• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2003년도 춘계학술발표대회 개요집
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• pp.216-218
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• 2003

A fracture mechanics analysis methodology for stress corrosion cracks (SCCs) existing in the Alloy 600 nozzle weld joint for control rod drive mechanisms (CRDMs) of pressurized water reactor is studied. Effects of weld residual stresses on the sub-critical crack behavior during the reactor operation are investigated by a fracture mechanics analysis, which is combined with the finite element alternating method. It is found that effects f the residual stresses on the stress intensity factor (SIF) and crack growth rate (CGR) are dominant and values of SIF and CGR of cracks in the region of weld joint are increased by a factor of three or more on an average.

• Nuclear Engineering and Technology
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• 제15권1호
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• pp.41-49
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• 1983

특정 원자력발전소 안전성 계통의 신뢰도 분석을 위한 자료평가의 목적으로 전산코드를 개발하였으며 그 유용성을 입증하였다. 가압 경수로 보조급수 계통 신뢰도 분석을 위하여 개발된 전산코드를 이용하여 관련자료를 평가하였다. 이를 위하여 부품고장률의 선분포는 미국의 원자력안전성 연구보고서, 특정 발전소의 운전경험은 기 발간된 인허가자 사상보고서에서 얻었다. 분석결과 후분포는 대수정규분포 곡선에 잘 점철되며 분포의 오차인자들은 현저히 감소하는 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제42권3호
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• pp.259-270
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• 2010

New generation nuclear reactors are designed using advanced safety analysis methods. A thorough understanding of different interacting physical phenomena is necessary to avoid underestimation and overestimation of consequences of off-normal transients in the reactor safety analysis results. This feature requires a multiphysics reactor simulation model. In this context, a coupled dynamics model based on a multiphysics formulation is developed indigenously for the transient analysis of large pressurized VVER reactors. Major simplifications are employed in the model by making several assumptions based on the physics of individual phenomenon. Space and time grids are optimized to minimize the computational bulk. The capability of the model is demonstrated by solving a series of international (AER) benchmark problems for VVER reactors. The developed model was used to analyze a number of reactivity transients that are likely to occur in VVER reactors.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2001년도 봄 학술발표회 논문집
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• pp.259-266
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• 2001

This paper presents the results of a structural analysis to determine design variables such as the inner basket array type, and thicknesses of the outer shell and the lid and bottom of a spent nuclear fuel disposal canister. The canister construction type introduced here is a solid structure with a cast iron insert and a corrosion resistant overpack, which is designed for the spent nuclear fuel disposal in a deep repository in the crystalline bedrock, entailing an evenly distributed load of hydrostatic pressure from the groundwater and large swelling pressure from the bentonite buffer. Hence, the canister must be designed to withstand these large pressure loads. Many design variables may affect the structural strength of the canister. In this study, among those variables, the array type of inner baskets and thicknesses of outer shell and lid and bottom are attempted to be determined through a linear static structural analysis. Canister types studied here are one for the pressurized water reactor (PWR) fuel and another for the Canadian deuterium and uranium reactor (CANDU) fuel.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.824-829
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• 1998

미국의 개량형 원자력 발전소 개념설계단계에서 중대사고시 사고완화를 위한 전략으로 원자로 압력용기 외부냉각 개념이 제안되었다. 중대사고 진행과정에서 노심용융물이 원자로 압력용기 하부헤드로 재배치 되었을 때 압력용기 외벽을 냉각함으로서 노심용융물을 압력용기 내부에 가두어 두어 격납건물 내로의 유출을 방지하는 방식이다. 이 연구에서는 원자로 압력용기 하부헤드 내의 노심용융물 거동중 자연 순환에 의한 거동을 수치적으로 모의하여 보았다. 연구결과, 정상상태의 온도 및 속도분포는 현상학적으로 적절하게 모의되나 고화와 액화의 경우에는 고유모델의 필요성이 요구되었다.

• Nuclear Engineering and Technology
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• 제34권1호
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• pp.68-79
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• 2002

The approach for evaluating the critical heat flux (CHF) margin using the departure from nucleate boiling ratio (DNBR) concept has been widely applied to PWR core design, while DNBR in this approach does not indicate appropriately the CHF margin in terms of the attainable power margin-to-CHF against a reactor core condition. The CHF power margin must be calculated by increasing power until the minimum DNBR reaches a DNBR limit. The Critical Power Ratio (CPR), defined as the ratio of the predicted CHF power to the operating power, is considered more reasonable for indicating the CHF margin and can be calculated by a CPR orrelation based on the heat balance of a test bundle. This approach yields directly the CHF power margin, but the calculated CPR must be corrected to compensate for many local effects of the actual core, which are not considered in the CHF test and analysis. In this paper, correction of the calculated CPR is made so that it may become equal to the DNB overpower margin. Exemplary calculations showed that the correction tends to be increased as power distribution is more distorted, but are not unduly large.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.463-478
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• 2019

The most important non-functional requirements for dependability of any Embedded Real-Time Safety Systems are safety, availability and reliability requirements. System architecture plays the primary role in achieving these requirements. Compliance with these non-functional requirements should be ensured early in the development cycle with appropriate considerations during architectural design. In this paper, we present an application of system architecture modeling for quantitative assessment of system dependability. We use probabilistic model checker (PRISM), for dependability analysis of the DTMC model derived from system architecture model. In general, the model checking techniques do not scale well for analyzing large systems, because of prohibitively large state space. It limits the use of model checking techniques in analyzing the systems of practical interest. We propose abstraction based compositional analysis methodology to circumvent this limitation. The effectiveness of the proposed methodology has been demonstrated using the case study involving the dependability analysis of safety system of a large Pressurized Water Reactor (PWR).

• Nuclear Engineering and Technology
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• 제53권1호
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• pp.322-336
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• 2021

This paper presents a numerical prediction of the transient hydraulic loads acting on the tubes and external supports of a pressurized water reactor (PWR) steam generator (SG) during blowdown following a sudden feedwater line break (FWLB). A simplified SG model was used to easily demonstrate the prediction. The blowdown discharge flow was treated as a flashing flow to realistically simulate the transient flow fields inside the SG and the connected broken feedwater pipe. The effects of the SG initial pressure or the broken feedwater pipe length on the intensities or magnitudes of transient hydraulic loads were investigated. Then predictions of the decompression pressure wave-induced impulsive pressure differential loads on SG tubes and the transient blowdown loads on SG external supports were demonstrated and the general aspects of transient responses of such transient hydraulic loads to the FWLB were discussed.

• Nuclear Engineering and Technology
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• 제26권3호
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• pp.420-424
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• 1994

가압경수로형 원자력발전소의 저출력 및 과도상태에서의 개선된 증기발생기 수위 제어 방식을 제시하였다. 수축 및 팽창 현상에 의한 수위의 요동을 줄이기 위해 기존의 비례·적분 제어기에 증기발생기 압력 및 급수온도를 고려한 앞먹임 보상부를 첨가하였다. 원전 훈련용 시뮬레이터를 이용하여 시뮬레이션을 수행한 결과 기존방식에 비해 적은 수위오차, 훨씬 빠른 진정시간을 얻을 수 있었다. 제시된 알고리즘은 구현이 용이하고 실제 적용도 가능하리라 판단된다.

• Nuclear Engineering and Technology
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• 제25권3호
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• pp.371-380
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• 1993

운전원의 언어적인 제어 논리를 이용한 모델과 무관한 제어기가 증기 발생기의 수위 제어를 위해 개발되었다. 오직 9개의 룰을 수위 오차와, 응축 수위 변화를 표현하는 유량 오차로부터 얻어졌다. 벨형의 소속함수는 민감도 분석을 통하여 조절되었고, 결과 과도 상태와 정상상태 공히 양호한 제어성능을 보였다. 저출력에서 전 출력까지의 다단계 출력 증가에 대하여, 전 운전 영역에서 좋은 제어를 보였다.