• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1448-1459
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• 2023

A key element of the safety analysis is Loss of Coolant Analysis (LOCA) which must be performed using system thermal-hydraulic codes. These codes are extensively validated against separate effect and integral experiments. RELAP/SCDAPSIM is one such code that may be used to predict LBLOCA response in a CANDU reactor. The RD-14M experiment selected for the Best Estimate Plus Uncertainty study is a 44 mm (22.7%) inlet header break test with no Emergency Coolant Injection. This work has two objectives first is to simulate pipe break with RELAP and compare these results to those available from experiment and from comparable TRACE calculations. The second objective is to quantify uncertainty in the fuel element sheath (FES) temperature arising from model coefficient as well as input parameter uncertainties using Integrated Uncertainty Analysis package. RELAP calculated results are found to be in good agreement with those of TRACE and with those of experiments. The base case maximum FES temperature is 335.5 ℃ while that of 95% confidence 95^th percentile is 407.41 ℃ for the first order Wilk's formula. The experimental measurements fall within the predicted band and the trends and sensitivities are similar to those reported for the TRACE code.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 춘계학술발표회요약집
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• pp.248.2-248
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• 2001

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 춘계학술발표대회 요약집
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• pp.191.1-191.1
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• 2003

• Nuclear Engineering and Technology
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• 제36권4호
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• pp.285-294
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• 2004

Pressure wave propagation in the discharge piping with a sparger submerged in a water pool, following the opening of a safety relief valve, is analyzed. To predict the pressure transient behavior, a RELAP5/MOD3 code is used. The applicability of the RELAP5 code and the adequacy of the present modeling scheme are confirmed by simulating the applicable experiment on a water hammer with voiding. As a base case, the modeling scheme was used to calculate the wave propagation inside a vertical pipe with sparger holes and submerged within a water pool. In addition, the effects on wave propagation of geometric factors, such as the loss coefficient, the pipe configuration, and the subdivision of sparger pipe, are investigated. The effects of inflow conditions, such as water slug inflow and the slow opening of a safety relief valve are also examined.

• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3526-3539
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• 2022

Nuclear thermal-hydraulic system analysis codes have been developed to simulate nuclear reactor systems, which solve simplified governing equations by replacing source terms with constitutive relations for simulating entire reactor systems with low computational resources. For half a century, many efforts have been made for wider versatility and higher accuracy of system codes, but various factors can affect the code analysis results, and it was difficult to isolate these factors and interpret them individually. In this study, two system codes, RELAP5 and TRACE, which have many users and are highly reliable, are selected to analyze only the effects of constitutive relations. The influence of constitutive relations is analyzed using in-house platforms that replicate constitute relations of RELAP5 and TRACE equally to exclude factors that may affect analysis results, such as governing equation solvers and user effects. Among the various constitutive relations, the analysis is performed on the wall variables expected to have the most influence on the analysis results. Part 1 paper presents the methodology and wall friction model comparison, while Part 2 paper shows wall heat transfer comparison of the two selected codes.

• Nuclear Engineering and Technology
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• 제30권5호
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• pp.424-434
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• 1998

RELAP5/Mod3.1 code was assessed with the semiscale experiment S-NC-3, and S-NC-4, which simulated the two-phase natural circulation and reflux condensation for the SBLOCA of PWR, respectively . Test S-NC-3 and S-NC-4 calculation results showed that RELAP5/Mod3.1 quite well describes the influence of steam generator secondary side heat transfer degradation on both two-phase natural circulation and reflux condensation. A comparison between the calculated and measured two-phase mass flow rate in test S-NC-3 shows good agreement for primary mass inventory more than 92%. And RELAP5/Mod3.1 have a good mass flow rate prediction capability for the transient such as S-NC-4 except some flow oscillations. The reflux flow rate for S-NC-4 test is under predicted, and the overall results verify that the correct prediction of the reduced liquid level appears to be required for the correct calculation of the overall phenomena.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1990-1997
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• 2020

To support the safe operation of the Miniature Neutron Source Reactor (MNSR), a thermo-hydraulic transient model using the RELAP5/Mod3.2 code was simulated. The model was verified by comparing the results with the measured and the previously calculated data. The comparisons consisted of comparing the MNSR parameters under normal constant power operation and reactivity insertion transients. Reactivity Insertion Accident (RIA) for three different initial reactivity values of 3.6, 6.0, and 6.53 mk have been simulated. The calculated peaks of the reactor power, fuel, clad and coolant temperatures in hot channel were calculated in this model. The reactor power peaks were: 103 kW at 240 s, 174 kW at 160 s and 195 kW at 140 s, respectively. The fuel temperature reached its maximum value of 116 ℃ at 240 s, 124 ℃ at 160 s and 126 ℃ at 140 s respectively. These calculation results ensured the high inherently safety features of the MNSR under all phases of the RIAs.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.547-552
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• 1998

The standard RELAP5/MOD3.2 code were assessed with the condensation experiment in the presence of noncondensable gas in a vortical tube of PCCS of CP-1300. There are two wall film condensation models, the default model and the alternative model, in RELAP5/MOD3.2. The experimental apparatus was modeled with the two models, md simulations were performed for several sub-tests to be compared with the experimental results. In overall sense the simulation results showed that the default model of RELAP5/MOD3.2 under-predicts the heat transfer coefficients, while the alternative model over-predicts them throughout the condensing tube.

• Nuclear Engineering and Technology
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• 제27권5호
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• pp.645-660
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• 1995

표준원전을 대상으로하여 저수위 운전시의 잔열제거제통상실사고를 RELAP5/MOD3 및 RELAP5/MOD3.1 전산프로그램을 이용하여 분석하였다. 증기발생기가 이용가능할 때 원자로냉각재계통에 배기 경로가 없는 경우와 배기경로가 있는 경우에 대하여 분석을 수행하였다. 배기경로가 없는 경우에 대해 RELAP5 /MOD3 전산프로그램과 RELAP5 /MOD3.1 전산프로그램으로 비교 분석을 수행하였다. 분석 결과 두 전산프로그램의 계산결과는 정성적인 면 뿐 아니라 정량적 인면도 비교적 잘 일치하였다. 그러나 계산결과로부터 RELAP5 /MOD3의 경우에는 벽 열전달모델의 결함이 발견되어 배기경로가 있는 경우에 대해서는 RELAP5 /MOD3.1 전산프로그램을 이용하여 분석을 수행하였다. 분석결과 원자로정지후 하루가 지났을때 배기경로가 없는 경우에는 두개의 증기발생기로도 잔열이 충분히 제거되지 않아 원자로계통의 압력이 지속적으로 증가하여 사고개시 후4,000초 정도에 원자로계통의 임시밀봉재의 설계압력인 0.24MPa에 도달하였다. 가압기 안전밸브 용량의 세배정도 크기의 배기경로가 있는 경우에는 10,000 초가 지나도 원자로냉자재계통의 압력이 0.24 MPa에 도달하지 않았으며 노심노출이 초래되지 않았다. 분석결과의 상세한 검토를 통해서 저수위 운전시 잔열제거능력 상실사고가 발생하였을 경우 REL-AP5/MOD3.1을 이용한 사고해석 방법론의 타당성을 제안하였으며 또한 적절한 배기용량을 산정하기 위한 자료를 제공하였다.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.709-718
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• 2012

The pressurized water reactor APR1400 adopts DVI (Direct Vessel Injection) for the emergency cooling water in the upper downcomer annulus. The International Standard Problem number 50 (ISP-50) was launched with the aim to investigate thermal hydraulic phenomena during a 50% DVI line break scenario with best estimate codes making use of the experimental data available from the ATLAS facility located at KAERI. The present work describes the calculation results obtained for the ISP-50 using the RELAP5/MOD3.3 system code. The work aims at validation and assessment of the code to reproduce the observed phenomena and investigate about its limitations to predict complicated mixing phenomena between the subcooled emergency cooling water and the two-phase flow in the downcomer. The obtained results show that the overall trends of the main test variables are well reproduced by the calculations. In particular, the pressure in the primary system show excellent agreement with the experiment. The loop seal clearance phenomenon was observed in the calculation and it was found to have an important influence on the transient progression. Moreover, the collapsed water levels in the core are accurately reproduced in the simulations. However, the drop in the downcomer level before the activation of the DVI from safety injection tanks was underestimated due to multi-dimensional phenomena in the downcomer that are not properly captured by one-dimensional simulations.