• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.376-380
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• 1996

RD-14은 CANDU형 발전소를 묘사하기 위한 실험장치로서 여러 초기조건에서 자연순환 실험이 수행되었다. 본 연구에서는 RD-14 실험장치의 자연순환 실험을 RELAP5 코드로 모사하여 결과를 비교·평가하였다. 본 연구는 CANDU형 발전소 사고해석의 검증용 코드로서 RELAP5 코드의 적용 타당성을 평가하기 위한 과정으로 수행되었다. 계산결과, RELAP5 코드는 RD-14 실험장치의 자연순환 실험을 잘 예측하고 있음을 보여주고 있어 CANDU형 발전소의 자연순환 평가에 유용하다는 결론을 얻었다.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.415-421
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• 1997

자연대류 현상의 특성파악과 주요 관련변수들에 대한 이론적인 연구를 위하여 단일 순환계통 자연대류 실험인 SNC-3, 4를 RELAP5/Mod3.1로 모사계산하여, RELAP5/Mod1의 모사결과와 비교하여, RELAP5의 버전에 따른 결과의 차이를 비교하도록 하였다. 이를 위하여 Mod1의 input을 수정하여Mod3.1로 바꾸고 이를 계산하도록 하였다. 두 실험에 대한 Mod3.1 모사계산 결과는 Mod 1의 계산결과와는 달리 실험을 매우 훌륭히 모사함을 보이고 있다.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.559-564
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• 1998

The film condensation models in RELAP5/MOD3.1 and RELAP5/WOD3.2 are assessed with the data experiment performed in the scaled down condensation experimental facility with a single vertical tube inner diameter 46 mm in the range of pressure 0.1∼7.5 Mpa for the PSCS(Passive Secondary Condenser System) Both MOD3.1 and MOD3.2 don't shows any reliable predictions the experimental data The RELAP5/MOD3.1 overpredicts the heat transfer coefficients experiment, whereas the RELAP5/MOD3.2 underpredicts those data it is recommended that the film condonation model in RELAP5/MOD3.2 should be modified to hue a larger heat transfer coefficient than those the present model to give the reliable predictions.

• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.225-236
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• 1994

The sudden discharge of the loop seal water, which is present upstream of the SEBIM POSRV, creates large momentum and inertia forces on the downstream of the discharge piping system. This study provides the procedures and results of analysis of the thermal-hydraulic transient in the SEBIM POSRV discharge piping during the valve opening. The analysis is peformed by RELAP5/MOD3. The appropriate modeling of the discharge piping system, SEBIM POSRV opening characteristics, and loop seal water discharge for the RELAP5/MOD3 analysis is suggested. Also performed is the sensitivity study for the selection of proper options for the junction and volume control. flags. The analysis results demonstrate the adequacy of the RELAP5/HOD3 for the thermal-hydraulic transient analysis of the loop seal water discharge of the SEBIM POSRV discharge piping system. From the sensitivity analysis results, it is shown that the smooth area change option with reasonable geometric pressure drop distribution, non-equilibrium option, and proper time step should be selected for loop seal water discharge analysis.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.109-109
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• 1999

The standard RELAP5/MOD3.2 code was modified using the non-iterative modeling. which is developed to simulate steam condensation in the presence of noncondensable gases ill a vertical tube. The modified RELAP5/MOD3.2 code was used to simulate two kinds of vertical in-tube experiments involving the condensation phenomenon in the presence of noncondensable gases. The modeling capabilities of the modified RELAP5/MOD3.2 codc as well as the standard code for the condensation in the presence of noncondensable gases are assessed using two PCCS condensation experiments and four reflux condensation experimcnts. The modified RELAP5/MOD3.2 code gives good prediction over the data of both PCCS condensation and reflux condensation experiments

• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.258.1-258
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• 2001

• Nuclear Engineering and Technology
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• v.22 no.1
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• pp.12-19
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• 1990

The Loss of Offsite Power Transient at 77.5% power which occurred on June 9, 1981 at the Kori Unit 1 PWR (Pressurized Water Reactor) is simulated using the RELAP5/MOD2 system thermal-hydraulics computer code. Major thermal-hydraulic parameters are compared with the available plant data. The comparison of the analysis results with the plant data demonstrates that the RELAP5/MOD2 code has the capability to simulate the thermal-hydraulic behaviour of PWRs under accident conditions of this type with accuracy, except the pressurizer pressure and level. The pressurizer pressure increase is sensitive to the in surge now it is believed that the interracial heat transfer in a horizontal stratified flow regime may be estimated low and the compression effect due to insurge flow may be high. In the nodalization sensitivity study it is found that S/G noding with junctions between bypass plenum and steam dome is preferred to simulate the S/G water level decreasing and avoid the spurious level peak at trubine trip.

• Nuclear Engineering and Technology
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• v.30 no.5
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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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• v.31 no.5
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• pp.465-475
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• 1999

The objective of the present work is to assess the analysis capability of two wall film condensation models, the default and the alternative models, of RELAP5/MOD3.2 on condensation experiments in the presence of noncondensable gas in a vertical tube of PCCS of CP-1300. In the calculation of a base case the default model of RELAP5/MOD3.2 under-predicts the heat transfer coefficients, and Its alternative model over-predicts them throughout the condensing tube, Also, both models over-predict the void fractions. The nodalization study shows that the variation of the node number does not change both modeling results of RELAP5/MOD3.2 Sensitivity study for varying input parameters shows that the inlet steam-air mixture flow rate, the inlet air mass fraction, and the inlet saturated steam temperature give significant changes of their heat transfer coefficients Run statistics show that the grind time of the default model is always higher than that of the alternative model by about 23%.