• 부식과 방식
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• 제12권1호
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• pp.30-38
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• 2013

원전의 구조적 건전성에 문제가 될 수 있는, 오스테나이트계 합금의 환경조장균열(EAC)에 대한 거동을 실험적인 결과와 문헌 조사를 통해 분석하였다. 일차측 환경에서 주기적인 반복하중을 받을 때에는 기계적인 피로균열에 더해 수소유기균열이나 동적변형시효 등으로 인한 가속화 메커니즘을 통해 피로수명 감소가 나타났다. 따라서 EAF에 대한 저항성은 전반적인 부식저항성이 우수한 니켈기합금이 스테인리스강보다 크게 나타났다. 그러나 일정한 하중을 받을 때에는 내부산화에 의해 국부적인 취약부인 입계로의 빠른 균열의 생성과 진전이 나타나 일차수 응력부식균열(PWSCC)이라는 형태로 발생한다고 여겨진다. 이때는 니켈-크롬의 비율이 내부산화 저항성에 영향을 미쳐, 비율이 낮은 스테인리스강은 높은 저항성을 가지고, 비율이 높은 니켈기합금은 낮은 저항성을 가진다. 그러나 아직 이러한 균열 메커니즘에 대한 명확한 이해가 부족하므로, 명확히 규명하기 위해서는 추가적인 연구가 필요하다.

• 한국압력기기공학회 논문집
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• 제15권2호
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• pp.58-63
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• 2019

This paper presents preliminary analysis and results on IASCC sensitivity of a core shroud in the reactor pressure vessel. First, neutron irradiation flux distribution of the reactor internals was calculated by using the Monte Carlo simulation code, MCNP6.1 and the nuclear data library, ENDF/B-VII.1. Second, based on the neutron irradiation flux distribution, temperature and stress distributions of the core shroud during normal operation were determined by performing finite element analysis using the commercial finite element analysis program, ABAQUS, considering irradiation aging-related degradation mechanisms. Last, IASCC sensitivity of the core shroud was assessed by using the IASCC sensitivity definition of EPRI MRP-211 and the finite element analysis results. As a result of the preliminary analysis, it was found that the point at which the maximum IASCC sensitivity is derived varies over operating time, initially moving from the shroud plate located in the center of the core to the top shroud plate-ring connection brace over operating time. In addition, it was concluded that IASCC will not occur on the core shroud even after 60 years of operation (40EFPYs) because the maximum IASCC sensitivity is less than 0.5.

• 한국방사선학회논문지
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• 제15권6호
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• pp.803-811
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• 2021

본 연구에서는 공정 시뮬레이션 기술을 적용하여 조사유기응력부식균열 시험 작업자의 피폭량 평가를 하였다. 상용 공정 시뮬레이션 코드인 DELMIA Version 5를 사용하여 조사유기응력부식균열 분석 시험 설비, 핫셀 및 작업자를 작성하고 조사유기응력부식균열 시험 공정을 구현하였으며, 사용자 코딩을 통해 선량이 분포된 공간을 지나는 작업자의 누적 피폭량을 평가할 수 있도록 하였다. 작업자 모사를 위해 시험 공정별로 인체의 근골격계를 모방하여 약 200 개 이상의 자유도를 가지는 휴먼 마니킨 자세를 작성하였다. 작업자 피폭량 계산을 위하여 휴먼 마니킨 작업의 하위정보에 접근하여 자세 별 좌표, 시작 시간 및 유지 시간을 추출하였으며, 공간 선량 값과 자세 유지 시간을 곱하여 누적 피폭량을 계산하였다. 피폭량 평가를 위한 공간 선량은 MCNP6 Version 1.0을 사용하여 핫셀 내·외부 공간 선량을 계산하였으며, 계산된 공간 선량은 공정 시뮬레이션 도메인에 입력하였다. 공정 시뮬레이션을 이용한 피폭량 평가 결과와 전형적인 피폭량 평가 결과를 비교 분석한 결과, 상시 출입구역 내 일상 시험 작업에 대한 연간 피폭량은 각각 0.388 mSv/year 및 1.334 mSv/year로서 공정 시뮬레이션을 이용한 피폭량 평가 결과가 전형적인 방법의 피폭량 평가 결과 대비 70 % 낮게 예측되었다. 공간 선량 높은 구역에서 수행되는 특수작업에 대해서도 공정 시뮬레이션을 이용한 피폭량 평가를 수행하였으며, 피폭량이 높은 작업을 쉽게 선별할 수 있었고, 해당 작업의 휴먼 마니킨 자세와 공간 선량 가시화를 통해 직관적으로 작업 개선안을 도출할 수 있었다.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.1-10
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• 2016

The number of nuclear power plants operating in Korea was 24 as of year 2015. Nine units out of 24 units have been operated for a period over 20 years. Kori unit 1 has been in operation for 40 years, and an extended operation for Wolsong unit 1 was decided in 2015. There has been reported some crackings in reactor internals in PWR have been reported in Europe, USA, Japan and Korea, and some of them were replaced with new one. Repair and replacement technologies for the reactor internals have been developing in order to meet the regulatory requirements for long term operation in Korea. The technologies will also be used for the exported nuclear units. It is required to review degradation history of the reactor internals worldwide as a part of the degradation management program development. Schematics of reactor internals designed and supplied by Westinghouse, Framatome and Combustion Engineering are described herein. Materials degradation history of reactor internals of PWR plants in USA, Japan and Europe is surveyed and summarized. Some events from Korean plants are also described. Aging management strategy for the internals is suggested.

• 대한금속재료학회지
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• 제50권8호
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• pp.583-589
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• 2012

This work is concerned with a statistical approach to the life prediction on irradiation-assisted stress corrosion cracking (IASCC) of stainless steel (SS) for core internals of a pressurized water reactor (PWR). The previous results of the time-to-failure of IASCC measured on neutron-irradiated stainless steel components were statistically analyzed in terms of stress and irradiation. The accelerating life testing model of IASCC of cold worked Type 316 SS was established based on an inverse power model with two stress-variables, the applied stress and irradiation dose. Considering the variation of the yield strength and applied stress with the irradiation dose in the model, the remaining life of the baffle former bolt was statistically predicted during operation under complex environments of stress and irradiation.

• 대한금속재료학회지
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• 제47권12호
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• pp.819-827
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• 2009

This work is concerned with a statistical analysis of factors affecting the irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels for core internals of pressurized water reactors (PWR). The microstructural and environmental factors were reviewed and critically evaluated by the statistical analysis. The Cr depletion at grain boundary was determined to have no significant correlation with the IASCC susceptibility. The threshold irradiation fluence of IASCC in a PWR was statistically calculated to decrease from 5.799 to 1.914 DPA with increase of temperature from 320 to $340^{\circ}C$. From the analysis of the relationship between applied stress and time-to-failure of stainless steel components based on an accelerated life testing model, it was found that B2 life of a baffle former bolt exposed to neutron fluence of 20 and 75 DPA was at least 2.5 and 0.4 year, respectively, within 95% confidence interval.