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Fontenay-aux-Roses, France 2nd – 18th July 2014 IRSN, Dimitrov Borislav

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Fontenay-aux-Roses, France 2nd – 18th July 2014 IRSN, Dimitrov Borislav

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  1. TECHNICAL ASSISTANCE FOR IMPROVING THE LEGAL FRAMEWORK FOR NUCLEAR SAFETY AND STRENGTHENING THE CAPABILITIES OF THE REGULATORY AUTHORITY OF VIETNAM (VARANS) AND ITS TSOProject VN/RA/01, Task 3, Topic 1 On-Job-Training on Safety requirements, systems, protection against external hazards and accidental studies GENERAL REQUIREMENTS FOR THE PSAR GEN III REACTORS TO BE VERIFIED IN THE SAFETY REVIEW OF THE PSAR OF BELARUS NPP , IRSN PRACTICE Fontenay-aux-Roses, France 2nd – 18th July 2014 IRSN, Dimitrov Borislav

  2. MAIN TOPICS • SAFETY ANALYSIS REPORT – GENERAL • SAFETY ANALYSIS REPORT – SPECIFICS • SAR content • SAR logical organization • Safety regulations • Innovative improvements • Defense in depthpresentation • Design principles and rules • Safety margins • Codes validation • Uncertainties • PSA - Probabilistic safety assessment • Design details • CONCLUSIONS Project VN/RA/01, Task 3 – On-Job-Training, Safety requirements, Paris, 02 -18 June 2014

  3. For defined safety objectives, the stakeholders need a detailed demonstration on the plant safety and radiation effects According to the IAEA guidelines, the document containing such demonstration can be referred to as the Safety Analysis Report According to the European practice, the SAR is the basic document used in the licensing process for all stages as design, construction, commissioning, and operation The SAR shows-up an important communication tool between the Operator and the Regulatory Body The SAR is a mandatory document for IRSN in performing safety assessment and expertise for NPP in the design or in operation These paper shares IRSN practice (PWR, EPR, VVER) in the assessment of PSAR some general/common requirements, independent of the presented technology SAFETY ANALYSIS REPORT - GENERAL Project VN/RA/01, Task 3 – On-Job-Training, Safety requirements, Paris, 02 -18 June 2014

  4. Thanks to the internationalization of the nuclear safety activities, for the three projects EPR, AP 1000 and VVER 92 for BELENE the SAR documents show very similar content. This facilitates the safety evaluation through the comparison of different design options. In spite of the consolidation on the SAR content, it is to be mentioned that differences in understanding and appreciation remain between west and east practice as far as the way the information is to be presented. West practice - the SAR should prove, demonstrate and provide evidence that defined safety objectives are achieved with a high confidence level. The SAR presentation should facilitate as much as possible his understanding. East practice (VVER) is more caring about to the amount of information and the design details to be provided. SAFETY ANALYSIS REPORT - GENERAL Project VN/RA/01, Task 3 – On-Job-Training, Safety requirements, Paris, 02 -18 June 2014

  5. PSAR CONTENTS The SAR completeness should be carefully examined. That could be facilitated by a logical presentation of the topics from the main objectives to the technical details. International consensus has been reached concerning the general content of the SAR for APWR nowadays. PSAR for EPR, AP 1000 and VVER 92 for BELENE NPP are developed and presented in compliance with the US NRC RG 1.70 (1.206) and the IAEA requirements. Belarus NPP PSAR is developed on the base of Belarusian and Russian Regulations, SAFETY ANALYSIS REPORT - SPECIFICS Project VN/RA/01, Task 3 – On-Job-Training, Safety requirements, Paris, 02 -18 June 2014

  6. PSAR CONTENTS (2) AP 1000 • INTRODUCTION AND GENERAL DESCRIPTION • SITE CHARACTERISTICE • DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT AND SYSTEMS • REACTOR • REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS • ENGINEERED SAFETY FEATURES • INSTRUMENTATION AND CONTROLS • ELECTRIC POWER • AUXILIARY SYSTEMS • STEAM AND POWER CONVERSION   • RADIOACTIVE WASTE MANAGEMENT • RADIATION PROTECTION • CONDUCT OF OPERATIONS • INITIAL TEST PROGRAM • ACCIDENT ANALYSES • TECHNICAL SPECIFICATIONS • QUALITY ASSURANCE • HUMAN FACTORS ENGINEERING • PROBABILISTIC RISK ASSESSMENT EPR • INTRODUCTION AND GENERAL DESCRIPTION • SITE CHARACTERISTICE • GENERAL DESIGN BASES OF STRUCTURES, COMPONENTS, EQUIPMENT AND SYSTEMS   • REACTOR • REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS • CONFINEMENT AND ENGINEERED SAFETY FEATURES • INSTRUMENTATION AND CONTROLS • ELECTRIC POWERSUPPLY • AUXILIARY SYSTEMS • STEAM AND POWER CONVERSION  SYSTEMS • EFFLUENTS  • RADIATION PROTECTION • PLANT OPERATIONS • PLANT COMISSIONING • SAFETY ANALYSES (PCC) • QUALITY ASSURANCE FOR CONSTRUCTION  • MAN MACHINE INTERFACE  • PROBABILISTIC SAFETY ASSESSMENT  • RISK REDUCTION • DECOMMISSIONING Project VN/RA/01, Task 3 – On-Job-Training, Safety requirements, Paris, 02 -18 June 2014

  7. Belene NPP AES 92 INTRODUCTION AND GENERAL DESCRIPTION SITE CHARACTERISTICE DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT AND SYSTEMS   REACTOR REACTOR COOLANT SYSTEM   ENGINEERED SAFETY FEATURES INSTRUMENTATION AND CONTROLS ELECTRIC POWERSUPPLY AUXILIARY SYSTEMS STEAM AND POWER CONVERSION  SYSTEMS RADIOACTIVE WASTE MANAGEMENT RADIATION PROTECTION CONDUIT OF OPERATIONS INITIAL TEST PROGRAM ACCIDENTANALYSES TECHNICAL SPECIFICATIONS  QUALITY ASSURANCE HUMAN FACTOR ENGINEERING  SEVERE ACCIDENTS OTHER ASPECTS FROM IAEA SAFETY GUIDE NO GS-G-4.1 • PSAR CONTENTS (3) Belarus NPP AES 91 • GENERAL DESCRIPTION OF THE PLANT • DESCRIPTION OF THE REGION AND NPP SITE   • GENERAL PROVISIONS AND APPROACHES TO THE DESIGNING OF BUILDINGS, STRUCTURES, SYSTEMS AND COMPONENTS • REACTOR  • PRIMARY CIRCUIT ANT ASSOCIATED SYSTEMS   • STEAM AND TURBINE FACILITY • INSTRUMENTATION AND CONTROLS • POWER SUPPLY • AUXILIARY SYSTEMS  • RADIOACTIVE WASTE DISPOSAL • RADIATION PROTECTION • SAFETY SYSTEMS • OPERATIONS • COMMITIONING • ACCIDENTANALYSES • LIMITS AND CONDITIONS OF THE SAFE OPERATION. OPERATIONAL LIMITS •  QUALITY ASSURANCE • DECOMITIONING  Project VN/RA/01, Task 3 – On-Job-Training, Safety requirements, Paris, 02 -18 June 2014

  8. SAR LOGICAL ORGANIZATION The SAR should substantiate and demonstrate that the pre-defined safety and radiation objectives are reached, accordingly to a logic and understandable path. This demonstration requires different chapters and sub-chapters, which are part of this path, be comprehensively and clearly addressed through suitable internal references. Moreover, it should be easy to understand how the objectives, the assumptions, the design principles and rules are addressed in different technical aspects and design details. Otherwise, it would be difficult verifying the completeness of the document. A very big information volume doesn’t necessarily mean good substantiation. SAFETY ANALYSIS REPORT - SPECIFICS Project VN/RA/01, Task 3 – On-Job-Training, Safety requirements, Paris, 02 -18 June 2014

  9. SAFETY REGULATIONS The demonstration provided in the SAR usually raises additional question and clarifications on the way the national and international regulations are matched in the design, General lists of regulations usually presented are not sufficient. They should be clearly referred to for each safety issue they concern, Additional clarifications and substantiations should be provided about the ranking and the priority of different regulations, national for the operator (1), the designer (2), or the producer (3), international as IAEA and EU It would be worth justifying the validity or not validity of the documentation, when originating from abroad. Moreover, the internationalization of nuclear activities requires all regulations be carefully examined and accounted for, whenever useful. SAFETY ANALYSIS REPORT - SPECIFICS Project VN/RA/01, Task 3 – On-Job-Training, Safety requirements, Paris, 02 -18 June 2014

  10. INNOVATIVE IMPROVEMENTS Usually, the operability of new and advanced technical solutions and their functional aptitude to perform their safety function are not demonstrated satisfactory. A robust and proven design requires that, for innovative improvements beyond current practices, additional information should be presented in the SAR and corresponding technical reports and publications referred as supporting documentation to the SAR, such as: theoretical analyses adopted to substantiate decisions, experimental works carried-out in on scale and / or full scale facilities and prototypes, tests results from real plants or commissioning tests, when available, licensing process documentation. These supporting documents should be made available during the evaluation process on demand. SAFETY ANALYSIS REPORT - SPECIFICS Project VN/RA/01, Task 3 – On-Job-Training, Safety requirements, Paris, 02 -18 June 2014

  11. DEFENSE IN DEPTHPRESENTATION The SAR should prove, demonstrate and provide evidence that defined safety objectives are achieved with a high level of confidence and reliability. Many different aspects have to be presented in SAR : Nuclear safety and radiation protection, mechanical, physical and thermal-hydraulic characteristics, technical and organization measures, systems, equipment, Accident analyses , Probabilistic studies, etc. For better understanding and easier assessment, all that huge information have to be systematically and logically structured and presented. Today, an international consensus exists, based and documented by many IAEA publications, that The Defense in Depth (DiD) principle should be considered as a basis for systematic safety substantiations and demonstration. The DiD presentation and demonstration needs very carful examination SAFETY ANALYSIS REPORT - SPECIFICS Project VN/RA/01, Task 3 – On-Job-Training, Safety requirements, Paris, 02 -18 June 2014

  12. DESIGN PRINCIPLES AND RULES The Defense in Depth implementation requires demonstration on how the design principles and rules are respected through adoption of design provisions and implementation of margins. Concerning systems design, for example, design rules should be presented for: Redundancy, Diversity, Single failure criterion, Common Cause Failures Passive and active safety systems combination, Common use for safety and normal operation, etc. Design requirements, principals and considerations should be presented to demonstrate safety objectives validity with important credibility in : internal hazards, and external hazards conditions. SAFETY ANALYSIS REPORT - SPECIFICS Project VN/RA/01, Task 3 – On-Job-Training, Safety requirements, Paris, 02 -18 June 2014

  13. SAFETY MARGINS To be substantiated that there is a wide margin to failure of any structures, systems or components for any of the AOO or AC that could occur. Safety margins are typically specified in codes and standards as well as by the regulatory body. The safety assessment shall determine whether acceptance criteria for each aspect of the safety analysis are such that an adequate margin is ensured. Analysis and substantiations of the safety margins has to be systematically presented. The methods adopted to evaluate and establish the safety margins have to be presented and discussed. Today safety margins in extreme external hazards (seism, flooding, high temperature etc.) should be reassessed (after Fukushima) SAFETY ANALYSIS REPORT - SPECIFICS Project VN/RA/01, Task 3 – On-Job-Training, Safety requirements, Paris, 02 -18 June 2014

  14. CODES VERIFICATION AND VALIDATION All computer codes used for the safety analysis have to be well documented in a consistent manner making reference to the objective of the code, Providing a comprehensive technical description, the validation reports, including all phases of the process (verification of the models against analytical approach, qualification through suitable benchmarks and verification against suitable and representative experimental results), The range of applicability and the scaling effects have also to be addressed, in tight connection with the adoption of uncertainty values and addition of design margins, Potential certification of the current version used. (not in all countries) SAFETY ANALYSIS REPORT - SPECIFICS Project VN/RA/01, Task 3 – On-Job-Training, Safety requirements, Paris, 02 -18 June 2014

  15. UNCERTAINTIES A systematic approach has to be adopted for the quantification of the uncertainties, their source, nature and degree. The general statements concerning uncertainties should be with sufficient justification and insight. Detailed and precise information has to be provided on the way the uncertainties affecting the main parameters are evaluated through both experiments and simulations, on the ways they are affected to the safety parameters and the procedure adopted to combine them, when needed, to obtain uncertainty figures which are not directly accessible, via experimental evidence. This issue has to be supported by a comprehensive documentation. This item applies to both uncertainties in the assumptions forming the basis of accident analyses and the implementation of uncertainties in the design in close connection with code validation, operation experience and current design practice. SAFETY ANALYSIS REPORT - SPECIFICS Project VN/RA/01, Task 3 – On-Job-Training, Safety requirements, Paris, 02 -18 June 2014

  16. PSA - PROBABILISTIC SAFETY ASSESSMENT Third generation reactors (GEN-III and GEN-III+) are considered as risk oriented designs. According to the current international practices, the PSA (Probabilistic Safety Assessment) is an important part of SAR. The IAEA safety standard No GS-G-4.1 considers the PSA as part of the SAR. The Technical Guidelines (TGL) also requires well balanced safety approach: The safety demonstration for the nuclear power plants of the next generation has to be achieved in a deterministic way, supplemented by probabilistic methods and appropriate research and development work. As part of EPR safety and design considerations, the PSA is seen as design tool to improve the overall safety level of the plant, through the implementation of design solutions and the addition of suitable safety systems. For EPR, the summary of PSA is presented in the additional chapter 18, PROBABILISTIC SAFETY ASSESSMENT, with the same relevance as chapter 15, SAFETY ANALYSES, which presents the deterministic safety approach. As for the AP 1000, this topic is addressed in chapter 19. According to Russian practice, the PSA analysis is presently provided in a separate document from SAR. SAFETY ANALYSIS REPORT - SPECIFICS Project VN/RA/01, Task 3 – On-Job-Training, Safety requirements, Paris, 02 -18 June 2014

  17. DESIGN DETAILS Obviously, it is not possible to include all the design information in the SAR. Accordingly, the material presented should be mainly safety-oriented. The volume of information and the level of the details provided should be primarily linked to the safety importance and relevance of the system or equipment under investigation. The information needed should be provided under a suitable form for SAR with essential and safety-oriented description and simplified drawings. Direct introduction of the design files is to be avoided. SAFETY ANALYSIS REPORT - SPECIFICS Project VN/RA/01, Task 3 – On-Job-Training, Safety requirements, Paris, 02 -18 June 2014

  18. CONCLUSIONS Nowadays there is an important internationalization of the nuclear activities and Vendors, Operators and Safety Authorities are looking for further harmonization of safety practices for better understanding and enhanced mastering of safety issues In particularity, additional efforts are needed to achieve a common approach between east and west practices to present different safety aspects in the plant Safety Analysis Report –SAR -, a basic document in the NPP licensing process Presented general/common requirements for PSAR should be very carefully examined in the licensing process and corrective measures could be taken if necessary Project VN/RA/01, Task 3 – On-Job-Training, Safety requirements, Paris, 02 -18 June 2014

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