Flaw Evaluation and Allowable Flaw Sizes of Nuclear Piping in JSME Code

Author(s):  
Kunio Hasegawa ◽  
Hideo Kobayashi ◽  
Koichi Kashima

A flaw evaluation code for nuclear power plants has been developed at the Japan Society of Mechanical Engineers (JSME) in 2000 and revised adding inspection rules in 2002. Then the code consists of inspection for nuclear components and evaluation procedures of flaws in Class 1 components detected during in-service inspection. This paper introduces the summary of the JSME Code and describes two kinds of allowable flaw sizes, Acceptance Standards and Acceptance Criteria, for Class 1 pipes in the flaw evaluation procedures. In addition, these allowable flaws are compared with those in the ASME (American Society of Mechanical Engineers) Code Section XI.


Author(s):  
Koichi Kashima ◽  
Tomonori Nomura ◽  
Koji Koyama

JSME (Japan Society of Mechanical Engineers) published the first edition of a FFS (Fitness-for-Service) Code for nuclear power plants in May 2000, which provided rules on flaw evaluation for Class 1 pressure vessels and piping, referring to the ASME Code Section XI. The second edition of the FFS Code was published in October 2002, to include rules on in-service inspection. Individual inspection rules were prescribed for specific structures, such as the Core Shroud and Shroud Support for BWR plants, in consideration of aging degradation by Stress Corrosion Cracking (SCC). Furthermore, JSME established the third edition of the FFS Code in December 2004, which was published in April 2005, and it included requirements on repair and replacement methods and extended the scope of specific inspection rules for structures other than the BWR Core Shroud and Shroud Support. Along with the efforts of the JSME on the development of the FFS Code, Nuclear and Industrial Safety Agency, the Japanese regulatory agency approved and endorsed the 2000 and 2002 editions of the FFS Code as the national rule, which has been in effect since October 2003. The endorsement for the 2004 edition of the FFS Code is now in the review process.



Author(s):  
Koichi Kashima ◽  
Tomonori Nomura ◽  
Koji Koyama

Following a recognition of the need to establish a FFS (Fitness-for-Service) Code in Japan, JSME (Japan Society of Mechanical Engineers) published its first edition in May 2000, which provided rules on flaw evaluation for Class 1 pressure vessels and piping, referring to the ASME Code Section XI. The second edition of the FFS Code was published in October 2002, to include rules on in-service inspection, which also referred to the ASME Code Section XI incorporating independent Japanese concepts. In addition, individual inspection rules for specific structures, such as shroud and shroud support for BWR plants, were prescribed in consideration of aging degradation by SCC. Furthermore, the third edition, which includes requirements on repair and replacement methods, will be published in 2004. Along with the efforts of the JSME on the preparation of the FFS Code, the Japanese Regulatory Agency has approved and endorsed this Code as the national rule, which has been in effect since October 2003.



Author(s):  
Ronald C. Lippy

The nuclear industry is preparing for the licensing and construction of new nuclear power plants in the United States. Several new designs have been developed and approved, including the “traditional” reactor designs, the passive safe shutdown designs and the small modular reactors (SMRs). The American Society of Mechanical Engineers (ASME) provides specific Codes used to perform preservice inspection/testing and inservice inspection/testing for many of the components used in the new reactor designs. The U.S. Nuclear Regulatory Commission (NRC) reviews information provided by applicants related to inservice testing (IST) programs for Design Certifications and Combined Licenses (COLs) under Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” in Title 10 of the Code of Federal Regulations (10 CFR Part 52) (Reference 1). The 2012 Edition of the ASME OM Code defines a post-2000 plant as a nuclear power plant that was issued (or will be issued) its construction permit, or combined license for construction and operation, by the applicable regulatory authority on or following January 1, 2000. The New Reactors OM Code (NROMC) Task Group (TG) of the ASME Code for Operation and Maintenance of Nuclear Power Plants (NROMC TG) is assigned the task of ensuring that the preservice testing (PST) and IST provisions in the ASME OM Code to address pumps, valves, and dynamic restraints (snubbers) in post-2000 nuclear power plants are adequate to provide reasonable assurance that the components will operate as needed when called upon. Currently, the NROMC TG is preparing proposed guidance for the treatment of active pumps, valves, and dynamic restraints with high safety significance in non-safety systems in passive post-2000 reactors including SMRs.



Author(s):  
Thomas G. Scarbrough

In a series of Commission papers, the U.S. Nuclear Regulatory Commission (NRC) described its policy for inservice testing (IST) programs to be developed and implemented at nuclear power plants licensed under 10 CFR Part 52. This paper discusses the expectations for IST programs based on those Commission policy papers as applied in the NRC staff review of combined license (COL) applications for new reactors. For example, the design and qualification of pumps, valves, and dynamic restraints through implementation of American Society of Mechanical Engineers (ASME) Standard QME-1-2007, “Qualification of Active Mechanical Equipment Used in Nuclear Power Plants,” as accepted in NRC Regulatory Guide (RG) 1.100 (Revision 3), “Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants,” will enable IST activities to assess the operational readiness of those components to perform their intended functions. ASME has updated the Operation and Maintenance of Nuclear Power Plants (OM Code) to improve the IST provisions for pumps, valves, and dynamic restraints that are incorporated by reference in the NRC regulations with applicable conditions. In addition, lessons learned from performance experience and testing of motor-operated valves (MOVs) will be implemented as part of the IST programs together with application of those lessons learned to other power-operated valves (POVs). Licensee programs for the Regulatory Treatment of Non-Safety Systems (RTNSS) will be implemented for components in active nonsafety-related systems that are the first line of defense in new reactors that rely on passive systems to provide reactor core and containment cooling in the event of a plant transient. This paper also discusses the overlapping testing provisions specified in ASME Standard QME-1-2007; plant-specific inspections, tests, analyses, and acceptance criteria; the applicable ASME OM Code as incorporated by reference in the NRC regulations; specific license conditions; and Initial Test Programs as described in the final safety analysis report and applicable RGs. Paper published with permission.



Author(s):  
Alexander Mutz ◽  
Manfred Schaaf

Abstract The Nuclear Power Plant KKG in Gösgen, Switzerland was designed according to the ASME Boiler and Pressure Vessel Code. The ASME BPVC, Section III, Appendix 11 regulates the flange calculation for class 2 and 3 components, it is also used for class 1 flanges. A standard for the determination of the required gasket characteristics is not well established which leads to a lack of clarity. As a hint different y and m values for different kinds of gasket are invented in ASME BPVC Section III [1]. The KTA 3201.2[2] and KTA 3211.2[3] regulate the calculation of bolted flanged joints in German nuclear power plants. The gasket characteristics required for these calculation methods are based on DIN 28090-1[4], they can be determined experimentally. In Europe, the calculation code EN 1591-1 [5] and the gasket characteristics according to EN 13555[6] are used for flange calculations. Because these calculation algorithms provide not only a stress analysis but also a tightness proof, it would be preferable to use them also in the NPP’s in Switzerland. Additionally, for regulatory approval also the requirements of the ASME BPVC must be fullfilled. For determining the bolting up torque moment of flanges several tables for different nominal diameters of flanges using different gaskets and different combinations of bolt and flange material were established. As leading criteria for an allowable state, the gasket surface pressure, the allowable elastic stress of the bolts and the strain in the flange should be a good and conservative basis for determining allowable torque moments. The herein established tables show only a small part according to a previous paper [7] where different calculation methods for determining bolting up moments were compared to each other. In this paper the bolting-up torque moments determined with the European standard EN 1591-1 for the flange, are assessed on the strain-based acceptance criteria in ASME BPVC, Section III, Appendices EE and FF. The assessment of the torque moment of the bolts remains elastically which should lead to a more conservative insight of the behavior of the flanges.





Author(s):  
Claude Faidy

Two major Codes are used for Fitness for Service of Nuclear Power Plants: one is the ASME B&PV Code Section XI and the other one is the French RSE-M Code. Both of them are largely used in many countries, partially or totally. The last 2013 RSE-M covers “Mechanical Components of Pressurized Water Reactors (PWRs): - Pre-service and In-service inspection - Surveillance in operation or during shutdown - Flaw evaluation - Repairs-Replacements parts for plant in operation - Pressure tests The last 2013 ASME Section XI covers “Mechanical components and containment of Light Water Reactors (LWRs)” and has a larger scope with similar topics: more types of plants (PWR and Boiling Water Reactor-BWR), other components like metallic and concrete containments… The paper is a first comparison covering the scope, the jurisdiction, the general organization of each section, the major principles to develop In Service Inspection, Repair-Replacement activities, the flaw evaluation rules, the pressure test requirements, the surveillance procedures (monitoring…) and the connections with Design Codes… These Codes are extremely important for In-service inspection programs in particular and essential tools to justify long term operation of Nuclear Power Plants.



Author(s):  
Gurjendra S. Bedi

The U.S. Nuclear Regulatory Commission (NRC) staff issued Revision 2 to NUREG-1482, “Guidelines for Inservice Testing at Nuclear Power Plant,” to assist the nuclear power plant licensees in establishing a basic understanding of the regulatory basis for pump and valve inservice testing (IST) programs and dynamic restraints (snubbers) inservice examination and testing programs. Since the Revision 1 issuance of NUREG-1482, certain tests and measurements required by earlier editions and addenda of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) have been clarified, updated, revised or eliminated. The revision to NUREG-1482 incorporates and addresses those changes, and includes the IST programs guidelines related to new reactors. The revised guidance incorporates lessons learned and experience gained since the last issue. This paper provides an overview of the contents of the NUREG-1482 and those changes and discusses how they affect NRC guidance on implementing pump and valve inservice testing (IST) programs. For the first time, this revision added dynamic restraint (snubber) inservice examination and testing program guidelines along with pump and valve IST programs. This paper highlights important changes to NUREG-1482, but is not intended to provide a complete record of all changes to the document. The NRC intends to continue to develop and improve its guidance on IST methods through active participation in the ASME OM Code consensus process, interactions with various technical organizations, user groups, and through periodic updates of NRC-published guidance and issuance of generic communications as the need arises. Revision 2 to NUREG-1482 incorporates regulatory guidance applicable to the 2004 Edition including 2005 and 2006 Addenda to the ASME OM Code. Revision 0 and Revision 1 to NUREG-1482 are still valid and may continue to be used by those licensees who have not been required to update their IST program to the 2004 Edition including the 2005 and 2006 Addenda (or later Edition) of the ASME OM Code. The guidance provided in many sections herein may be used for requesting relief from or alternatives to ASME OM Code requirements. However, licensees may also request relief or authorization of an alternative that is not in conformance with the guidance. In evaluating such requested relief or alternatives, the NRC uses the guidelines/recommendations of the NUREG, where applicable. The guidelines and recommendations provided in this NUREG and its Appendix A do not supersede the regulatory requirements specified in Title 10 of the Code of Federal Regulations (10 CFR) 10 CFR 50.55a, “Codes and standards”. Further, this NUREG does not authorize the use of alternatives to, grant relief from, the ASME OM Code requirements for inservice testing of pumps and valves, or inservice examination and testing of dynamic restraints (snubbers), incorporated by reference in 10 CFR 50.55a. Paper published with permission.



2020 ◽  
Vol 1 (46) ◽  
pp. 387-404
Author(s):  
Kharytonova L ◽  
◽  
Kutsenko O ◽  
Kadenko I ◽  
◽  
...  

The paper focuses on the one of the persperctive approaches to the increasing of thje safety of Nuclear Power Plants - Flaw Handbook Concept. Object of study - equipment and piping of Nuclear Power Plants. Purpose of study - the description of the Flaw Handbook Concept and the application of the concept for the specific example. Method of the study - numerical procedures of the finite-element method and fracture mechanics. In the modern economics the optimization of the performance and operation of industry and power systems is of the main importance. The Flaw Handbook Concept is considered in the paper. This concept is applied on the nuclear power plants in the leading states with the aim of the optimization of the procedures of in-service inspection and repair. The main steps of the concept are considered and applied for the specific example. An example of Flaw Handbook using is analysed. The results of the paper can be incorporated into the procedures of in-service inspection for the safety-significant equipment and piping. KEYWORDS: FLAW HANDBOOK, BRITTLE FRACTURE, FATIGUE, FINITE-ELEMENT METHOD, SURGE PIPE.



Author(s):  
Gaetano Ruggeri ◽  
Luigi Brusa

Abstract Scope of the paper is to summarise the experience about management of materials arising from decommissioning of Italian NPPs, and to illustrate criteria, procedures and systems, which Sogin is defining to manage the problem of the clearance of sites and materials, considering the international experience and boundary conditions existing in the Country. Since 1962 Enel (the largest Italian utility for electric power) has operated the four Italian nuclear power plants: Garigliano (160 MWe BWR), Latina (210 MWe GCR), Trino (270 MWe PWR) and Caorso (882 MWe BWR). These NPPs were shutdown in the 80’s: Garigliano NPP was shutdown in 1982 following a decision made by Enel, based on technical and economical reasons, Latina, Trino and Caorso NPPs following decisions made by the Italian Government after the Chernobyl accident. The “deferred decommissioning (SAFSTOR)” was the decommissioning strategy selected by Enel and approved by the competent Authorities, due to the lack of a repository for the disposal of radioactive materials and of release limits for clearance of materials. Activities have been started aimed at reaching the “Safe Enclosure” condition, which would have lasted for some decades, before final dismantling of plants. In 1999 the liberalisation of the Italian electricity market led Enel to separate its nuclear activities, forming a new Company, named Sogin, to which decommissioning Italian NPPs was committed. At the same time, considering pressures, both at national and local level, to adopt the “prompt decommissioning (DECON)” strategy, in December 1999 the Italian Minister of Industry, with the intent to accelerate the dismantling of Italian NPPs, presented the plans to create a national repository for nuclear waste, and asked Sogin to revise the decommissioning plans, according to the new global strategy, taking into account all the relevant technical, organisational, financial and legislative aspects of the problem. As the DECON strategy enhances the importance of “clean-up” both of sites and materials, the related aspects are held in due consideration in developing the decommissioning plans, which deal with the following: • characterisation of plant systems, components and structures; • decontamination and dismantling techniques; • monitoring of dismantled materials for clearance; • treatment of dismantled, radioactive materials (which cannot be cleared), prior to disposal; • treatment and conditioning of radioactive waste, prior to disposal; • final clearance of sites. Authorisation requirement related to the release, recycle and reuse of materials produced during plant decommissioning, together with the acceptance criteria for disposal of radioactive materials, are of key importance, considering that the change in decommissioning strategy increases the quantity of radioactive waste to be disposed of, the costs for waste treatment, transportation and disposal, and the capacity of the national repository. In this connection, Sogin is discussing with competent Authorities and Bodies in order to define clearance criteria and disposal acceptance criteria, which neither impair nor complicate the future dismantling operations. In (1) details are given about Italian decommissioning Regulation, decommissioning strategy and Organisation, in order to show the boundary conditions, which exist in Italy for planning and development of NPPs Decommissioning Projects. In the following paragraphs the decommissioning strategy is summarised first together with some critical items of decommissioning; then the Italian regulation about the management of radioactive waste is reported. The management of waste and materials, which will arise from the decommissioning of Italian nuclear power plants, is driven by the requirements imposed by the competent Authorities basing on this regulation.



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