• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.441-446
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• 1996

INEL에서 가압 경수로에서 일어날 수 있는 가상 사고 해석을 위해 개발된 Relap5/Mod3.2를 PC Version으로 변환 설치하였다. CPU Time 비교 및 검증 계산을 수행했으며 중요 변수 비교를 통해 변환의 적절함을 확인했다. Accumulator 주입에 의한 Numerical Oscillation이 일어나는 부분에서 서로 다른 시간 간격 제어를 함으로써 HP Version 결과와 차이가 있는 것을 제외하고 대체적으로 유사한 결과를 보임을 알 수 있었다. 경제적 측면과 Microprocessor의 발전 속도 측면에서 볼 때 PC Version 사용이 매우 유리한 선택이 될 수 있을 것이다.

• Nuclear Engineering and Technology
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• v.36 no.2
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• pp.142-152
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• 2004

The liquid entrainment and vapor pull-through offtake model of RELAP5/MOD3 had been developed for SBLOCA (Small Break Loss of Coolant Accident). The RELAP5/MOD3 model for horizontal volumes accounts for the phase separation phenomena and computes the flux of mass and energy through a branch when stratified conditions occur in the horizontal pipe. In the case of CANDU reactor, this model should be used in the coolant flow of 95 feeders connected to the reactor header component under the horizontal stratification in header. The current RELAP5 model can treat the only 3 directions junctions; vertical upward, downward, and side oriented junctions, and thus improvements for the liquid entrainment and vapor pull-through model were needed for considering the exact angles. The RELAP5 off-take model was modified and generalized by considering the geometric effect of branching angles. Based on the previous experimental results, the critical height correlation was reconstructed by use of the branch line connection angle and validation analyses were also performed using SET. The new model can be applied to vertical upward, downward and angled branch, and the accuracy of the new correlations is more improved than that of RELAP5.

• Nuclear Engineering and Technology
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• v.52 no.9
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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.

• Nuclear Engineering and Technology
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• v.18 no.3
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• pp.175-182
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• 1986

The KNUI (Korea Nuclear Unit 1) loss of offsite power transient as a design-base accident has been simulated using the RELAP5/MOD1/NSC computer code. The analysis is carried out using the best-estimate methodology, but the sequence and its assumptions are based on the evaluation methodology th at emphasizes conservatism. Important thermal-hydraulic parameters such as average temperature, steam generator level and pressurizer water volume are compared with the results in the KNU1 Final Safety Analysis Report (FSAR). The present analysis gives much lower RCS average temperature and pressurizer water volume, and much higher S/G water volume at the turnaround point, which may be considered to be additional improved safety margins. This is expected since the present analysis deals with the best-estimate thermal-hydraulic models as well as the initial conditions on a best-estimate basis. These additional safety margins may contribute to further validate the safety of the KNU1 in this type of accidents(Decrease in Heat Removal by the Secondary System).

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.396-402
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• 1995

영광 3,4호기의 부분충수 운전중 정지냉각계통이 상실되고 가압기 Manway가 개방된 사고에 대하여 RELAP5/MOD3.1.2의 열수력 코드를 이용하여 모의하였다. 계산결과 계통의 압력은 최고 1.74bar 까지 도달하였으며, 사고 발생 후 약 1시간 이후부터 계통은 노심이 노출될 때까지 유사 정상상태를 유지한다. 이때 가압기 Manway를 통해 방출되는 증기량은 약 4 kg/s로 붕괴열의 약 80%를 담당하고 증기발생기 2차측에 의해 나머지 20% 가량 제거된다. 또한 비응축성 가스는 계통에 남아 있는 한 계통의 압력 상승율을 증가시키며, RELAP5/MOD3.1.2 계산결과는 일차계통 전체 냉각재의 약 26 %의 질량오차를 나타냈다.

• Nuclear Engineering and Technology
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• v.27 no.4
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• pp.470-482
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• 1995

The Ulchin 3&4, which are 2825 MWt PWRs, adopted Safety Depressurization System (SDS) to mitigate the beyond design basis event of Total Less of Feedwater(TLOFW). In this study the results and methodology of the analyses for the determination of SDS bleed capacity are discussed. The SDS design bleed capacity has been determined from the CEFLASH-4AS/REM simulation according to the following design criteria : 1) Each SDS flow path, in conjunction with one of two High Pressure Safety Injection (HPSI) pumps, is designed to have a sufficient capacity to prevent core uncovery if one SDS path is opened simultaneously with the opening of the Pressurizer Safety Valves (PSVs). 2) Both SDS bleed paths are designed to have sufficient total capacity with both HPSI pumps operating to prevent core uncovery if the Feed and Bleed (F&B) initiation is delayed up to thirty minutes from the time of the PSVs lift. To verify the results of CEFLASH-4AS/REM simulation a comparative analysis kas also been per-formed by more sophisticated computer code, RELAP5/MOD3. The TLOFW event without operator recovery and TLOFW event with F&B are analyzed. The predictions by the CEFLASH-4AS/REM of the transient too phase system behavior are in good qualitative and quantitative agreement with those by the RELAP5/MOD3 simulation. Both of the results of analyses by CEFLASH-4AS/REM and RELAP5/MOD3 have demonstrated that decay heat removal and core inventory make-up can be successfully accomplished by F&B operation during now event for the Ulchin 3&4.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.173-173
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• 2001

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.981-988
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• 2018

An experiment was performed for the OECD/NEA ROSA-2 Project with the large-scale test facility (LSTF), which simulated a steam generator tube rupture (SGTR) accident due to a double-ended guillotine break of one of steam generator (SG) U-tubes with operator recovery actions in a pressurized water reactor. The relief valve of broken SG opened three times after the start of intact SG secondary-side depressurization as the recovery action. Multi-dimensional phenomena specific to the SGTR accident appeared such as significant thermal stratification in a cold leg in broken loop especially during the operation of high-pressure injection (HPI) system. The RELAP5/MOD3.3 code overpredicted the broken SG secondary-side pressure after the start of the intact SG secondary-side depressurization, and failed to calculate the cold leg fluid temperature in broken loop. The combination of the number of the ruptured SG tubes and the HPI system operation difference was found to significantly affect the primary and SG secondary-side pressures through sensitivity analyses with the RELAP5 code.

• Nuclear Engineering and Technology
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• v.32 no.1
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• pp.57-66
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• 2000

The RELAP5 code has been developed for best-estimate simulation of transients and accidents for pressurized water reactors and their associated systems, but it has not been fully assessed for those of CANDU reactors. However, a previous study suggested that the RELAP5 code could be applicable to simulate the transients and accidents for CANDU reactors. Nevertheless, it is indicated that there are some works to be resolved, such as modeling of headers and multi-channel simulation for the reactor core, etc. Therefore, this study has been initiated with an aim to identify the code applicability for all the postulated transients and accidents in CANDU reactors. In the present study, the small inlet header break experiment (B8604) in the RD-14 test facility was simulated with RELAP5/MOD3.2 code. The RELAP5 results were also compared with both experimental data and those of CATHENA analyses performed by AECL and the analyses demonstrated the code's capability to predict major . phenomena occurring in the transient with sufficient accuracy for both Qualitative and quantitative viewpoint However, some discrepancies in the depressurization of the primary heat transport system after the break and the consequent time delay of the major phenomena were also observed.

• Nuclear Engineering and Technology
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• v.33 no.1
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• pp.12-24
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• 2001

The post-LOCA long term cooling (LTC) performance of the Korean Standard Nuclear Power Plant (KSNPP) is analyzed for both small break loss-of-coolant accidents (LOCA) and large break LOCA at cold leg. The RELAP5/MOD3.2.2 beta code is used to calculate the LTC sequences based on the LTC plan of the Korean Standard Nuclear Power Plants (KSNPP). A standard input model is developed such that LOCA and the followed LTC sequence can be calculated in a single run for both small break LOCA and large break LOCA. A spectrum of small break LOCA ranging from \ulcorner.02 to 0.5 k2 of break area and a double-ended guillotine break are analyzed. Through the code calculations, the thermal-hydraulic behavior and the boron behavior are evaluated and the effect of the important action including the safety injection tank (SIT isolation and the simultaneous injection in LTC procedure is investigated. As a result, it is found that the sufficient margin is available in avoiding the boron precipitation in the core. It is also found that a further specific condition for the SIT isolation action need to be setup and it is recommended that the early initiation of the simultaneous injection be taken for larger break LTC sequences.