• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.243-251
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• 1992

Behavior of loop seal formation and clearing during small break loss of coolant accident is investigated using the RELAP5/MOD 2 and /MOD3 codes with the test of SB-CL-18 of the LSIF (Large Scale Test Facility）. The present study examines the thermal-hydraulic mechanisms responsible for early core uncovery including the manometric effect due to an asymmetric coolant holdup in the steam generator upflow and downflow side. The analysis with the RELAP5/MOD2 demonstrates the main phenomena occuring in the depressurization transient including the loop seal formation and clearing with sufficient accuracy. Nevertheless, several differences regarding the evolution of phenomena and their timing have been pointed out in かe base calculations. The RELAP5/MOD3 predicts overall phenomena, particularly the steam generator liquid holdup better than the RELAP5/MOD2. The nodalization study in the components of the steam generator U-tubes and the cross-over legs wiか the RELAP5/MOD3 results in good prediction of the loop seal clearing phenomena and their timing.

• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.585-590
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• 1997

The condonation models in the standard RELAP5/MOD3.2 code are assessed and improved based on the database, which is constructed from the previous experimental data on various condonation phenomena The default model the laminar film condonation in RELAP5/MOD3.2 does not give any reliable predictions, and its alternative model always predicts higher values than the experimental data Therefore, it is needed to develop a new correlation based on the experimental data of various operating ranges in the constructed database. The Shah correlation, which is used to calculate the turbulent film condensation heat transfer coefficients in the standard RELAP5/MOD3.2, well predicts the experimental data in the database. The horizontally stratified condonation model of RELAP5/MOD3.2 overpredicts both cocurrent and countercurrent experimental data The correlation proposed by H.J.Kim predicts the database relatively well compared with that of RELAP5/MOD3.2 The RELAP5/MOD3.2 model should use the liquid velocity for the calculation of the liquid Reynolds number and be modified to conifer the effects of the gas velocity and the film thickness.

• Nuclear Engineering and Technology
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• v.34 no.5
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• pp.455-467
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• 2002

RELAP5/MOD3.2 was modified to perform the thermal-hydraulic safety analysis for HANARO transients. Several aspects of RELAP5/MOD3.2 were modified or replaced by new features to properly simulate the unique HANARO characteristics such as the finned fuel element, the cooling mechanisms by both plate type heat exchanger and the natural circulation. Especially, the heat transfer packages were modified to be more appropriate for the safety analysis and the heat transfer models were developed for the plate type heat exchanger as well as natural circulation through the pool water. This modified version of RELAP5/MOD3.2 is renamed as RELAP5/HANARO. The thermal-hydraulic simulations of the single fuel pin test and plate type heat exchanger were peformed to assess the realistic predicting capabilities of RELAP5/HANARO and compared with experimental results and manufacturer＇s data in this paper. In addition, the natural circulation experiment using the scaled bundle was simulated to validate the capability of RELAP5/HANARO. The simulation results show almost similar trend with experimental data. Therefore, it is proved that RELAP5/HANARO has a confidence to use for the safety analyses of HANARO.

• Nuclear Engineering and Technology
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• v.35 no.5
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• pp.426-441
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• 2003

A reactor inlet header break experiment, B9401, performed in the RD-14M multi channel test facility was analyzed using RELAP5/MOD3.2 and RELAP5/CANDU[1]. The RELAP5 has been developed for the use in the analysis of the transient behavior of the pressurized water reactor. A recent study showed that the RELAP5 could be feasible even for the simulation of the thermal hydraulic behavior of CANDU reactors. However, some deficiencies in the prediction of fuel sheath temperature and transient behavior in athe headers were identified in the RELAP5 assessments. The RELAP5/CANDU has been developing to resolve the deficiencies in the RELAP5 and to improve the predictability of the thermal-hydraulic behaviors of the CANDU reactors. In the RELAP5/CANDU, critical heat flux model, horizontal flow regime map, heat transfer model in horizontal channel, etc. were modified or added to the RELAP5/MOD3.2. This study aims to identify the applicability of both codes, in particular, in the multi-channel simulation of the CANDU reactors. The RELAP5/MOD3.2 and the RELAP5/CANDU analyses demonstrate the code's capability to predict reasonably the major phenomena occurred during the transient. The thermal-hydraulic behaviors of both codes are almost identical, however, the RELAP5/CANDU predicts better the heater sheath temperature than the RELAP5/MOD3.2. Pressure differences between headers govern the flow characteristics through the heated sections, particularly after the ECI. In determining header pressure, there are many uncertainties arisen from the complicated effects including steady state pressure distribution. Therefore, it would be concluded that further works are required to reduce these uncertainties, and consequently predict appropriately thermal-hydraulic behaviors in the reactor coolant system during LOCA analyses.

• 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.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.143-143
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• 1999

• 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%.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.178.2-178
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• 2001