• Nuclear Engineering and Technology
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• v.23 no.1
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• pp.56-65
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• 1991

The LOFT LOCE L2-3 was simulated using the RELAP5/MOD2 Cycle 36.04 code to assess its capability in predicting the thermal-hydraulic phenomena in LBLOCA of a PWR. The reactor vessel was simulated with two core channels and split downcomer modeling for a base case calculation using the frozen code. The result of the base calculation showed that the code predicted the hydraulic behavior, and the blowdown thermal response at high power region of the core reasonably and that the code had deficiencies in the critical How model during subcooled-two-phase transition period, in the CHF correlation at high mass flux and in the blowdown rewet criteria. An overprediction of coolant inventory due to the deficiencies yielded the poor prediction of reflood thermal response. Improvement of the code, RELAP5 / MOD2 Cycle 36.04, based on the sensitivity study increased the accuracy of the prediction of the rewet phenomena.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.759-764
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• 1997

본 논문에서는 영광 3,4호기의 초기 부분충수 운전중 정지냉각 상실 사건에 대하여 확률론적 안전성평가(Probabilistic Safety Assessment; PSA)를 수행하였다. 1단계 PSA 결과인 노심손상빈도에 크게 영향을 끼치는 인간행위는 THERP(technique for human error rate prediction)를 사용하여 평가하였고, 사고경위는 KIRAP(KAERI integrated reliability analysis code package)을 이용하여 정량화하였다. 영광 3,4호기의 부분충수 운전중 정지냉각 상실 사건에 대한 예비적인 PSA 결과, 노심손상 빈도는 1.43E-6로 평가되었고 노심손상 빈도에 주요하게 기여하는 것은 원자로 냉각재 보충에 대한 운전원의 진단 실패로 나타났다. 노심손상빈도를 감소하는 방안의 하나는 운전원의 진단오류 확률을 낮추기 위해 노심손상까지의 운전원 여유시간을 확장하는 것이다. 그러나 보다 구체적인 결과는 분석에 필요한 여러 가지 자료들을 검토하고 PSA를 다시 수행해야 얻을 수 있을 것으로 판단된다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.275-282
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• 1996

일체형 원자로는 증기발생기 및 가압기를 압력용기 내에 설치한 것으로서 연결배관이 없기 때문에 배관의 파단에 의한 대형 냉각재 상실사고를 근원적으로 배제하고 전체계통을 단순화 시킬 수 있다. 증기발생기는 대부분 관류식으로서 일체식과 모듈식이 사용되고 있다. 본 연구에서는 모듈식 나선형 증기발생기를 사용한 일체형 신형경수로의 예비 개념설계를 수행하였다. 가압기는 원자로 내에 별도의 용기를 설치하는 내장형 자기가압기를 채택하였다. 제어봉 구동장치는 핵분열 반응열을 이용한 원자로 기동을 위하여 반응도를 미세하게 조정하는 것이 가능하고 지진하중과 같은 동하중의 영향을 최소화하기 위하여 원자로용기 외부로의 돌출부분을 최소화하는 설계개념을 채택하였다. 냉각재 펌프는 Canned Motor Pump를 원자로용기 상부에 수직으로 직접 부착하는 개념을 사용하였다.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.254-254
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.171.1-171
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• 1999

• Transactions of the Korean Society of Mechanical Engineers
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• v.7 no.3
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• pp.344-352
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• 1983

격납 용기 내에 비응축성 가스(공기)가 존재하는 경우에 증기의 응축 열전달 계수를 평가하는 방 법을 연구하였다. 유일한 대규모 격납 용기 실험인 CVTR자료를 이용하여 응축 열전달 계수를 계산하여, 현재 원자력 발전소의 냉각재 상실 사고(LOCA) 및 주 증기 배관 파열사고(MSLB)시에 격납 용기의 안전 해석에서 공식적으로 사용되고 있는 Tagami와 Uchida열전달 계수 관계식과 비교해 본 결과 좋은 일치를 보여 주었다.

• Nuclear Engineering and Technology
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• v.23 no.1
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• pp.66-80
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• 1991

The LOFT intermediate break experiment L5-1, which simulates 12 inch diameter ECC line break in a typical PWR, has been analyzed using the reactor thermal/hydraulic analysis code RELAP5/MOD2, Cycle 36.04. The base calculation, which modeled the core with single flow channel and two heat structures without using the options of reflood and gap conductance model, has been successfully completed and compared with experimental data. Sensitivity studies were carried out to investigate the effects of nodalization at reactor vessel and core modeling on major thermal hydraulic parameters, especially on peak cladding temperature(PCT). These sensitivity items are : single flow channel and single heat structure (Case A), two flow channel and two heat structures (Case B), reflood option added (Case C) and both reflood and gap conductance options added (Case D). The code, RELAP5/MOD2 Cycle 36.04 with the base modeling, predicted the key parameters of LOFT IBLOCA Test L5-1 better than Cases A,B,C and D. Thus, it is concluded that the single flow channel modeling for core is better than the two flow channel modeling and two heat structure is also better than single heat structure modeling to predict PCT at the central fuel rods. It is, therefore, recommended to use the reflood option and not to use gap conductance option for this L5-1 type IBLOCA.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.423-428
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• 1996

피동형원자로 AP600을 참조발전소로하여 설정된 1000MWe급 대용량 피동형원자로의 계통개념에 대한 안전계통 성능 평가 및 코드의 적용성 평가를 목적으로 RELAP5/MOD3코드를 사용하여 대형냉각재상실사고를 모의 해석하였다. 피동형 안전계통으로 축압기, CMT IRWST를 모델하였으며 가압기에 연결된 1단계부터 3단계까지의 자동감압밸브계통을 모델링 하고 4단계 자동감압밸브계통은 각 루프의 고온관에 연결되어 있는 것으로 모델링 하였다. 피동형 안전계통의 모델이 향상된 RELAP5/MOD3.2와 그 이전의 코드인 RELAP5/MOD3.1의 냉각재상실사고 모의계산결과 원자로내의 압력변화, 노심냉각수 주입유량 및 핵연료 피복재 온도 거동이 거의 유사하게 나타났으며 1000MWe급 대용량 피동형원자로의 안전계통은 냉각재 상실사고시 충분한 노심냉각능력을 가지는 것으로 분석되었다.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.4 no.2
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• pp.13-19
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• 2008

In order to predict a remaining life of a plant, it is necessary to select the components that are critical to the plant life. The remaining life of those components shall be evaluated by considering the aging effect of materials used as well as numerous factors. However, when evaluating reliability of nuclear structural components, some problems are quite formidable because of lack of information such as operating history, material property change and uncertainty in damage models. Accordingly, if structural integrity and safety are evaluated by the deterministic fracture mechanics approach, it is expected that the results obtained are too conservative to perform a rational evaluation of plant life. The probabilistic fracture mechanics approaches are regarded as appropriate methods to rationally evaluate the plant life since they can consider various uncertainties such as sizes and shapes of cracks and degradation of material strength due to the aging effects. The objective of this study is to evaluate the structural integrity for a reactor pressure vessel under the small break loss of coolant accident by applying the deterministic and probabilistic fracture mechanics. The deterministic fracture mechanics analysis was performed using the three dimensional finite element model. The probabilistic integrity analysis was based on the Monte Carlo simulation. The selected random variables are the neutron fluence on the vessel inside surface, the content of copper, nickel, and phosphorus in the reactor pressure vessel material, and initial RTNDT.

• Nuclear Engineering and Technology
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• v.19 no.1
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• pp.10-21
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• 1987

Uncertainty analysis of the FLB accident is performed for KNU-1 using the response surface methodology and Monte Carlo simulation. The FLB analyses using the RELAP4/Mod6 were performed a number of times to generate the data base for the uncertainty analysis, along with the EM calculation for comparison purpose. Two kinds of input sets are utilized for response surface method to investigate and compare the effects of the uncertainty of input variables on the RCS peak pressure following a FLB. The first set is composed of six major plant operational parameters and the second set is composed of five major modelling parameters. It is found through the analysis of results that the uncertainties of modelling parameters have more influence on the RCS peak pressure than the uncertainties of plant operational parameters and that the extra margin of 9% of peak pressure is gained. And one of the assumptions of EM calculation, which is usually accepted as conservative is found to be erroneous, that is, the initial core inlet temperature is found to act negatively on the RCS pressure following a FLB.