[en] The objective of the present work is to verify the prediction and analysis capability of MELCOR code about the progression of severe accidents in light water reactor and also to evaluate appropriateness of thermal-hydraulic models used in MELCOR code. Comparing the results of experiment and calculation with MELCOR code is carried out to achieve the above objective. Specially, the comparison between the CORA-13 experiment and the MELCOR code calculation was performed

[en] The objectives of the present work are to verify MELCOR code and to suggest reasonable bases for the application of this code to the evaluation procedure of severe accidents in a nuclear power plant. To do this, thermal-hydraulic models used in the MELCOR code were investigated and the calculation results were compared with the available experimental results. In the present work, lET series were analysed to verify the DCH models in the MELCOR and CONTAIN code by calculating lET-I, IR, 3, 4 and 7 test. The calculation results of the MELCOR code for the temperature in the subcompartment, regardless of the existence of the oxygen and hydrogen within the SURTSEY vessel, show similar results with the experimental results and for the pressure in the SURTSEY vessel show that IET-3, 4 and 7 test having the oxygen and hydrogen were overestimated by the MELCOR code more than those of the CONTAIN code. Also the calculation results of the MELCOR code show abrupt increase of temperature in the subcompartment in the latter time of accident in case of IET-7 having the oxygen within the SURTSEY vessel and the calculation results of the CONTAIN code show abrupt increase of temperature in the latter time of accident in cases of IET-3, 4 and 7. From the calculation results of the MELCOR and CONTAIN code for IET series, it is known that the calculation results of the CONTAIN code show faster progression of accident more than those of the MELCOR code. Also it is known that the temperature of subcompartment is overestimated by the CONTAIN code and the pressure in the SURTSEY vessel is overestimated by the MELCOR code in cases of IET series having the oxygen and hydrogen by comparing the calculation results of the both codes with the experimental results. In the future work, more exact analyses will be carried out by mutually offsetting the weak points of the both codes and by applying the results of analyses to a nuclear power plant

[en] In order to provide a technical base for evaluation of the applicability of NUREG-1465 source term, which the domestic nuclear industry has tried to use for the KNGR design as well as other cases, postulated large break LOCA-induced core-melt scenarios, not accompanying recovery through core reflooding, have been analyzed for the KSNP. For this purpose the MELCOR code (Version 1.8.4) was used to estimate the characteristics and the quantity of fission product release. Sensitivity analysis for some parameters were done. Comparison of these results with the realistic source term showed that the durations of fission product release into containment during the design basis accident phases were longer than those of NUREG-1465 and the release fractions of alkali metals, halogens and the tellurium group were bigger than those of NUREG-1465. Further supporting calculations for this study and additional sensitivity analyses, e.g. for the accident sequences terminated by injection of ECCS, are required