emplacement at a Geological Disposal Facility and the longer-term performance of the system have been considered, together with the implications for the size and design of a Geological Disposal Facility. Overall, it is concluded that the spent fuel and ILW wastes should be disposable and that no new issues are expected to arise.- file size 1,024 Kb GDA: Disposability Assessment of Wastes and Spent Fuel - Part 2 Data Sheets and Inventory TablesThe Disposability Assessment undertaken by the Radioactive Waste Management Directorate (RWMD) of the Nuclear Decommissioning Authority (NDA) on the disposability of any proposed arisings of higher activity wastes or spent fuel in a Geological Disposal Facility assesses the implications of the disposal of the proposed waste packages against waste package standards and specifications. Part 1 of the report is the Main Report. Part 2 (this document) provides data summary sheets and inventory data for waste packages. It provides data summary sheets for the different waste packages advised by EDF and AREVA and detailed inventory tables which were developed for the purposes of the GDA Disposability Assessment.- file size 526 Kb Critique of the NDA RWMD Disposability AssessmentThis document provides an EDF/AREVA critique of the NDA RWMD Disposability Assessment reports. The principal issues arising from the review process are outlined, and each issue is briefly discussed, noting how the issue is expected to be resolved. Overall, EDF and AREVA concur with RWMD that the issues can be addressed by the operator under the Letter of Compliance process.- file size 46 Kb The Case for Disposability of Spent Fuel and ILWThis document provides information on how and when the issues raised by the Radioactive Waste Management Directorate (RWMD) Disposability Assessment of EPR™ waste and spent fuel will be addressed. Each of the issues is discussed and it is concluded that it will be the responsibility of the site operator to address the issues discussed as part of the normal site licensing and regulatory processes.- file size 230 Kb GDA UK EPR™ - DecommissioningThis document demonstrates the feasibility of decommissioning the UK EPR™ (including decommissioning of the interim storage facilities for Spent Fuel and ILW), using current technology, and demonstrates that the design allows decommissioning to be carried out in a controlled, careful and systematic manner. The baseline decommissioning plan covers the entire lifecycle of the site, including achieving operational shutdown, performance of Post Operational Clean Out, a care and maintenance period, if this should be required, and decommissioning to a state that is suitable for its future intended use.- file size 2,457 Kb Decontamination processes and techniques for the UK EPR™This document provides additional information on decontamination during operation, maintenance and decommissioning. The following issues are addressed: decontamination during operation and maintenance; minimisation of decontamination waste arising from the UK EPR™; and decontamination during Post Operation Clean Out (POCO) and during decommissioning.- file size 637 Kb |
