• Journal of Energy Engineering
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• v.25 no.1
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• pp.15-22
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• 2016

The heat transport system transients of Wolsong 2,3,4 nuclear power plants were analysed during abnormal operating conditions. The compliance with requirements of AECB Regulatory Document R-77 for CANDU reactor was estimated. The analysis results showed that for each postulated accident the peak pressure values in the reactor headers are within the acceptance criteria given in ASME code requirements. The effect of LRV that is one of the overpressure protection device was very minor.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.995-1000
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• 1995

ASME B&PV Code Section III NB-3200의 규정은 원자로 관련 1등급 부품의 설계시 지켜야할 사항이다. 이 규정은 운전조건별로 허용응력에 대한 분류를 하여 허용한도를 규정하고 있다. 따라서 응력해석시 이 규정을 적용하기 위해 해석결과의 검색, 추출정리, 추가계산 등 응력해석 후속작업을 위한 통합 program을 awk 언어를 사용하여 개발하였다. 이 통합 Program은 ASME에 규정된 응력별로 여러 개의 awk program module로 작성하였고 각각의 모듈을 통합하는 UNIX script file로 구성되어있다. 각각의 모듈은 독립된 batch 작업이 가능하고, 이것을 모두 연계한 batch 작업 역시 가능하도록 하였다. 문서작성시 도표작성을 용이하게 하기 위해 후처리결과가 하나의 디렉토리에 저장되도록 하였다.

• Nuclear industry
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• no.8_9
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• pp.48-50
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• 1979

• Nuclear industry
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• no.5_6 s.26
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• pp.4-15
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• 1981

• Corrosion Science and Technology
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• v.3 no.5
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• pp.216-221
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• 2004

The operating experience showed that the fatigue is one of the major piping failure mechanisms in nuclear power plants (NPPs). The pressure and/or temperature loading transients, the vibration, and the mechanical cyclic loading during the plant operation may induce the fatigue failure in the nuclear piping. Recently, many fatigue piping failure occurred at the socket weld area have been widely reported. Many failure cases showed that the gap requirement between the pipe and fitting in the socket weld was not satisfied though the ASME Code Sec. III requires 1/16 inch gap in the socket weld. The ASME Code OM also limits the vibration level of the piping system, but some failure cases showed the limitation was not satisfied during the plant operation. In this paper, the fatigue behavior of the socket weld in the nuclear piping was estimated by using the three dimensional finite element method. The results are as follows. (1) The socket weld is susceptible to the vibration if the vibration levels exceed the requirement in the ASME Code OM. (2) The effect of the pressure or temperature transient load on the socket weld in NPPs is not significant because of the very low frequency of the transient during the plant lifetime operation. (3) 'No gap' is very risky to the socket weld integrity for the specific systems having the vibration condition to exceed the requirement in the ASME OM Code and/or the transient loading condition. (4) The reduction of the weld leg size from $1.09*t_1$ to $0.75*t_1$ can affect severely on the socket weld integrity.

• Proceedings of the KSME Conference
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• 2003.04a
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• pp.654-659
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• 2003

PVAP (Pressure Vessel Analysis Program V1.0) was developed by adopting the finite element analysis program ANSYS V6.0, and Microsoft Visual Basic V6.0 was also utilized for the interfacing and handling of input and output data during the analysis. PVAP offers the end user the ability to design and analyze vessels in strict accordance with ASME Section VIII, Division 2. More importantly, the user is not required to make any design decisions during the input of the vessel. PVAP consists of three analysis modules for the finite element analysis of the primary components of pressure vessel such as head, shell, nozzle, and skirt. In each module, finite element analysis can be performed automatically only if the end user gives the dimension of the vessel. Furthermore, the calculated results are compared and evaluated in accordance with the criteria given in ASME Boiler and Pressure Vessel Code, Section VIII, Division 2. In particular, heat transfer analysis and consecutive thermal stress analysis for the junction between skirt and head can be carried out automatically in the skirt-tohead module. Finally, report including the above results is created automatically in Microsoft Word format.

• Journal of the KSME
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• v.18 no.4
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• pp.35-40
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• 1978

여기서 상기 ASME Code에 대하여 간단히 설명하기로 한다. ASME Code 는 첫부분에서 ASME Code의 적용을 받아야 하는 압력용기를 정의하고, 압력용기의 건설에 관한 일반원칙을 설명한후 그 다음에는 세개의 Subsection으로 나뉘어져 있다. 즉 Subsection A General Requirements Subsection B Requirements Pertaining to Methods of Fabrication of Pressure Vessels Subsection C Requirements Pertaining to Classes of Material 여기서 Subsection A는 압력용기 재료나 제작방법의 상위와 관계없이 적용하여야 할 일반적인 요구사항을 규정한 것이며, Subsedction B에서는 압력용기의 제작방법을 용접,리벳팅,단조,경납 땜의 4가지로 나누어 각 제작방법에 따른 특수 요구사항을 규정하였고, 마지막으로 Subsection C 는 재료에 따른 특수 요구사항을 규정한 것이다. 이 각 Subsection은 다시 General, Materials, Design, Fabrication, Inspection and Tests, Stamping and Reports, Pressure Relief Devices로 나누어 이에 대한 각각의 요구사항들을 설명하고 있다. 그러나 이 기술기준에서는 제정방향으로, 다음의 목차에서도 알 수 있는 바와 같이 이들의 순서를 바꾸어 총칙,재료,설계,제작, 검사 및 시험, 압력릴리프장치를 배치한 후 이미 KS B 6231에 제정되어 있는 것은 그 규정을 대부분 그 대로 인용하였고, 그렇지않은 것은 우리의 실정을 참작하여 삭제, 보완, 수정하였다. 삭제한 내용 중 대표적인 것으로 공인검사관(Authorized Inspector) 및 Stamping and Reports 에는 9개의 Mandatory Appendix 와 16개의 Nonmandatory Appendix가 있는데, 이둘 중 이 기술기준에서 필요하다고 생각되는 것은 발췌 수록하였다. 단위에 대해서는 국가시책에 따라 메트릭 시스템을 사용하였고 단위의 환산에서 야기되는 소수점등의 처리는 공학적인 판단에 의거하였다.

• Journal of the Korean Society for Nondestructive Testing
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• v.30 no.4
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• pp.380-385
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• 2010

The integrity of components and piping of operating nuclear power plants has been identified by in-service inspection(ISI) requirements and activities commensurate with standards and codes such as KEPIC MI or ASME Code Section XI. However, the other various degradation mechanisms not considered during design stage of nuclear power plants have been checked by enhanced ISI. The requirements of enhanced ISI have been voluntarily developed by the industry itself or strickly issued by regulatory body. Even though the requirements were developed by the industry, they should be reviewed by regulatory body for their application in nuclear power plants. The enhanced ISI activities and requirements of non-destructive examination(NDE) which reflect the degradation issues in nuclear power industry will be primarily discussed in this paper.

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.1115-1126
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• 2009

As one of the ways to enhance the economical features in sodium-cooled fast reactor development, the concept of an integrated steam generator and pump system (ISGPS) is proposed from a structural point of view. And the related intermediate heat transfer system (IHTS) piping layout compatible with the ISGPS is described in detail. To assure the creep design lifetime of 60 years, the structural integrity is investigated through high temperature structural evaluation procedures by the SIE ASME-NH computer code, which implements the ASME-NH design rules. From the results of this study, it is found that the proposed ISGPS concept is feasible and applicable to a commercial SFR design.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1616-1629
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• 2023

To quantify the conservatism of existing ASME strain-based evaluation methods for seismic loading, this paper presents very low cycle fatigue test data of elbows under various cyclic loading conditions and comparison of evaluation results with experimental failure cycles. For strain-based evaluation methods, the method presented in ASME BPVC CC N-900 and Sec. VIII are used. Predicted failure cycles are compared with experimental failure cycle to quantify the conservatism of evaluation methods. All methods give very conservative failure cycles. The CC N-900 method is the most conservative and prediction results are only ~0.5% of experimental data. For Sec. VIII method, the use of the option using code tensile properties gives ~3% of experimental data, and the use of the material-specific reduction of area can reduce conservatism but still gives ~15% of experimental data.