[en] The header of CANDU reactor is an important component to simulate the fuel channel behavior because the headers' hydraulic behavior controls the feeder void fraction which affects on the fuel bundle coolability. In CANDU accident analyses, the liquid entrainment and vapor pull-through (off-take) phenomena should be considered when horizontal stratification achieved inside the header. The current RELAP5 off-take model can treat only 3 directions; vertical upward, downward, and side oriented junctions. The RELAP5 off-take model was modified and generalized by considering the geometric effect of branch angles. Based on the previous experimental results, the critical height correlation was reconstructed by use of the branch line connection angle. The new model in RELAP5/CANDU could be applied to vertical upward, downward and angled branch. The verification and validation analyses for the new model were performed using Separate Effect Test (SET) and Integral Effect Test (IET). The verification and validation analyses show improved accuracy with the new model. As an application, the Wolsong nuclear power plant 35% RIH break was analyzed and the results showed consistent behavior compared to the analysis results of AECL

[en] This work is devoted to the technique application of lock-in infrared thermography in the shipbuilding and ocean engineering industry. For this purpose, an exploratory study to find the optimized test conditions is carried out by the design of experiments. It has been confirmed to be useful method that the phase contrast images were quantified by a reference image and weighted by defect hole size. Illuminated optical intensity of lower or medium strength give a good result for getting a phase contrast image. In order to get a good phase contrast image, lock-in frequency factors should be high in proportion to the illuminated optical intensity. The integration time of infrared camera should have been inversely proportional to the optical intensity. The other hand, the difference of specimen materials gave a slightly biased results not being discriminative reasoning

[en] In a postulated core melt accident, if the molten core is not retained in-vessel despite taking severe accident mitigation actions, the core debris will relocate in the reactor cavity region. There, it will interact with structural concrete and could potentially result in basemat failure (through erosion or overpressurization) and consequent fission product release to the environment. Although a methodology of cooling the molten core by adding water on its top is selected as a severe accident management strategy in case molten core is released outside a reactor vessel, the possibility of a long-term cooling is still unresolved. In the OECD/MCCI project scheduled for 4 years from 2002. 1 to 2005. 12, a series of tests are being performed to secure the data for cooling the molten core spread out at the reactor cavity and for the long-term CCI (Core Concrete Interaction). The tests include not only separate effect tests such as a melt eruption, water ingression, and crust failure tests with prototypic material but also 2-D CCI tests with a prototypic material under dry and flooded cavity conditions. In this study, SWICCS-4, one of the SWICCS tests of OECD/MCCI project, was assessed using MELCOR1.8.5 and the computer code deficiencies were investigated including several sensitivity studies

[en] In a postulated core melt accident, if the molten core is not retained in-vessel despite taking severe accident mitigation actions, the core debris will relocate in the reactor cavity region. There, it will interact with structural concrete and could potentially result in basemat failure (through erosion or overpressurization) and consequent fission product release to the environment. Although a methodology of cooling the molten core by adding water on its top is selected as a severe accident management strategy in case molten core is released outside a reactor vessel, the possibility of a long-term cooling is still unresolved. In the OECD/MCCI project scheduled for 4 years from 2002. 1 to 2005. 12, a series of tests are being performed to secure the data for cooling the molten core spread out at the reactor cavity and for the long-term CCI (Core Concrete Interaction). In this study, SWICCS-1, 2 and 3, of the SWICCS tests of OECD/MCCI project, was assessed using MELCOR1.8.5 and the sensitivity parameters' selection of CAV package were investigated in SWICCS-1 and the same options were applied to another experiments SWICCS-2 and -3

[en] The KINS Reactor Thermal-hydraulic Analysis System (KINS-RETAS) under development is directed toward a realistic analysis approach of best-estimate (BE) codes and realistic assumptions. In this system, MARS is pivoted to provide the BE Thermal-Hydraulic (T-H) response in core and reactor coolant system to various operational transients and accidental conditions. As required for other BE codes, the qualification is essential to ensure reliable and reasonable accuracy for a targeted MARS application. Validation is a key element of the code qualification, and determines the capability of a computer code in predicting the major phenomena expected to occur. The MARS validation was made by its developer KAERI, on basic premise that its backbone code RELAP5/MOD3.2 is well qualified against analytical solutions, test or operational data. A screening was made to select the test data for MARS validation; some models transplanted from RELAP5, if already validated and found to be acceptable, were screened out from assessment. It seems to be reasonable, but does not demonstrate whether code adequacy complies with the software QA guidelines. Especially there may be much difficulty in validating the life-cycle products such as code updates or modifications. This paper presents the plan for MARS validation, and the current implementation status

[en] There have been many activities to evaluate the possibility of recriticality, such as Fred et al., Monsteller and Rahn, EPRI report, Darnowski et al., and etc. However, those activities are focused on BWR. Moreover, release of fission products, which can increase the reactivity is not considered properly. In this study, we perform an evaluation of possibility for recriticality on a PWR assembly loaded in APR1400 if unborated water is injected during a severe accident. The evaluation is done by Monte Carlo method with explicit modeling of the assembly geometry. We also perform a MELCOR simulation to select a geometry with high possibility of recriticality and to calculate remaining fractions of fission products to estimate the inventories during a severe accident. In this paper, using the Monte Carlo method, evaluations of the possibility for recriticality were performed on a fuel assembly in PWR during the reflooding phase of a severe accident. The geometric and isotopic configurations of the fuel assembly were selected from the MELCOR simulation on LBLOCA scenario. Inventories of fission products were obtained from the ORIGEN-ARP calculations and the MELCOR simulation. In this study, we assumed that the absorber material in the control rods is B4C. However, some of reactors in Korea such as Kori Unit 2, etc. use Ag-In-Cd as absorber materials which has a lower melting point than the temperature for eutectic mixture formation of B₄C and cladding of control rods. Therefore, we expect that there will be more severe case for recriticality evaluations. As future works, we will perform sensitivity studies on the control rod types. We will also perform the evaluations on the whole core of the reactor. From the whole core analysis, we will develop a model to analyze the impact of the recriticality in whole core/part of the core on the progression of a severe accident.

[en] It is difficult for the approximated solutions to apply when the mass concentrations of aerosol particles change with time. Moreover, it is also difficult to solve the system explicitly, since the system is very stiff due to large differences between magnitudes of mean charges and those of the number densities of ions. The differences are on the order of ~1020. In order to enhance such drawbacks, we develop a calculation module using a concept of separate calculations between mean charges of each section of aerosol size and the ion concentrations. Then, we apply the module to analyze mean charge distributions of Cs and I-131 aerosols to check if it is appropriate for the assumption on the sticking efficiency in the conventional calculations to be used in aerosol dynamics.We compare the results with those of the approximated steady-state solutions given by Clement and Harrison as well. In this study, we developed the calculation module for mean charge distributions of the radioactive aerosol particles in order to perform stable and accurate calculations when the mass concentrations of the aerosol particles changes with the time. Using the module, we performed the analyses to check if it is appropriate for the assumption on the sticking efficiency in the conventional calculations to be used in aerosol dynamics. If such aspects on sticking efficiency are considered in the calculation of aerosol dynamics, coagulations and settling rates will be different from those used in the conventional calculations. Therefore, as a future work, the module will be coupled with a calculation module of the aerosol dynamics to analyze behaviors of aerosol particles in the containment. The results of the analyses will be compared with those from conventional calculations. In addition, in order to calculate the charge distributions for the various radioactive aerosols, construction of database on the properties for other radioactive aerosols will be done as well.

[en] IVR-ERVC is selected as a Severe Accident Management Strategy for APR1400. Design Requirements: • To provide flexibility in severe accident management; • To provide an additional function for in-vessel retention by using the existing systems; • To flood the external surface of reactor vessel lower plenum before the relocation of molten corium. Implementation: • Initial flooding of the reactor vessel using a shutdowm colling pump (SCP); • Reactor insulation design for effective water intrusion and flow; • Supplementary water injection by boric acid makeup pump (BAMP) to compensate boiling-out.

[en] In this paper, we select important fission products for the estimation of the source term during a severe accident of a PWR. The selection is based on the numerical results obtained from depletion calculations for the typical PWR fuel via the in-house code named DEGETION (Depletion, Generation, and Transmutation of Isotopes on Nuclear Application), release fractions of the fission products derived from NUREG-1465, and effective dose conversion coefficients from ICRP 119. Then, for the selected fission products, we obtain the adjoint solutions of the Bateman equations for radioactive decay in order to determine the importance of precursors producing the aforementioned fission products via radioactive decay, which would provide insights into the assumption used in MACCS 2 for a level 3 PSA analysis in which up to six precursors are considered in the calculations of radioactive decays for the fission product after release from the reactor

[en] 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. 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