• 비파괴검사학회지
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• 제7권1호
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• pp.51-53
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• 1987

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 1999년도 특별강연 및 춘계학술발표대회 개요집
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• pp.141-142
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• 1999

• 한국압력기기공학회 논문집
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• 제16권1호
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• pp.65-73
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• 2020

Integrity evaluations on a platen superheater were conducted as per ASME Section VIII Division 2(hereafter 'ASME VIII(2)') which was originally used for design with implicit consideration of creep effects. A platen superheater subjected to severe loading conditions of high pressure and high temperature at creep regime in a supercritical thermal plant in Korea was chosen for present study. Additional evaluations were conducted as per nuclear-grade high-temperature design rule of RCC-MRx that takes creep effects into account explicitly. Comparisons of the two results from ASME VIII(2) and RCC-MRx were conducted to quantify the conservatism of ASME VIII(2). From present analyses, it was shown that the design evaluation results exceeded allowable limits of RCC-MRx for the plant design conditions although limits of ASME VIII(2) were satisfied regardless of operation time, which means that design as per ASME VIII(2) might be potentially non-conservative in case of operation in creep range. A high-temperature design evaluation program as per RCC-MRx, called 'HITEP_RCC-MRx' has been used and it was shown that pressure boundary components can be designed reliably with the program especially for the loading conditions of long-term creep conditions.

• 비파괴검사학회지
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• 제33권5호
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• pp.465-471
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• 2013

원전 증기발생기는 원자로 냉각재 계통에서 발생한 열에너지를 터빈 계통의 주급수에 전달하여 터빈을 회전시키기 위한 증기를 생산하는 일종의 열교환기이다. 증기발생기 전열관의 손상은 증기발생기의 구조적 및 누설 건전성 유지 능력을 저해시키기 때문에 주기적으로 와전류검사를 수행하여 전열관의 건전성을 평가한다. 증기발생기 전열관의 건전성 평가는 보통 원자로 연료 재장전 기간 중에 수행된다. 현재 국내 증기발생기 전열관에 적용되는 와전류검사는 KEPIC 및 ASME 코드 요건에 따라 수행되며, 와전류검사 수행에 필요한 검사 시스템은 와전류검사 장치와 수집된 신호를 평가하기 위한 평가 프로그램으로 구성된다. 검사에 적용되는 와전류검사 시스템을 구성하는 핵심기기인 와전류검사 장치는 ASME 및 KEPIC 코드에서 총 고조파 왜곡율, 입출력 임피던스, 증폭기 직선성 및 안정성, 위상 직선성, 대역폭 및 복조필터 응답, 디지털 변환, 채널 간섭 등과 같은 전기적 특성을 측정하도록 규정하고 있다. 이에 따라 본 논문에서는 국내 최초로 개발한 원전 증기발생기 와전류검사 장치의 전기적 특성 측정을 위한 ASME 및 KEPIC 코드 요건을 설명하고, 이 요건에 따른 증기발생기 와전류검사 장치의 전기적 특성의 측정 결과를 제시하였다.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.101-106
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• 2016

NSSS (Nuclear Steam Supply System) and BOP (Balance of Plant) design works for PGSFR (Prototype Gen-IV Sodium Fast Reactor) have been conducted in Korea. NSSS major components, e.g. reactor vessel, steam generator and secondary sodium main pipes, are designed according to the rule of ASME boiler and pressure vessel code division 5, in which DBA (Design by Analysis) methods are used in the stress assessments. However, there is little discussions about detail rules for BOP piping design. In this paper, the detail methodologies of BOP piping stress assessment are discussed including safety systems and non-safety system pipings. It is confirmed that KEPIC MGE(ASME B31.1) and ASME BPV code division 5 HCB-3600 can be used in stress assessments of non-safety pipes and class B pipes, respectively. However, class A pipe design according to ASME BPV code division 5 HBB-3200 has many difficulties applying to PGSFR BOP design. Finally, future development plan for class A pipe stress assessment method is proposed in this paper.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 춘계학술대회
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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.

• 한국안전학회지
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• 제22권1호
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• pp.13-18
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• 2007

Structural integrity assessment of thin-walled pipes and pipe items has become one of the major issues in the nuclear power plant. ASME Section XI Code Case N-597-2 provides a criterion for acceptance of the pipes. But the code case has several limitations for application and sometimes gives too conservative or non-conservative results. So it is necessary to understand fully the technical bases of the code case. In the code case N-597, the allowable local thicknesses of thinned straight pipes are given for three different cases. Because of the different technical base, each case gives different thickness values and sometimes gives contradictory values. In this paper attempts were made in order to propose a unified rule for the allowable local thickness and in order to remove or relax the restrictions on the application of the code case. For this purpose elastic stress analyses were made using the finite element method and the stress results were examined. Based on the obtained bending stress results, a very simple procedure was proposed to obtain the consistent allowable local thickness for the thinned straight pipes.

• 한국산학기술학회논문지
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• 제9권6호
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• pp.1500-1505
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• 2008

본 논문은 원자력 내진등급 IIA 글로브 밸브의 지진하중에 대한 구조 건전성을 평가하였다. 밸브 구조물을 3차원 모델링하여 유한요소법을 사용하여 모달 해석 및 등가 정적 응력해석을 수행하였다. 모달 해석 결과 밸브 구조물은 충분히 강건하여 등가 정적해석이 가능하였고 해석에서 구한 응력들을 조합하여 원자력 규격집에서 제시한 허용응력과 비교하여 밸브의 구조 건전성을 판단하였다. 모든 지진하중에 대한 밸의 구조물에서의 응력은 허용값 보다 작게 분포하므로 구조적 건전성을 유지한다고 평가할 수 있다.

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2407-2417
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• 2021

A structural integrity evaluation program (STEP) was developed for the high temperature reactor design evaluation according to the ASME Boiler and Pressure Vessel Code (ASME B&PV), Section III, Rules for Construction of Nuclear Facility Components, Division 5, High Temperature Reactors, Subsection HB. The program computerized HBB-3200 (the design by analysis procedures for primary stress intensities in high temperature services) and Appendix T (HBB-T) (the evaluation procedures for strain, creep and fatigue in high temperature services). For evaluation, the material properties and isochronous curves presented in Section II, Part D and HBB-T were computerized for the candidate materials for high temperature reactors. The program computerized the evaluation procedures and the constants for the weldment. The program can generate stress/temperature time histories of various loads and superimpose them for creep damage evaluation. The program increases the efficiency of high temperature reactor design and eliminates human errors due to hand calculations. Comparisons that verified the evaluation results that used the STEP and the direct calculations that used the Excel confirmed that the STEP can perform complex evaluations in an efficient and reliable way. In particular, fatigue and creep damage assessment results are provided to validate the operating conditions with multiple types of cycles.

• 기계저널
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• 제20권6호
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• pp.457-467
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• 1980

가동중조사(ISI)에 적용되는 ASME-MI 의 1974년 Edition에 ISI시 검출된 흠함을 평가하는 절 차가 제시되었다는 것은 이미 언급한 바 있다. 가동중검사의 개념은 ASME-III에 따라 건설된 부품이라 하더라도 원자력발전소의 온 수명기간동안 또는 정해진 수명기간동안 재료 내부에 흠 함이 전혀 생기지 않는다거나 또는 성장하지 않는다고 볼 수 없다는데 에 근거를 둔 것이다. 즉, 부품을 건설할 때 ASME-III의 규정에 따라 수행한 비파괴시험(nondestructive examination)에 의하여 합격된 부품이라 할지라도 시험방법의 유효성 및 합격기준등에 의하여 발견되지 않았 거나, 발견되었더라도 합격된 흠합등이 있을 수 있다. 이러한 흠함들이 원자로의 가동과 더불어 성장하거나, 또는 부품의 사용조건에 따라 흠함이 새로이 발생할 경우는 RCPB의 구조적 건전성 및 안정성은 위협을 받게된다. 이에 대처하기 위하여 원자력발전소의 전수명기간의 통하여 정 기적으로 ISI를 수행해 줄 것이 10CFR50의 50, 55a, "Codes and Standards"에 법적으로 명문 화되어 있다. ISI 시발견된 흠함은 그 크기를 평가하여 합격기준을 초과하는 경우에는, 흠함의 제거(removal), 항리(repair), 부품의 대체(replacement), 또는 파괴력학적인 해석에 의하여 그러한 흠함이 구조건 안전성을 손상시키지 않는다는 것을 입증해 주어야 한다. ISI 시에 검출된 흠함을 평가하고 파괴력학적으로 해석하는 규정으로서 ASME-XI에 명시된 절차는 다음과 같다.차는 다음과 같다.