• 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

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.257.1-257
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• 2001

• Nuclear Engineering and Technology
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• v.27 no.6
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• pp.859-869
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• 1995

The present study is to understand the physical phenomena anticipated during the accident with RHR loss under mid-loop operation in a PWR and, at the same time, to examine the prediction capability of RELAP5/MOD3.1 on such an accident, by simulating an integral test relevant to this accident for reliable analysis in an actual PWR. The selected experiment, i.g. BETHSY Test 6.9a, represents the configuration with the pressurizer manway open and steam generators unavailable during the accident. Accordingly, the results of this ok are sure to contribute to understanding both the key events as well as the sensitive parameters, anticipated in the accident, for validity of the actual analysis. In the simulation result overall behavior as well as major phenomena observed in the experiment have been predicted reasonably by RELAP5/MOD3.1, however, the problem associated with enormous computing time .due to small time step size has been encountered. Besides, the code prediction of higher swollen level in the pressure vessel has given rise to overestimation of both pressurizer level and RCS pressure. Subsequently, overprediction of the break flow through the manway has led to earlier core uncovery than that in the experiment by about 400 seconds. As a whole, it is demonstrated from both the experiment and the analysis that gravity feed has not been sufficient to recover the core level and thus additional forced feed has been necessary in this configuration.

• 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.50 no.8
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• pp.1412-1420
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• 2018

An experiment was performed using the large-scale test facility (LSTF), which simulated a 1% vessel upper head small-break loss-of-coolant accident with an accident management (AM) measure under an assumption of total-failure of high-pressure injection (HPI) system in a pressurized water reactor (PWR). In the LSTF test, liquid level in the upper head affected break flow rate. Coolant was manually injected from the HPI system into cold legs as the AM measure when the maximum core exit temperature reached 623 K. The cladding surface temperature largely increased due to late and slow response of the core exit thermocouples. The AM measure was confirmed to be effective for the core cooling. The RELAP5/MOD3.3 code indicated insufficient prediction of primary coolant distribution. The author conducted uncertainty analysis for the LSTF test employing created phenomena identification and ranking table for each component. The author clarified that peak cladding temperature was largely dependent on the combination of multiple uncertain parameters within the defined uncertain ranges.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.531-536
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• 1996

계통분식 코드인 RELAP5/MOD3와 격납용기 분석 코드인 CONTEMPT4/MOD5의 통합코드를 구성하였다. 두 코드는 process제어의 개념을 이용하여 가능한 한 코드의 수정을 최소화하고 각 코드의 특성을 유지시키면서 explicit coupling되게 하였다. 통합코드를 간단한 피동형 계통 분석에 적용시켜 계통과 격납용기의 계산을 동시에 수행함으로써 얻을 수 있는 격납용기-계통 간의 열수력 현상을 파악 할 수 있게 하여줌으로써 피동형 원전의 열수력 분석도구로서 사용할 수 있음을 검증하였다.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.403-410
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• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.591-598
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• 1997

The COBRA/RELAPS code, an integrated version of the COBRA-TF and RELAP5/MOD3 codes, has been developed for the realistic simulations of complicated, multi-dimensional, two-phase, thermal-hydraulic system transients in light water reactors. Recently, KAERA developed an unified version of the COBRA/RELAP5 code, which can run in serial mode on both workstations and personal computers. This paper provides the brief overview of the code integration scheme, the recent code modifications, the developmental assessments, and the future development plan.

• Nuclear Engineering and Technology
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• v.31 no.3
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• pp.344-363
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• 1999

A multi-dimensional thermal-hydraulic system analysis code, MARS 1.3.1, has been developed in order to have the realistic analysis capability of two-phase thermal-hydraulic transients for pressurized water reactor (PWR) plants. As the backbones for the MARS code, the RELAP5/MOD3.2.1.2 and COBRA-TF codes were adopted in order to take advantages of the very general, versatile features of RELAP5 and the realistic three-dimensional hydrodynamic module of COBRA-TF. In the MARS code, all the functional modules of the two codes were unified into a single code first. Then, the source codes were converted into the standard Fortran 90, and then they were restructured using a modular data structure based on "derived type variables" and a new "dynamic memory allocation" scheme. In addition, the Windows features were implemented to improve user friendliness. This paper presents the developmental work of the MARS version 1.3.1 including the hydrodynamic model unification, the heat structure coupling, the code restructuring and modernization, and their verifications.their verifications.

• Journal of Mechanical Science and Technology
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• v.20 no.7
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• pp.1030-1042
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• 2006

An experimental study has been performed to investigate the off-take phenomena at the header-feeder systems (horizontal header pipe with multiple feeder branch pipes) in a CANDU (CANadian Deuterium Uranium) reactor with the branch orientation varies ${\pm}36^{\circ}\;or\;{\pm}72^{\circ}$. In order to evaluate the applicability of the conventional correlations used in the safety analysis code, RELAP5-Mod3, the test facility is designed with the 1/2 scale of the. CANDU 6. It was found that the data set for the top, bottom and side branches are in a good agreement with the correlations used. However, for the specific angled branches, ${\pm}36^{\circ}\;and\;{\pm}72^{\circ}$, the onsets of off-take data and quality data showed large deviation with the conventional model inside RELAP5-MOD3. Furthermore, based on the uncertainty analysis, the conventional 2.5 power law needs to be modified. The present experimental data set can be useful for the construction of the general correlation considering the arbitrary branch orientation.