• 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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• 2001.05a
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• pp.178.2-178
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.163.1-163
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.191-191
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.172.2-172
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.135.1-135
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• 1999

• 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.

• 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.

• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.297-310
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• 1992

FLECHT SEASET unblocked forced reflood tests are assessed using Apollo version of RELAP5/MOD3 5M5. The main purpose of the study is to examine the code predictability under forced reflood conditions having different initial power levels and flooding rates. Among various test matrices, the assessment calculations are performed for the test numbers 31701 31302, 31203, 31805, 34524, 31021, 34006 and 35807 These have been selected because they have similar initial conditions but different initial peak rod powers or flooding rates. In addition, various sensitivity calculations are performed for test number 31203 on the improved models of RELAP5/MOD3. Those are for the effect of Counter Current Flow Limit (CCFL) option at the outlet junction of the test section, for the effect of grid modelling on the interfacial drag calculations as well as on the heat structure calculations, and for the effect of nodalization and the time step size. The results of sensitivity studies show that the improved models of RELAP5/MOD3 enhance the code predictability. The assessment results show that the RELAP5/MOD3 has a tendency to underpredict the turn around temperature and the turn around time. But RELAP5/MOD3 silghtly overpredicts the turn around temperature for high flooding rate. The results also show that the calculated quenching by RELAP5/MOD3 is delayed with the increase of the rod power or the decrease of the flooding rate.

• 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.