Volume 6: Nuclear Codes, Standards, Licensing, and Regulatory Issues最新文献

{"title":"Development of China’s Nuclear Power Standard System and Brief Comparison With American and French Standards","authors":"Ruiyuan Deng, Feifei Wu, Zhenbang Dong, Shangyuan Liu, Jianren Du, Hongwei Zhang, Yecong Sun, Xueyao Zhang","doi":"10.1115/icone29-93863","DOIUrl":"https://doi.org/10.1115/icone29-93863","url":null,"abstract":"\u0000 China’s nuclear power standards have generally experienced three stages: starting, moderate development and rapid development. Up to now, a PWR nuclear power standard system with clear hierarchy, complete and reasonable structure and meeting the needs of the industry has been basically formed. This paper analyzes the conception, planning principles, main scope and objects, hierarchical structure, project setting and evolution process of China’s nuclear power standard system. Finally, it briefly analyzes the characteristics and current situation of nuclear power standards in the United States and France, and expounds the differences in standardization mechanism and standard technology between the two countries and China.","PeriodicalId":220265,"journal":{"name":"Volume 6: Nuclear Codes, Standards, Licensing, and Regulatory Issues","volume":"46 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125323822","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Application of ALARP Concept in Nuclear Power Plant Design — A Case Study in HPR1000","authors":"Dongyuan Li, Shi-hong Qing, Haiying Luo, Jinkai Wang, Jiaoshen Xu","doi":"10.1115/icone29-92241","DOIUrl":"https://doi.org/10.1115/icone29-92241","url":null,"abstract":"\u0000 As Low As Reasonably Practicable (ALARP) is one of the most important nuclear regulatory concepts in the UK Generic Design Assessment (GDA). ALARP requires that nuclear power plants have the duty to achieve the lowest level of residual risk provided this is reasonably practicable. Unlike As Low As Reasonably Achievable (ALARA), ALARP concept should be considered not only in radioprotection, but also in nuclear safety areas such as fault studies, structural integrity, etc. Different from the traditional licensing requirements in other countries, ALARP concept is not prescriptive but emphasizes self-evaluation, risk-orientation and balance of all aspects. It is therefore difficult to demonstrate that the design has reduced nuclear risks to an ALARP level without a systematic approach during the licensing. In this study, the UK context regulatory requirements of ALARP (including the Tolerability of Risk, Relevant Good Practice, Gross Disproportion, and Numerical Targets) are investigated based on which an ALARP approach for Hua-long Pressurized Reactor (HPR1000) is developed. In this approach, potential improvements can be identified through the review against Relevant Good Practice and risk insights. Risks can be reduced through the implementation of preferred options determined by optioneering process. The ALARP position of the whole plant design can be evaluated through holistic risk review. The ALARP approach has been successfully applied to HPR1000 and recognized by the UK authority during the GDA process. The ALARP approach facilitates the balance of safety, economy and environmental friendliness during nuclear power plant design, and can be guidance to the robust licensing and regulation.","PeriodicalId":220265,"journal":{"name":"Volume 6: Nuclear Codes, Standards, Licensing, and Regulatory Issues","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"130683244","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Research on the Construction of the Advanced Nuclear Power Standards System Platform","authors":"Shaolin Zhang, Wen Yang, Ruilin Dong, Yun Liang, Shangyuan Liu","doi":"10.1115/icone29-91362","DOIUrl":"https://doi.org/10.1115/icone29-91362","url":null,"abstract":"\u0000 Advanced nuclear power standards system is a new type of nuclear power standards system composed of all kinds of standards in the nuclear energy field.\u0000 At present, the research results of the two stages of the subject and the achievements of nuclear power standard construction in recent years are scattered. It is necessary to establish a single nuclear power standard database to process and develop the standards in the field of nuclear power and various research results, reports, process documents, data and other databases, so as to provide a good data environment for the majority of nuclear engineering scientists and technicians.\u0000 In order to solidify the research results of advanced nuclear power standard system and nuclear power standard data, the application support platform of nuclear power standard system adopts innovative SaaS cloud service mode to build an Internet-based cloud service platform with process engine and knowledge map technology.\u0000 Through the integration of informatization and standardization, standardize the standardization process, we can provide convenient and fast information services, vigorously expand the scope of participants in nuclear power standardization and the depth of participation of all parties, and realize the standard system management mode of autonomy, industry participation and dynamic updating.","PeriodicalId":220265,"journal":{"name":"Volume 6: Nuclear Codes, Standards, Licensing, and Regulatory Issues","volume":"84 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126205134","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Analysis and Research of Code Revision for Design of Nuclear Safety Related Concrete Structures of Pressurized Water Reactor Nuclear Power Plants in China","authors":"Jun Huang, Weiguo Zhang, Xueyuan Liang, Feifei Wu, Jing Xu, Shaolin Zhang, Wen Yang, Zhiyao Tang","doi":"10.1115/icone29-93834","DOIUrl":"https://doi.org/10.1115/icone29-93834","url":null,"abstract":"\u0000 NB/T 20012-2010 design requirements for concrete structures related to nuclear safety of PWR nuclear power plants is a standard issued by the National Energy Administration in 2010. It is the top-level standard and important standard for concrete structure design. Due to technological progress and changes in relevant reference standards, some of its contents can not meet the requirements at this stage, On the basis of referring to the latest relevant standards and specifications at home and abroad, the engineering experience in the structural design of nuclear safety related plants of nuclear power plants built and under construction in China was summarized, and the chief editor revised it. This paper mainly analyzes the revised contents, mainly including revising the minimum strength grade of concrete was changed to C30, HRB500, HPB300 steel bars were added, and HPB235, HRB335 steel bars were canceled; improving the safety of the axial compression members that configure the ordinary stirrups, modifying the formula for calculating the bearing capacity of the transverse shear (that is, out-of-plane shear), modifying the formula for calculating the punching bearing capacity (considering the effects of membrane force) ; defining the load effect combination used in the calculation of crack width, and the maximum crack width is calculated according to the relevant specifications of GB 50010; stipulating that short-term stiffness is taken in deflection calculation under normal operation and severe environment, normal operation and extreme environment and abnormal operation conditions; adding contents of structural analysis, basic provisions of mechanism components and beyond the scope of design basis;The Appendix B (informative) “Performance Index of Reinforcement and Concrete at High Temperature” was added;.","PeriodicalId":220265,"journal":{"name":"Volume 6: Nuclear Codes, Standards, Licensing, and Regulatory Issues","volume":"7 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125271603","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"ISO 19443: Fundamental Requirements of the New Quality Management Standard for the International Nuclear Supply Chain","authors":"Jeanne Bargsten, Clemens Treier, Özlem Yilmaz, Amandine Pades, Andrea Kim, Jung-Kyu Lee","doi":"10.1115/icone29-91803","DOIUrl":"https://doi.org/10.1115/icone29-91803","url":null,"abstract":"\u0000 The new quality management standard ISO 19443 [2] defines specific requirements for supply chain organizations in the nuclear industry. The standard became increasingly popular in the last years and a certification is now often a prerequisite for equipment manufacturers and service providers to participate in the bidding process of new build projects.\u0000 These companies face the challenge, to adapt their existing quality management system to the new requirements of ISO 19443 [2], since this is necessary to receive a globally accepted accredited certification.\u0000 A conducted analysis of the ISO 19443 standard [2] identified major differences to the established quality management system standard ISO 9001 [3], the results are presented. Furthermore, the key requirement areas of the standard with particular significance to the nuclear industry, are evaluated.\u0000 ISO 19443 [2] enshrines nuclear safety culture as the fundamental element throughout the entire supply chain and enables a graded approach to the application of quality requirements. The standard aims at the global harmonization of management standards and facilitates the regulation of the international nuclear supply chain.\u0000 Best practice approaches are highlighted, on how to successfully adapt a quality management system. It is based on experience of one of the first accredited inspection bodies, already supporting several manufacturing companies on their way to certification.","PeriodicalId":220265,"journal":{"name":"Volume 6: Nuclear Codes, Standards, Licensing, and Regulatory Issues","volume":"4 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"123807126","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Overview of AFCEN Non-Linear Benchmark in China – Improving Rules for Vessel Ratcheting Evaluation","authors":"Yu Tan, Xuejiao Shao, Enming Liang, Shuchun Zuo, Feng Lv, T. Métais, Han Liu","doi":"10.1115/icone29-93561","DOIUrl":"https://doi.org/10.1115/icone29-93561","url":null,"abstract":"\u0000 The French nuclear standard, RCC-M 2017 edition, has seen the inclusion of an Appendix on non-linear calculation methods: the Appendix ZC. This non-mandatory Appendix encompasses a set of rules and guidelines which can help the designer perform non-linear analyses. It includes the definition of models adequate to use, fatigue Ke factor optimization methods, etc...\u0000 In fatigue analyses per the RCC-M B-3200 paragraph, rules on thermal ratcheting limits are exposed in paragraph B-3234.8. These rules aim at limiting any thermal ratcheting of the structure and the criteria exposed are sometimes difficult to meet, even though there is no threat to safe operation of the concerned equipment.\u0000 Non-linear analyses in this case could be used to help meet the criteria but would need extra rules to be defined: these are considered to be added to the Appendix ZC.\u0000 In early 2019, an international benchmark effort was launched on this topic. The case is a typical nozzle geometry subjected to both thermal and pressure transients.\u0000 In China, some main nuclear entities have taken part in this benchmark effort: NPIC, SNPI, CNPDC and CNPE, along with the coordination of EDF and Yuansuan.\u0000 • let participants get familiar with the benchmark and performing linear elastic calculations along with simplified RCC-M code evaluation methods. The outcome is a participant comparison of the respect or not of the code criteria.\u0000 • perform an analysis based first on one’s own parameter identification and an Armstrong-Frederick model (with 1 and 2 parameters) and then perform the comparison with a common set of parameters.\u0000 • use a defined cyclic stress-strain curve including full hysteresis loops to derive a common identification methodology and compare results.\u0000 This paper aims at providing the public with a progress of the benchmark case (using ANSYS and code_aster software), the Chinese participants findings, and recommendations for the structural integrity community on the first two bullet points above.","PeriodicalId":220265,"journal":{"name":"Volume 6: Nuclear Codes, Standards, Licensing, and Regulatory Issues","volume":"os-43 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127784020","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Analysis of Different Requirements of Dissimilar Metal Joint Welding Procedure Qualification Between RCC-M and ASME BPVC","authors":"W. Pei, Chen Dongxu, Yang Shumin","doi":"10.1115/icone29-91363","DOIUrl":"https://doi.org/10.1115/icone29-91363","url":null,"abstract":"\u0000 In order to enhance the comprehension of the welding procedure qualification standards, RCC-M (Design and Construction Rules for Mechanical Components of PWR Nuclear Islands) and ASME BPVC (American Society of Mechanical Engineers Boiler and Pressure Vessel Code), used by the components fabrication for nuclear power industry in the People’s Republic of China, this paper takes the qualification of dissimilar metal joint between inlet-nozzle and safe end of steam generator in Generation III nuclear power unit as an example, taking a comparative approach, compares the requirements of different welding variables one by one between two standards and then lists the differences and analyzes the differences. It shows that RCC-M has more strict requirements than ASME BPVC on the qualification of dissimilar metal joint with buttering, the type of test couple qualified by RCC-M is more similar to product than that by ASME BPVC, and qualification by RCC-M may need more test couples than that by ASME BPVC, if choosing same value of variables which have overlap between these two standards, test coupon qualified by RCC-M can used for that qualified by ASME BPVC, but not vice versa. At last, the paper gives some implementation advises and guidance of subsequent qualification works during design and fabrication.","PeriodicalId":220265,"journal":{"name":"Volume 6: Nuclear Codes, Standards, Licensing, and Regulatory Issues","volume":"97 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128695848","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Codes Comparative Study on Mechanical Properties of Core Forgings of Reactor Pressure Vessel","authors":"Hao Yu, Yinbiao He, Xinxin Wu, Shenjie Gu, Xin Liu, Hongchang Wang","doi":"10.1115/icone29-90363","DOIUrl":"https://doi.org/10.1115/icone29-90363","url":null,"abstract":"\u0000 In this paper, the mechanical properties of core forgings in Pressurized Water Reactor (PWR )NPPs in China were studied according to different rules of minimum yield strength at room temperature stipulated in ASME codes and RCC-M codes. First, this paper investigates the origin of two different codes for core forgings, then analyses the effects of yield strength, tensile strength, tempering temperature and toughness on the mechanical properties of forgings in both of codes. And then the results showed that the acceptance value of minimum yield strength at room temperature materials does not correspond to the mechanical design and analysis value. Finally, some suggestions for compiling the c’ore area forgings mechanical properties codes suitable for China’s PWR NPPs were given. This work provided a feasible implementation path for guiding the formulation of China’s PWR NPP codes.","PeriodicalId":220265,"journal":{"name":"Volume 6: Nuclear Codes, Standards, Licensing, and Regulatory Issues","volume":"32 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125917871","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"On Multiaxial Strain Characteristics of Piping Systems Subjected to Earthquake Loads","authors":"M. Morishita, Akihito Otani","doi":"10.1115/icone29-90172","DOIUrl":"https://doi.org/10.1115/icone29-90172","url":null,"abstract":"\u0000 The JSME Code Case N-CC-008 provides design rules for piping seismic design by inelastic response time history analysis and strain-based fatigue criteria. Since the strain time histories obtained from the analysis are multi-axial and random, the Code Case uses the Rainflow method for cycle counting in its procedure of evaluating the ranges of equivalent strain and the fatigue usage factor. The Rainflow method is used for identifying pairs of peaks/valleys of the representative strain and their occurrence times. Then, with the other components of strain and stress at these times, the range of equivalent strain is calculated sequentially. On the other hand, ASME Boiler and Pressure Vessel Code Section VIII, Division 2 reads “The Rainflow cycle counting method is recommended but not applicable for non-proportional loading.” The authors, then, analyzed the behavior of multi-axial strain from the inelastic response analyses of some dynamic tests of piping systems to test if the piping strains due to seismic excitations are proportional or non-proportional, or somewhere in between. The IS method, which was proposed by Itoh et.al for quantifying the degree of non-proportionality of multi-axial fatigue cycles, was used for the analysis. The analysis results showed that the multi-axial strain in piping systems induced by seismic excitation is almost proportional. Based on this result, the authors concluded that the Rainflow cycle counting method is well applicable for the piping seismic fatigue evaluation. In addition to these numerical analyses with the IS method, some discussions are also made in this paper from the viewpoint of the relation between vibration mode and strain components (compatibility condition between deformation and strain) to support the conclusion that the piping strain by seismic loads is proportional.","PeriodicalId":220265,"journal":{"name":"Volume 6: Nuclear Codes, Standards, Licensing, and Regulatory Issues","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129687476","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Practice and Discussion on Safety Categorisation and Classification of Defence in Depth Level 1/2 Functions Based on IAEA-SSG-30","authors":"Fan He, Jiangwei Ji, Jiaoshen Xu, Zengpei Zhu, Kewei Cai","doi":"10.1115/icone29-93238","DOIUrl":"https://doi.org/10.1115/icone29-93238","url":null,"abstract":"\u0000 The safety classification of structures, systems and components (SSC) in nuclear power plants (SSG-30) was issued in 2014 by the International Atomic Energy Agency (IAEA). Both the Hua-long Pressurized Reactor (HPR1000) and UK version of the Hua-long Pressurized Reactor (UK HPR1000) Safety Function Categorisation and System, Structure and Component (SSC) Classification methodology have been developed based on IAEA SSG-30 with consideration of different regulatory requirement. The SSG-30 is a high level guidance to perform the safety categorization and classification and there are still some issues that need to be analyzed in specific application such as failure consequence severity criteria and consequence analysis hypothesis. With the expansion of Defence in Depth (DiD) levels in nuclear power plant design, there are differences about function categorization requirements between different countries. This paper analyzes the differences of definition about DiD levels and the function categorisation requirements in different international standards. The IAEA SSG-30 methodology for application is evaluated to meet different regulatory requirements and the alignment or expansion is needed for application have been identified to determine the function categorisation in DID 1/2 levels. The principles of the HPR1000 and UK HPR1000 on DID 1/2 functions categorisation have been summarized and the subsequent introduce and discussion are about practices to apply these principles on special safety functions and safety features.","PeriodicalId":220265,"journal":{"name":"Volume 6: Nuclear Codes, Standards, Licensing, and Regulatory Issues","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115962690","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}