Public Version of UK EPR™ GDA Submission
 EDF and AREVA have applied 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.  The UK Health and Safety Executive 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. |
SSER
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 is undergoing 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 regulatory 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 98 Kb Sub-Chapter 1.2 - General Description of the UnitSub-chapter 1.2 provides a general overview of the UK EPR design. 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 plant.-file size 1,146 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 design 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 288 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 169 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 reactor 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 192 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, together with a set of representative site parameters representative for the UK in order to carry out 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 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. Following Generic Design Assessment (GDA) it will be necessary to further update this chapter to include site specific details and to utilise the available site-specific data.-file size 111 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). At this stage, 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 parameters appropriate for the development of an EPR in the UK, i.e. representative of a range of potential coastal and estuarine locations, in order to carry out impact studies for Generic Design Assessment (GDA). Following 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 242 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 principles used to define the safety and functional requirements of safety-related equipment and their application to systems and equipment 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 reactor operating in the UK to address UK specific safety principles.-file size 804 Kb Sub-Chapter 3.2 - Classification of structures, equipment and systemsSub-chapter 3.2 presents the classification principles used in the EPR design, the aim of which is to help define the safety and functional requirements (e.g. quality, redundancy) of safety-related equipment. The two main classification systems are described: the mechanical classification which addresses the pressure barrier role of mechanical components, and the functional classification, which addresses the performance of systems required during fault transients; together with the seismic and structural classification systems which describe how the safety significance is interpreted in terms of relevant specific technical requirements. The application of the classification principles to systems, electrical equipment, I&C equipment is presented. All safety-classified systems, components and structures are divided into different classes, with corresponding requirements depending on the safety functions to be fulfilled; requirements are graduated according to the importance of the safety duty being performed.-file size 404 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 design. The safety requirements are established as follows: a review is carried out of the relevant structures, identifying the various components of the structures to which particular requirements may apply; the various load cases applicable to the structures are specified, covering both normal reactor operation and faults (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, are identified; and the safety requirements are then incorporated into the design of the civil structures and into the definition of their associated behaviour requirements. The civil structures to which the safety requirements are applied are those having a role with regard to the three basic safety functions of: reactivity control, fuel cooling and containment of radioactive materials. These 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 563 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 stream 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 3,143 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 173 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 306 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 117 Kb Sub-Chapter 3.8 - Codes & 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 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.-file size 214 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 208 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 103 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 106 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 422 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 634 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 123 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 62 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 and assessment of mechanical integrity in accident conditions; manufacturing and inspection details. 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. The application of the break preclusion principle to the main coolant lines in the reactor coolant system is also detailed here.--- 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 135 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 702 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, and the main principles and parameters governing the reactor coolant system water chemistry are given. A description of how the break preclusion principle is applied to the main coolant lines in the reactor coolant system is also given. The safety approach is based on the concept of defence-in-depth, and the requirements related to the further levels of defence-in-depth are described, together with the design criteria for the over-pressure protection system. An outline of the in-service inspection requirements is presented, together with the requirements applied to the design, manufacture, inspection and in-service surveillance for "non breakable" nuclear pressurised equipment, including protection against internal and external hazards. Finally, a comparison of the design principles applied to break preclusion and "non-breakable" components with the requirements conventionally applied to "incredibility of failure components" in UK power reactors is provided.-file size 286 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 preliminary safety evaluation is given, including a description of the fracture mechanics analyses performed to assess the margins to fast fracture. In-service inspection requirements are given, together with manufacturing requirements.-file size 798 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 and assessment of mechanical integrity in accident conditions; manufacturing and inspection details. 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 1,487 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 unit.-file size 104 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 1,904 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 1,071 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 80 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 114 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 276 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 242 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 218 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 232 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 two I&C platforms, including the reliability, availability and functional requirements; the software and hardware programming design and configuration; system integration, installation, and commissioning; verification and validation. Also included are four appendices which provide a more detailed description of I&C systems, based on the specific design of the French Flamanville 3 EPR reactor.--- 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.-file size 175 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. 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 469 Kb Sub-Chapter 7.3 - F1 classified Instrumentation & Control systemsSub-chapter 7.3 describes the protection system, the safety automation system, the safety information and control system, and the priority and actuation 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 systems are either, required to attain a controlled or safe state after a fault, or failure during normal power plant operation would lead to a fault.-file size 1,340 Kb Sub-Chapter 7.4 - F2 classified and non-classified Instrumentation & Control systemsSub-chapter 7.4 describes the process information and control system, the process automation system, the reactor control surveillance and limitation system, the RRC-B safety 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 472 Kb Sub-Chapter 7.5 - InstrumentationSub-chapter 7.5 describes the instrumentation used in the UK EPR design. It covers: conventional, in-core and ex-core instrumentation; rod position measurement; reactor pressure vessel water level measurement; loose parts monitoring and vibration monitoring; radiation monitoring; accident and severe accident instrumentation; 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 373 Kb Sub-Chapter 7.6 - I&C procedures and toolsSub-chapter 7.6 provides information on the I&C procedures and tools used for the design of the two I&C platforms: the standard I&C system for the process information and control system, process automation system and safety automation system; and the Teleperm XS platform for the protection system and the reactor control surveillance and limitation 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.-file size 128 Kb Appendix 7A - Detailed description of the I&C systems - example - General description of Nuclear Island I&CAppendix 7A is included to provide a more detailed description of the I&C systems, based on the specific design of the Flamanville 3 EPR reactor, currently under construction in France. This appendix gives an overview of the Nuclear Island I&C, including: definition of the functional objectives assigned to I&C for safety and operation of the plant, a summary of the main processes and criteria applied in the design of I&C systems; a description of the I&C systems and equipment implemented to fulfil the functional objectives and the design criteria; and the human factors approach to the design of the equipment provided to operate the plant.-file size 1,821 Kb Appendix 7B - Detailed description of the I&C systems - example - Nuclear Island protection safety and monitoring systemsAppendix 7B is included to provide a more detailed description of the I&C systems, based on the specific design of the Flamanville 3 EPR reactor, currently under construction in France. This appendix contains a more detailed technical description of the Nuclear Island protection safety and monitoring systems: the protection system, the safety information and control system, and the safety instrumentation.-file size 962 Kb Appendix 7C - Detailed description of the I&C systems - example - Nuclear Island Control and Monitoring SystemsAppendix 7C is included to provide a more detailed description of the I&C systems, based on the specific design of the Flamanville 3 EPR reactor, currently under construction in France. This appendix contains a more detailed technical description of the nuclear island control and monitoring systems: the process information and control system, safety information control system, safety and process automation system, reactor control, surveillance and limitation system, the I&C systems for severe accidents and process instrumentation.-file size 2,317 Kb Appendix 7D - Detailed description of the I&C systems - example - System diversity and redundancy in control and protectionAppendix 7D is included to provide a more detailed description of the I&C systems, based on the specific design of the Flamanville 3 EPR reactor, currently under construction in France. This appendix contains a more detailed technical description of the diversity and redundancy in the control and protection system, including the basic principles for the definition of the I&C architecture, the independence and separation principles, and diversity concepts.-file size 819 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.--- 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 76 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 60 Kb Sub-Chapter 8.3 - Nuclear Island Power SupplySub-chapter 8.3 describes the emergency power supplies and power distribution systems 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.-file size 432 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 89 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 83 Kb Chapter 9 - Auxiliary SystemsChapter 9 provides details of the following systems: 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 1,639 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 522 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 the two separate sampling systems, one for the primary system and a second for the secondary side of the steam generators; 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 1,163 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 5,819 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; 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.-file size 789 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 application of the break preclusion principle 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 81 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, and 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.-file size 398 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; 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 feedwater plant, cooling water and turbine gland system, is operator and/or site specific, and is not included here.-file size 368 Kb Sub-Chapter 10.5 - Implementation of the Break Preclusion PrincipleSub-chapter 10.5 describes how the break preclusion principle is applied to the main steam lines inside and outside the reactor building containment. The safety approach is based on the concept of defence-in-depth, and the requirements related to the further levels of defence-in-depth are described, together with the areas of application of the break preclusion principle.-file size 103 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 202 Kb Chapter 11 - Discharges and Waste - Chemical and RadiologicalChapter 11 provides details of the radioactive waste 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, 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, 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.-file size 134 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, 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 401 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 maximum additional annual discharges are provided, together with estimates of monthly discharges of gaseous and liquid radioactive waste. The EPR expected performance is derived from recent operational feedback, taking credit for design features and guaranteed quantifiable gains.-file size 1,706 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 radioactive effluent systems, in normal operation, involves: collecting radioactive 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 radioactive 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 neighbouring the reactor building, the radioactive effluent 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 1,484 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 Intermediate Level Waste (ILW) radioactive operational waste, and the spent fuel, produced during the UK EPR reactor's 60 years of operation. 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.-file size 254 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 reactor 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 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 reactor 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 75 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 reactor, including optimisation of the plant radioactive inventory and shielding provisions, and compliance with individual and collective dose limits.-file size 61 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 143 Kb Sub-Chapter 12.3 - Radiation protection measuresSub-chapter 12.3 provides details of the radiation protection measures implemented on the UK EPR reactor including: radiation protection classification and zoning, and rules for the design (and installation) of equipment and rooms; shielding provisions, including material 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 589 Kb Sub-Chapter 12.4 - Dose uptake optimisationSub-chapter 12.4 describes the design measures adopted in the UK EPR plant 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 plant.-file size 208 Kb Sub-Chapter 12.5 - Post accident accessibilitySub-chapter 12.5 defines the systems and components to which access is required in 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 operators 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 62 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.-file size 594 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; 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.-file size 624 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 185 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 382 Kb Sub-Chapter 14.2 - Analysis of the Passive Single FailureThe PCC accident analyses presented in Sub-chapters 14.3 to 14.5 take into account an active single failure; these analyses do not take into account a passive single failure. Sub-chapter 14.2 provides the demonstration that the safe shutdown state can be maintained in the long term despite a passive single failure (i.e. a failure that occurs in a component which does not need to change state to carry out its function, such as a leak in the pressure boundary). An indication of the technical measures to detect and control any leak resulting from the passive single failure is also provided.-file size 108 Kb Sub-Chapter 14.3 - Analyses of PCC-2 eventsSub-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.This sub-chapter contains Security Sensitive material and is therefore not publically available--- 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 10,00 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.This sub-chapter contains Security Sensitive material and is therefore not publically available--- 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 10,00 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.This sub-chapter contains Security Sensitive material and is therefore not publically available--- 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.This sub-chapter contains Security Sensitive material and is therefore not publically available--- Sub-Chapter 14.7 - Fault and Protection ScheduleSub-chapter 14.7 provides a summary of the fault and protection schedule for the UK EPR design. The fault and protection schedule is a list of the postulated faults conditions that are considered in the EPR design due to their potentially unacceptable consequences, and it includes the initiating events, their frequencies and consequences, the classified safety systems (and beneficial non-safety classified systems) that protect against them and the overall protection claims. The rationale for and a justification of 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 the principles applied to modify or deactivate settings during changes in the reactor state.This sub-chapter contains Security Sensitive material and is therefore not publically available--- 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.--- 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 reactor. 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).This sub-chapter contains Security Sensitive material and is therefore not publically available--- 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 containment response in those situations, and assigns a release category to containment fault sequences that present the same characteristics of activity release to the off-site environment. 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. Both the numerical targets defined in the HSE/NII Safety Assessment Principles, and quantitative probabilistic safety objectives in accordance to the French Technical Guidelines, are presented.-file size 101 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).This sub-chapter contains Security Sensitive material and is therefore not publically available--- 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 reactor. 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, whereby an external hazard is screened out if: there is no impact expected on the plant safety; the expected frequency of core damage is sufficiently low; or the frequency of the external hazard is low. 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.This sub-chapter contains Security Sensitive material and is therefore not publically available--- 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. 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 events involving non-draining and draining of the spent fuel pool, are presented in terms of derived event sequence frequencies and the risk of fuel damage (per reactor per year). This sub-chapter contains Security Sensitive material and is therefore not publically available--- Sub-Chapter 15.4 - Level 2 PSASub-chapter 15.4 presents the Level 2 probabilistic safety analysis (PSA), which allows evaluation of the nature, magnitude and frequency of radioactive releases outside the containment boundary. 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 use of containment event trees; definition of fission product release categories; and treatment of sensitivities and uncertainties. The results of the Level 2 PSA are presented in terms of the frequencies, release characteristics, and release risk associated with each release category, and the frequencies of the individual core damage end states contributing to the total core damage frequency. This sub-chapter contains Security Sensitive material and is therefore not publically available--- Sub-Chapter 15.5 - Level 3 PSA: Assessment of off-site risk due to postulated accidentsSub-chapter 15.5 summarises the process used to perform a probabilistic assessment of the radiological impact of the UK EPR design in terms of the off-site risk to the public due to postulated accidents (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 eventual fatalities in the wider UK population. This sub-chapter contains Security Sensitive material and is therefore not publically available--- Sub-Chapter 15.6 - Seismic Margin AssessmentSub-chapter 15.6 presents the Seismic Margin Assessment (SMA) performed for the generic design assessment (GDA) of the UK EPR design. A full seismic probabilistic safety analysis cannot be performed at the GDA stage, as no specific site has been selected; hence a seismic margin assessment is carried out to assess the plant seismic capacity in order to demonstrate that sufficient margin exists in the seismic design of the main structures and components beyond the design basis conditions. The seismic margin assessment demonstrates that the seismic capacity of the UK EPR design is greater than 1.6 times the Design Basis Earthquake (DBE) and that there will be no disproportionate increase in risk from a realistic range of events which are more severe than the design basis event (i.e. there is no cliff edge effect).This sub-chapter contains Security Sensitive material and is therefore not publically available--- 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 options resulting from PSA studies are presented here. Unavailability due to repairs and preventive maintenance activities is included in the Level 1 PSA model; quantification of the results without preventive maintenance is also presented to assess the impact of the preventive maintenance programme.. Other sensitivity analyses presented here include: the effects of transients longer than 24 hours, 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. This sub-chapter contains Security Sensitive material and is therefore not publically available--- 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.--- 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 means (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.This sub-chapter contains Security Sensitive material and is therefore not publically available--- 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. This sub-chapter contains Security Sensitive material and is therefore not publically available--- 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 particularly considered in this approach are detailed here.-file size 130 Kb Sub-Chapter 16.4 - Specific StudiesSub-chapter 16.4 presents an assessment of two specific faults: a double-ended break of the main coolant line (2A-LOCA) and a double-ended break of the main steam line outside the containment (2A-SLB). Assessment of the faults is presented, including description of the analysis methods, main assumptions and models used. The 2A-LOCA analysis concludes that no rods rupture during the 2A-LOCA transient. The double-ended main steam line break is taken into account as a defence-in-depth measure, and a large margin to the safety criterion is demonstrated.This sub-chapter contains Security Sensitive material and is therefore not publically available--- 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 Chapter 16.2 these codes are used. Each code is described, together with an outline of its validation basis, where required. 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 145 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 analysis (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 design in the frame of the PCSR studies. The analysis presented in this appendix should be read in conjunction with section 3.8 of Sub-chapter 16.1.This sub-chapter contains Security Sensitive material and is therefore not publically available--- 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 Level 3 probabilistic safety analysis demonstrating that the risk to the public due to accidents is 'broadly acceptable' are summarised. An assessment of additional modification options requested by the US or Finnish regulators, and modifications implemented in the Sizewell B PWR, concludes that these are not considered to be reasonably practicable according to the ALARP 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 61 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 HSE/NII 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 253 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 172 Kb Sub-Chapter 17.4 - Review of PSA Results: Comparison with Numerical Risk TargetsSub-chapter 17.4 summarises the results of the level 3 probabilistic safety analysis (PSA) for the UK EPR design (presented in Sub-chapter 15.5 of the PCSR), which demonstrates that the risk to the public due to accidents meets the HSE 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 83 Kb Sub-Chapter 17.5 - Review of Possible Design Modifications to Confirm Design meets ALARP PrincipleSub-chapter 17.5 considers additional modification options requested by the US or Finnish regulators, and modifications implemented in the Sizewell B PWR, and assesses if these are warranted for the design of the UK EPR reactor, under the UK principles of ALARP. None of the modifications considered is indicated as reasonably practicable according to the ALARP methodology.-file size 212 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 plant complies with the overall requirements of the ALARP principle, which is a key safety design objective for the UK EPR plant.-file size 87 Kb Chapter 18 - Human-Machine Interface and Operational AspectsChapter 18 describes the human factors engineering (HFE) programme to be implemented for development of the man-machine interface; and the design principles applied to the man machine interface. The operating principles to be applied to the UK EPR plant to ensure safe operation in both normal and emergency conditions are also described. Achieving the safety and availability objectives required for the EPR plant requires advanced design skills. The part played by the human operators in running the installation requires careful consideration from the outset. Human operators may be fallible, but they are essential to nuclear power plant operation. They are key to performance and safety, particularly in managing unexpected situations and in optimising daily operation, if they form an integral part of a well-managed working environment. The HFE programme is presented here as it is implemented for the Flamanville 3 EPR, currently under construction in France; certain aspects will need to be revisited for the UK EPR design to take into account the UK socio-cultural context.--- Sub-Chapter 18.1 - Human-Machine InterfaceSub-chapter 18.1 describes the human factors engineering (HFE) programme to be implemented for development of the human-machine interface. It outlines the requirements, objectives and scope of the HFE programme, together with the methodological principles used and how they are implemented in practice through the HFE programme. The HFE programme includes: human-machine interface design, task allocation and organisation, procedures, organisational aspects, staffing and qualification requirements, for both operational and maintenance and testing activities, and training. The design principles applied to the human machine interface are described, covering areas such as the role of operating personnel, their information and control needs, alarms and operating procedures, together with the human machine interface systems (primarily the main control room, remote shutdown station and technical support centre, and the equipment contained therein).-file size 367 Kb Sub-Chapter 18.2 - Normal OperationSub-chapter 18.2 describes the operating principles and procedures to be applied to the UK EPR plant during normal operation. The main operating principles and procedures for normal operation are described, covering the full fuel cycle, together with the implementation of preventive maintenance principles related to maintenance activities performed as the plant is in power range conditions, which allow optimisation of outage duration. The plant must be operated in a manner consistent with the safety case assumptions and, to achieve this, operating rules are developed using Operating Technical Specifications, which set out the normal operating rules, and Periodic Tests, which are performed on systems that carry out safety functions identified in the Safety Case. Different approaches can be applied to define the Operating Technical Specifications and the chosen approach will be decided by the Licensee; however, two possible approaches are presented.-file size 197 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 195 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 56 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 129 Kb Chapter 20 - Design Aspects in relation to the DecommissioningChapter 20 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.--- Sub-Chapter 20.1 - General 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.-file size 106 Kb Sub-Chapter 20.2 - Implementation for the EPRSub-chapter 20.2 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.-file size 113 Kb Chapter 21 - Quality and Project ManagementChapter 21 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 can only be fully defined following a successful generic design assessment, and 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, post-GDA, to control the construction of a new plant in accordance with the applicable standards and regulations.-file size 163 Kb Sub-Chapter 21.2 - Quality & Environmental (Q&E) ManagementSub-chapter 21.2 outlines the quality assurance arrangements established for the UK EPR generic design assessment (GDA) project, and the quality and environmental management (QEM) systems of the supporting organisations (EDF, AREVA and AMEC), including documentation, control, records, design review and change management processes, and audits.-file size 128 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 unit 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 regulatory 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 156 Kb Sub-Chapter 1.2 - General Description of the UnitSub-chapter 1.2 provides a general overview of the UK EPR unit. 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 unit.- file size 953 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 unit 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 148 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 171 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 172 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 can only be fully defined following a successful generic design assessment, and 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, post-GDA, to control the construction of a new plant in accordance with the applicable standards and regulations. Sub-chapter 2.2 outlines the quality assurance arrangements established for the UK EPR generic design assessment (GDA) project, and the quality and environmental management (QEM) systems of the supporting organisations (EDF, AREVA and AMEC), including documentation, control, records, design review and change management processes, and audits.- file size 237 Kb Chapter 3 - Aspects having a Bearing on the Environment during Operation PhaseChapter 3 provides an overview of the interfaces of the UK EPR plant 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 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 504 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 864 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 the radioactive dose received by workers, and the amount of radioactive waste and hazardous material produced are reduced to as low as reasonably practicable. Sub-chapter 5.2 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.- file size 196 Kb Chapter 6 - Discharges and Waste - Chemical and RadiologicalChapter 6 provides details of the radioactive waste 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, 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 radioactive waste management systems, which include the liquid and gaseous waste processing systems, the solid waste treatment system.- file size 62 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.- file size 111 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 344 Kb Sub-Chapter 6.3 - Outputs for the operating installation.pdfSub-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 guaranteed quantifiable gains. Maximum additional annual discharges are provided, together with estimates of monthly discharges of gaseous and liquid radioactive waste.- file size 1,545 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 radioactive effluent systems, in normal operation, involves: collecting radioactive 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 radioactive 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 neighbouring the reactor building, the radioactive effluent 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 1,026 Kb Sub-Chapter 6.5 - Interim storage facilities and disposability for UK EPRSub-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. Additionally, the current status of the discussions with the NDA, with respect to the disposability of UK EPR reactor 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.- file size 308 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 plant. 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 the 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 an authorisation 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 313 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 613 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 design 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 plant. 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 154 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 sub-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 235 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 651 Kb Chapter 12 - Non-radiological Impact AssessmentChapter 12 describes the methodology used to assess the non-radiological impacts due to a UK EPR plant. 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 unit. 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 756 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 HSE 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 Safety, Security and Environmental Report (SSER) for the proposed UK EPR design addresses the requirements of the HSE 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.- file size 264 Kb |
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. 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 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 out options for waste and spent fuel management including 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 spent fuel (dry and wet storage options) and Intermediate Level Waste 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 EPR decommissioning waste, based on the assumption that decommissioning begins immediately after 60 years of of UK EPR operation, in order to provide envelope volumes of activated materials. 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 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 of 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 of 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 443 Kb Monitoring of liquid and gaseous dischargesThis 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 316 Kb 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 561 Kb RWMC MappingThis mapping document identifies existing documentation forming the basis of the Radioactive Waste Management Case (RWMC) for the UK EPR. It deals with higher activity radioactive waste 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 325 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 1024 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 are 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 |
