• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.18 no.1
• /
• pp.1-10
• /
• 2022

Currently, there is no technical standard and regulation for seismic analysis of non-nuclear safety piping. Accordingly, ASME Sec.III ND, a standards applied to safety class 3 piping, is applied. However, the technical standard applied for other than seismic analysis is ASME B31, which leads to controversy. In this study, the feasibility of applying ASME B31E was confirmed by reviewing rulescomparing technical standards, and evaluating piping allowable stress margins. The evaluation revealed that applying ASME B31.1 as a technical standard is too conservative compared to ASME Sec.III ND. On the other hand, ASME B31E (issued at the request of the industry) clearly presents the technical standards for seismic analysis of ASME B31 piping, and shows a similar level of conservatism compared to ASME Sec.III ND. It is expected to reduce the controversy over technical standards for seismic analysis of non-nuclear safety piping by applying ASME B31E.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.15 no.2
• /
• pp.1-8
• /
• 2019

The integrity of the Reactor Pressure Vessel (RPV) is affected by the neutrons bombarding the vessel wall leading to embrittlement. This irradiation-induced embrittlement leads to reduction in the fracture toughness of RPV materials. This paper presents a comparative study of typical Optimized Power Reactor (OPR)1000 reactor pressure-temperature (P-T) limit curves using the pre-2006 American Society of Mechanical Engineers (ASME) editions used in the power plant and the current ASME edition of 2010. The current ASME Code utilizes critical reference stress intensity factor based on the lower bound of static, while the Pre-2006 ASME editions are based the critical reference stress intensity factor based on the lower bound of static, dynamic and crack arrest. Model-Based Systems Engineering approach was used to evaluate ASME Code Section XI Appendix G for generating the P-T limit curves. The results obtained from this analysis indicate decrease in conservatism in P-T limit curves constructed using the current 2017 ASME code, which can potentially increase operational flexibility and plant safety. Hence it is recommended to use ASME code edition after 2006 be used in all operating nuclear power plants (NPPs) to establish P-T limit curve.

• Proceedings of the Computational Structural Engineering Institute Conference
• /
• 1997.04a
• /
• pp.114-120
• /
• 1997

원자력공사에 관련된 장치와 부속 기기를 설계 및 제작 또는 설치하려면 ASME에서 인증하는 자격이 필요하고, 자격을 취득하거나 갱신하기 위하여 시설, 품질보증 프로그램, 설계 및 제작능력등을 ASME 위원회의 실사를 받아야 한다. 이를 위한 일련의 과정 중에서 설계능력을 증명할 수 있는 설계해석 보고서를 제출하여야 하며 이를 위하여 현대중공업이 설계한 Chiler Surge Vessel의 설계조건 및 하중 및 하중조합에 대하여 Design Report를 작성하였으며, ASME Code의 요구조건을 모두 만족하는 것으로 평가되었다.

• The KSFM Journal of Fluid Machinery
• /
• v.2 no.3 s.4
• /
• pp.24-29
• /
• 1999

The ASME strainers have been newly installed at the suction side of each reactor coolant pump to get rid of the foreign materials which may damage the pump impeller or interfere with the coolant path of fuel flow tube or primary plate type heat exchanger. The strainer was designed in accordance with ASME SEC. III, DIV. 1, Class 3 and the structural integrity was verified by seismic analysis. The screen was designed in accordance with the effective void area from the result of flow analysis for T-type strainer. After installation of the strainer, it was confirmed through the field test that the flow characteristics of primary cooling system were not adversely affected. The pressure loss coefficient was calculated by Darcy equation using the pressure difference through each strainer and the flow rate measured during the strainer performance test. And these are useful data to predict flow variations by the pressure difference.

• Proceedings of the Korean Institute of Building Construction Conference
• /
• 2015.05a
• /
• pp.9-10
• /
• 2015

Generally significant reinforcement is used in nuclear power plant structures and may cause potential problems when concrete is poured. In particular pouring concrete into structural member joint area is more difficult than other areas since the joint area is very congested due to the crossed bars and the embedded plates, The purpose of this study is to solve these problems by applying Gr.80(550MPa) shear bars to containment structures of nuclear power plant. In order to apply them to containment structures, it is necessary to change ASME-CC code (ASME Sec.III Div.2). The structural performance tests of wall & beam have been done to compare Gr.80(550Mpa) with Gr.60(420MPa) shear bars. The test results and code change proposal were presented to ASME-CC Committee last year and the discussion for code change will be expected to proceed in the near future.

• Journal of Mechanical Science and Technology
• /
• v.20 no.12
• /
• pp.2061-2078
• /
• 2006

The main purpose of this paper is to establish the high temperature structural integrity evaluating procedures for the next generation reactors, which are to be operated at over 500$^{\circ}C$ and for 60 years. To do this, comparison studies of the high temperature structural design codes and assessment procedures such as the ASME-NH (USA), RCC-MR (France), DDS (Japan), and R5 (UK) are carried out in view of the accumulated inelastic strain and the creep-fatigue damage evaluations. Also the application procedures of the ASME-NH rules with the actual thermal and structural analysis results are described in detail. To overcome the complexity and the engineering costs arising from a real application of the ASME-NH rules by hand, all the procedures established in this study such as the time-dependent primary stress limits, total accumulated creep ratcheting strain limits, and the creep-fatigue damage limits are computerized and implemented into the SIE ASME-NH program. Using this program, the selected high temperature structures subjected to two cycle types are evaluated and the parametric studies for the effects of the time step size, primary load, number of cycles, normal temperature for the creep damage evaluations and the effects of the load history on the creep ratcheting strain calculations are investigated.

• Journal of the KSME
• /
• v.20 no.6
• /
• pp.428-433
• /
• 1980

우리는 수차에 걸친 경제개발5개년 계획을 성공적으로 추진하여 오며 지금에는 공업구조의 고 도화 즉 중화학공업의 급속한 성장이 착실히 진전되어 나가고 있다. Plant산업은 기술축적이 기 반이 되어야 하는 것은 물론 양질의 기술축적 위에 완전한 품질보증체계를 갖추기 위하여는 전 사적 품질 보증체계를 갖추어야 한다. 이런 전사적 품질보증체계를 갖추기 위하여 미국기계기 술자협회(ASME)발생하는 ASME Boiler and Pressure Vessel Code에 의한 중공업부문및 원자력 발전소 부문의 설계 제조 시험에 대한 풍질보증장증(ASME Stamp)을 소개하고자 한다.

• Nuclear industry
• /
• no.9_10
• /
• pp.35-40
• /
• 1980

• Nuclear industry
• /
• no.5_6 s.21
• /
• pp.44-50
• /
• 1980

• Nuclear industry
• /
• no.7_8 s.22
• /
• pp.20-28
• /
• 1980