• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.647-652
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• 1998

The YGN Units 3&4 plant conditions during shutdown operation were reviewed to identified the possible even scenarios following the loss of shutdown cooling. The Thermal hydraulic analyses were performed for the five cases of RCS configurations under the worst event scenario, unavailable secondary cooling and no RCS inventory makeup, using the RELAP5/MOD3.2 code to investigate the plant behavior, From the analyses results, times to boil, times to core uncovery and times to core heat up were estimated to determined the containment closure time to prevent the uncontrolled released of fission products to atmosphere, These data provide useful information to the abnormal procedure to cope with event.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.699-706
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• 1998

한국원자력연구소에서는 원자력중장기연구의 일환으로 한국형 표준원전을 모의하는 종합열수력실증실험을 계획하고 있으며, 현재 실험장치에 대한 척도해석(Scaling Analysis), 예비해석(Scoping Analysis) 및 개념설계를 수행하고 있다. 본 논문에서는 영광 3/4호기를 대상으로 척도해석을 통하여 실험장치를 개념설계하고, 저온관 6인치 소형냉각재 상실사고에 대하여 예비해석을 수행한 결과를 보여준다. 개념설계된 실험장치는 높이비가 참조원자로와 동일하고, 체적비가 1/200이다. 실험장치의 개념설계는 이상유동에 대한 3단계 척도법을 적용하였으며, 개념설계의 타당성을 입증하기 위해 RELAP5/MOD3.1 코드를 사용하여 정상상태 및 저온관 6인치 소형냉각재 상실사고시 계통의 거동을 예비 계산하였다. 실험장치에 대한 예비해석결과 사고 거동이 참조원자로와 잘 일치하는 것으로 나타났다. 또한 수평관 및 주냉각재펌프의 척도기준이 사고의 진행과정에 영향을 미치는 중요한 인자로 밝혀졌다

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.806-811
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• 1998

핵연료 피복관의 산화반응 현상은 중대사고시 원자로와 격납건물의 건전성을 위협하는 중요한 원인중의 하나이다 본 논문에서는 MELCOR에서 사용증인 Urbanic-Heidrich 상관식과 SCDAP/RELAP5/MOD3.1에서 사용중인 MATPRO-EG&G 상관식을 사용하여 산화 반응 모델이 노심손상에 미치는 영향을 울진원전3,4호기를 대상으로 MELCOR의 입력변수의 변화에 따른 민감도를 분석하였다. 분석결과, Urbanic-Heidrich 상관식이 MATPRO-EG&G상관식에 비해 핵연료 용융시작을 약 394초, 원자로 노심 하부에서의 용융물 재배치 (relocation)시작을 약 434초 가량 빨리 초래하여 사고진행에는 큰영향이 없음을 나타내고 있으나 노심하부 파손시점까지 발생한 수소량은 Urbanic-Heidrich 상관식이 MATPRO-EG&G상관식에 비해 약 1.4배정도 더 많이 발생시켜 격납건물 건전성에 대한 영향이 매우 크므로 보다 자세한 모델검토가 요구된다.

• 한국안전학회지
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• 제27권5호
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• pp.229-234
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• 2012

Operator's action time is evaluated from MAAP4 analysis used in conventional probabilistic safety assessment(PSA) of a nuclear power plant. MAAP4 code which was developed for severe accident analysis is too conservative to perform a realistic PSA. A best-estimate code such as RELAP5/MOD3, MARS has been used to reduce the conservatism of thermal hydraulic analysis. In this study, operator's action time of core cooling recovery operation is evaluated by using the MARS code, which its Fussell-Vessely(F-V) value was evaluated as highly important in a small break loss of coolant(SBLOCA) event and loss of component cooling water(LOCCW) event in previous PSA. The main conclusions were elicited : (1) MARS analysis provides larger time window for operator's action time than MAAP4 analysis and gives the more realistic time window in PSA (2) Sufficient operator's action time can reduce human error probability and core damage frequency in PSA.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.23-28
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• 1996

The potential for the RELAP5/MOD3.2 was assessed for the loss-of-RHR event during the mid-loop operation and the predictability of major thermal-hydraulic phenomena was also evaluated for the long term transient. The analysis results of the typical two cases(cold leg opening case and pressurizer opening case) were compared with experimental data which was conducted at ROSA-IV/LSTF in Japan. As a result, it was shown that tile code was capable of simulating the thermal-hydraulic transport process with appropriate time step during the reduced inventory operation with the loss-of- RHR system.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.310-316
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• 1996

현재 사용중인 Non-LOCA 해석용 인허가 코드들은 특정한 형태의 가압경수로에 맞게 짜여진 것들이어서 모든 형태의 가압경수로에 적용할 수 있는 범용 코드의 개발이 필요한 실정이다. 이를 위하여 한국원자력연구소에서는 웨스팅하우스 및 CE형 발전소에 공히 적용할 수 있는 과도현상 해석 코드인 TASS코드를 개발하고 있다. 이 TASS코드는 실시간보다 빠르게 핵증기계통에 대한 모의계산을 수행하며 대화식의 입출력을 통하여 사용자가 원하는 과도현상을 정확히 모사할 수 있다. 이 TASS코드의 웨스팅하우스형 발전소에 대한 적용타당성은 이미 검증되었으며, 본 논문에서는 CE형 발전소에 대하여 TASS 코드를 적용하여 Non-LOCA 인허가 해석을 하기위한 검증을 위해 주급수관 파단사고 및 주증기관 파단사고에 대하여 RELAP5/MOD3 코드와의 비교계산을 수행하였다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.551-556
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• 1997

A present study addresses a loss-of-RHR event as an initiating event under specific low power or shutdown conditions. Two typical plant configurations, cold leg opening case with water-filled steam generators and pressurizer opening case with emptied steam generators, were evaluated using the RELAP5/ MOD3.2 code. The calculation was compared with the experiment conducted at ROSA-IV/LSTF in Japan. As a result, the code was capable of simulating the system transient behavior following the event. Especially, thermal hydraulic transport processes including non-condensable gas behavior were reasonably predicted with an appropriate time step and CPU time. However, there were some code deficiencies such as too large system mass errors and severe flow oscillations in core region.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.315-320
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• 1995

The CHF phenomenon has been investigated for water flow under forced and natural circulation modes with vertical round tubes at low pressure and low flow condition. Experiments have been performed by using three different test sections for mass fluxes below 400 kg/㎡s under near atmospheric pressure. The experimental data for forced and natural circulation are compared with each other. To predict the flow rate at the two-phase region our test condition has been analyzed by RELAP5/MOD3 because the local two-phase condition inside the stainless steel tube cannot be directly measured. To predict the CHF with accuracy we have to consider the parameters at the single-phase region as well as the flow behavior at the two-phase region.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.713-718
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• 1998

Henry-Fauske critical flow model was incorporated into TRAC-PF1 to correct some errors in the original TRAC-PFI critical flow model. Henry-Fouske mode1 was numerically implemented and tested against steady-state steam-water experimental data. The model was incorporated into TRAC-PFI and code assessment against Marviken Critical Flow Tests 15 and 24 was carried out. Calculations using RELAP5/MOD3 were also made for comparison. Ten cases were calculated each test and sensitivity study on nodalization as well as critical flow or model was performed Stand-alone numerical model test and code assessment were done for verification and validation of code modification. Calculation results show that the modified version of TRAC-PF1 has a capability to model critical flow correctly in various conditions.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3787-3800
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• 2023

Reactivity Initiated Accident (RIA) scenarios require special attention using advanced simulation techniques due to their complexity and importance for nuclear power plant (NPP) safety. While the conservative approach has traditionally been used for safety analysis, it may lead to unrealistic results which calls for the use of best estimate plus uncertainty (BEPU) approach, especially with the current advances in computational power which makes the BEPU analysis feasible. In this work an Uncontrolled Control Element Assembly (CEA) Withdrawal at Full Power accident scenario is analyzed using the BEPU approach by loosely coupling the thermal hydraulics best-estimate system code (RELAP5/SCDAPSIM/MOD3.4) to the statistical analysis software (DAKOTA) using a Python interface. Results from the BEPU analysis indicate that a realistic treatment of the accident scenario yields a larger safety margin and is therefore encouraged for accident analysis as it may enable more economic and flexible operation.