• Nuclear Engineering and Technology
• /
• 제55권8호
• /
• pp.3102-3113
• /
• 2023

Following the Fukushima nuclear disaster, the simulation of accidents in the spent fuel pool has become more noticeable. Despite the low amount of decay heat power, the consequences of the accidents in a spent fuel pool (SFP) can be severe due to the high content of long-lived radionuclides and lack of protection by the pressure vessel. In this study, the loss-of-cooling accident (LOFA) for the VVER-1000/V446 spent fuel pool is simulated by employing RELAP5 and MELCOR 1.8.6 as the best estimate and severe accident analysis codes, respectively. For two cases with different total power levels, decay heat of spent fuels is calculated by ORIGEN-II code. For modeling SFP of a VVER-1000, a qualified nodalizations are considered in both codes. During LOFA in SFP, the key sequences such as heating up of the pool water, boiling and reducing the water level, uncovering the spent fuels, increasing the temperature of the spent fuels, starting oxidation process (generating Hydrogen and extra power), the onset of fuel melting, and finally releasing radionuclides are studied for both cases. The obtained results show a reasonable consistency between the RELAP5 and MELCOR codes, especially before starting the oxidation process.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
• /
• pp.368-373
• /
• 1996

A simple set of the transition criterion of the condensation regimes and the heat transfer coefficients on the direct contact condensation of the core makeup tank is developed, and implemented in RELAP5/MOD3.1 The condensation regimes are divided into two regimes: supply limit and condensation limit. In mode]ing the transition criterion between two regimes, a large-eddy model developed by Theofanous is used, and the empirical coefficient of the present large-eddy model is close to that of the large-eddy model. It turns out that the modified code better predicts the experimental data, especially the injection flow rate and the water level trend than the original code does.

• Nuclear Engineering and Technology
• /
• 제21권4호
• /
• pp.259-266
• /
• 1989

LOFT L2-5 대형 냉각재상실사고 모사실험에 대해 RELAP5/MOD2/Cycle 36.04로부터 개선된 코드를 이용, 코드평가를 수행하였다. 강수관(Downcomer)모델 및 노심유로모델에 따른 코드 민감도분석을 위해 보충계산을 수행하였다. 계산결과는 1차계통의 압력, 파단부를 통과하는 질량유량, 노심 고온부의 피복재 온도등에 대해서 실험결과와 비교 분석되었다. 분석결과 RELAP5/MOD2 계산에 의해 1차계통의 수력학적 거동은 잘 묘사될 수 있으며, 단일 노심유로 모델을 이용한 계산에서는 사고발생이후 감압기간동안 노심이 과대 냉각되는 현상이 발견되었다. 노심의 고온유로에서의 수력학적 거동을 잘 묘사할 수 있는 이중 노심유로 모델계산을 이용하여 이 현상을 극복하고 실험치에 근사하는 결과를 얻을 수 있음을 알 수 있었다.

• 시스템엔지니어링학술지
• /
• 제18권2호
• /
• pp.58-74
• /
• 2022

Deterministic accident analysis plays a central role in the nuclear power plant (NPP) safety evaluation and licensing process. Traditionally the conservative approach opted for the point kinetics model, expressing the reactor core parameters in the form of reactivity and power tables. However, with the current advances in computational power, high fidelity multi-physics simulations using real-time code coupling, can provide more detailed core behavior and hence more realistic plant's response. This is particularly relevant for transients where the core is undergoing reactivity anomalies and uneven power distributions with strong feedback mechanisms, such as reactivity initiated accidents (RIAs). This work addresses a RIA, specifically a control element assembly (CEA) withdrawal at power, using the multi-physics analysis tool RELAP5/MOD 3.4/3DKIN. The thermal-hydraulics (TH) code, RELAP5, is internally coupled with the nodal kinetics (NK) code, 3DKIN, and both codes exchange relevant data to model the nuclear power plant (NPP) response as the CEA is withdrawn from the core. The coupled model is more representative of the complex interactions between the thermal-hydraulics and neutronics; therefore the results obtained using a multi-physics simulation provide a larger safety margin and hence more operational flexibility compared to those of the point kinetics model reported in the safety analysis report for APR1400. The systems engineering approach is used to guide the development of the work ensuring a systematic and more efficient execution.

• Nuclear Engineering and Technology
• /
• 제23권1호
• /
• pp.66-80
• /
• 1991

전형적 PWR 비상노심냉각계통에서의 12 inch 파단사고에 대응하는 LOFT 중형냉각재 상실사고 모사 실험 자료 L5-1을 이용하여 RELAP5/MOD2 Cycle 36.04 전산코드의 평가가 수행되었다. 평가 근거는 기준 코드와 nodalization에 의한 계산 결과가 L5-1 실험 결과와 잘 일치하는지, 추가적인 민감도 분석 연구로는 이중 노심 유로 및 열적 모델을 고려하고 model 민감도 분석으로는 reflood, gap conductance option 사용 여부에 따른 피복재 온도에 미치는 영향을 관찰하였다. 기준 계산 결과 기준 모델이 L5-1 현상을 대체로 잘 모사하였으나, 피복재가 천천히 가열되고 주변 부위의 피복재 온도가 과대하게 예견되었다. 민감도 분석 결과 단일 열적 모델이 피복재 가열 시작을 10초개선 하였고, 이중 유로 모델이 주변 온도를 20K 개선하였으나 최대 피복재 온도는 기준 계산시 보다 정확치 못하였으므로, 기준 모델인 단일 유로, 이중 열적 구조 그리고 reflood option은 사용하고 gap conductance option은 사용치 않는 것이 코드의 중형 냉각재 상실사고 해석시 피복재 온도 관찰의 관점에서 바람직하다.

• Nuclear Engineering and Technology
• /
• 제26권1호
• /
• pp.126-139
• /
• 1994

본 연구에서는 BETHSY 실험장치에서 수행한 6" 소형 냉각재 상실사고(LOCA) 실험을 최적 열수력 코드인 CATHARE2 V1.2와 RELAP5/MOD3를 이용하여 계산했다. 본 연구의 주 목적은 소형 LOCA시 관심을 가지는 주요 물리현상인 이상 임계유동, 감압과정, 노심수위 감소, loop seal clearing 등에 대한 두 코드의 소형 LOCA 계산모의능력을 평가하는 것이다. 두코드는 이상 유동현상의 전개 경향이나 발생시점을 비교적 잘 예측하는 것으로 나타났고, CATHARE2의 경우가 실험과 더 잘 일치했다. 그렇지만 두 코드는 loop seal clearing 현상, loop seal clearing 발생후의 노심수위, accumulator 유량거동 등의 예측에는 약간의 편차를 보였는데, 편차의 정도는 RELAP5가 CATHARE2보다 더 큰 것으로 나타났다. 두 코드의 편차요인을 보다 상세히 분석하기 위하여 계면 마찰력, mesh크기, 파단노즐 junction에서의 방출계수(Discharge coefficient)등에 대하여 민감도분석을 수행하였다. 그 결과 CATHARE2의 경우는 계면 마찰력을 증가시킴으로써 감압과정시 일차계통의 질량분포, 즉 증기 발생기 입구 공동(SG inlet plenum)에서의 차압과 Cross√er leg의 차압이 개선되었으며, 증기발생기 외측 열전달계수를 증가시킴으로써 중기발생기의 압력변화를 개선할 수 있었다. RELAP5의 경우는 어떤 하나의 입력변수를 변화시켜서 과도기의 결과를 개선할 수 없었으며 다만, 계면 마찰력 모델링에 여전히 많은 불화실성이 내포되어 있음을 확인했다.확인했다.

• Nuclear Engineering and Technology
• /
• 제54권12호
• /
• pp.4776-4797
• /
• 2022

In the one-dimensional thermal-hydraulic (TH) codes, a subcooled boiling model to predict the void fraction profiles in a vertical channel consists of wall heat flux partitioning, the vapor condensation rate, the bubbly-to-slug flow transition criterion, and drift-flux models. Model performance has been investigated in detail, and necessary refinements have been incorporated into the Safety and Performance Analysis Code (SPACE) developed by the Korean nuclear industry for the safety analysis of pressurized water reactors (PWRs). The necessary refinements to models related to pumping factor, net vapor generation (NVG), vapor condensation, and drift-flux velocity were investigated in this study. In particular, a new NVG empirical correlation was also developed using artificial neural network (ANN) techniques. Simulations of a series of subcooled flow boiling experiments at pressures ranging from 1 to 149.9 bar were performed with the refined SPACE code, and reasonable agreement with the experimental data for the void fraction in the vertical channel was obtained. From the root-mean-square (RMS) error analysis for the predicted void fraction in the subcooled boiling region, the results with the refined SPACE code produce the best predictions for the entire pressure range compared to those using the original SPACE and RELAP5 codes.

• Nuclear Engineering and Technology
• /
• 제54권5호
• /
• pp.1860-1873
• /
• 2022

Nuclear thermal-hydraulic system analysis codes have been developed to comprehensively model nuclear reactor systems to evaluate the safety of a nuclear reactor system. For analyzing complex systems with finite computational resources, system codes usually solve simplified fluid equations for coarsely discretized control volumes with one-dimensional assumptions and replace source terms in the governing equations with constitutive relations. Wall boiling heat transfer models are regarded as essential models in nuclear safety evaluation among many constitutive relations. The wall boiling heat transfer models of two widely used nuclear system codes, RELAP5 and TRACE, are analyzed in this study. It is first described how wall heat transfer models are composed in the two codes. By utilizing the same method described in Part 1 paper, heat fluxes from the two codes are compared under the same thermal-hydraulic conditions. The significant factors for the differences are identified as well as at which conditions the non-negligible difference occurs. Steady-state simulations with both codes are also conducted to confirm how the difference in wall heat transfer models impacts the simulation results.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1995년도 추계학술발표회논문집(1)
• /
• pp.403-410
• /
• 1995

A loss of the residual heat removal (RHR) accident during mid-loop operation of Yong-gwang Nuclear Units 3/4 was analyzed using the RELAP5/MOD3.1.2 code. In this work the following assumptions are used; (i) initially the reactor coolant system (RCS) above the hot leg center line is filled with nitrogen gas, (ii) two 3/4-inch diameter vent valves on the reactor vessel head and the top of pressurizer in the reactor coolant system are always open, and a level indicator is connected to the RMR suction line, (iii) the two steam generators are in wet layup status and the steam generator atmospheric dump valve assemblies are removed so that the secondary side pressure remains at nearly atmospheric condition throughout the accident, and (iv) the loss of RHR is presumed to occur at 48 hours after reactor shutdown. Findings from the RELAP5 calculations are (i) the core boiling begins at ∼5 min, (ii) the peak RCS pressure is ∼3.0 bar, which implies a possibility of temporary seal break, (iii) ∼94 % of the decay heat is removed by reflux condensation in the steam generator U-tubes in spite of the presence of noncondensable gas, (iv) the core uncovery time is evaluated to be 7.2 hours. Significant mass errors were observed in the calculations.

• Nuclear Engineering and Technology
• /
• 제29권3호
• /
• pp.229-239
• /
• 1997

To investigate the realistic behavior of mass and energy release and resultant containment response during large break Loss of Coolant accident (LOCA), analyses are performed for Yonggwang (YGN) 3&4 nuclear power plants by using a merged version of RELAP5/CONTEMPT4 computer code. Comparative analyses by using conservative design computer codes are also peformed. The break types analyzed are the double-ended guillotine breaks at the cold leg and hot leg. The design analysis resulted in containment peak pressure during post-blowdown phase for the cold leg break. However, the RELAP5/CONTEMPT4 analyses show that the containment pressure has a peak during blowdown phase, thereafter it decreases monotonously without the second port-blowdown peak. For the hot leg break, revised design analysis shows much lower pressure than that reported in YGN 3&4 final safety analysis report. The RELAP5/CONTEMPT4 analysis shoos similar trend and confirmed that the bypass flow through the broken loop steam generator during post-blowdown is negligibly small compared to that of cold leg break. The low pressure and temperature predicted tv realistic analysis presented in this paper suggest that the design analysis methodology contains substantial margin and it can be improved to provide benefit in investment protection, such as, relaxing plant technical specifications and reducing containment design pressure.