[en] If high pressure safety injection pumps are not working for a total loss of feed water sequence, then this results in a severe accident inevitably. Nevertheless, a rapid depressurization using the POSRV could still mitigate the severe accident by providing a cooling water into a damaged core from safety injection tanks(SIT) which are passive systems. The purpose of this paper is to estimate how long the reactor vessel failure can be delayed by using the passive safety injection tanks for typical high pressure sequences at the KNGR. Based on MAAP calculation, the results show that the reactor vessel failure can be delayed about 7 hours if the inventory of four SITs is effectively used to remove the decay heat through the primary feed and bleed operation

[en] Sensitivity analyses for the in-vessel combustible gas generation, using the MELCOR program, are described in this report for the APR1400. The typical accident sequences of a station blackout and a large LOCA scenario are selected. A lower head failure model, a Zircaloy oxidation reaction model and a B₄C reaction model are considered for the sensitivity parameters. As for the base case, 1273.15K for a failure temperature of the penetrations or the lower head, an Urbanic-Heidrich correlation for the Zircaloy oxidation reaction model and the B4C reaction model are used. Case 1 used 1650K as the failure temperature for the penetrations and Case 2 considered creep rupture instead of penetration failure. Case 3 used a MATPRO-EG and G correlation for the Zircaloy oxidation reaction model and Case 4 turned off the B₄C reaction model. The results of the studies are summarized below : (1) Both the higher penetration failure temperature and the creep rupture failure model cause a large amount of hydrogen generation for a station blackout sequence. For a large LOCA, however, the hydrogen generation doesn't change much. (2) The MATPRO-EG and G correlation for a Zircaloy oxidation reaction results in less hydrogen generation than the Urbanic-Heidrich correlation (Base case) for both scenarios. (3) When the B₄C reaction model turns off, the amount of hydrogen decreases for two sequences. (4) The amount of CO, CO₂, and CH₄ do not significantly change for two sequences. It is found that the selection of hydrogen generation model affects fuel relocation characteristics inside the core include fuel channel blockage. Since it changes the amount of hydrogen generated from the core, a future study is recommended on this area

[en] The report contains the studies on a molten corium-concrete interaction, which has been recognized as important phenomena of severe reactor accident, where APR-1400 plant has been selected as a reference plant. The purposes of report are to review phenomenological models related to the molten corium-concrete interaction, and to investigate modelling uncertainties by performing sensitivity analysis, and finally to determine that the domestic design requirements of relevant phenomena can be satisfied. Concrete properties, corium amount, corium distribution in the reactor cavity, debris bed configuration, debris power, and heat transfer to the overlying coolant or concrete are considered as important uncertain parameters affecting basemat melt-through. The relevant code modelling have been reviewed and the effects of these parameters are studied through sensitivity analysis

[en] The superior in-pile performance of the HANA claddings have been verified by the successful irradiation test and in the Halden research reactor up to the high burn-up of 67GWD/MTU. The in-pile corrosion and creep resistances of HANA claddings were improved by 40% and 50%, respectively, over Zircaloy-4. HANA claddings have been also irradiated in the commercial reactor up to 2 reactor cycles, showing the corrosion resistance 40% better than that of ZIRLO in the same fuel assembly. Long-term out-of-pile performance tests for the candidates of the next generation cladding materials have produced the highly reliable test results. The final candidate alloys were selected and they showed the corrosion resistance 50% better than the foreign advanced claddings, which is beyond the original target. The LOCA-related properties were also improved by 20% over the foreign advanced claddings. In order to establish the optimal manufacturing process for the inner and outer claddings of the dual-cooled fuel, 18 different kinds of specimens were fabricated with various cold working and annealing conditions. Based on the performance tests and various out-of-pile test results obtained from the specimens, the optimal manufacturing process was established for the inner and outer cladding tubes of the dual-cooled fuel

[en] The objective of this paper is to establish Auxiliary FeedWater (AFW) operational technical bases to the Korean Next Generation Reactor (APR1400) by modeling the plant, and analyzing Station BlackOut (SBO) using the MELPCOR code. The severe accident phenomena at nuclear power plants have large uncertainties. For the integrity of the reactor vessel and containment safety against severe accidents, it is essential to understand severe accident sequences and to assess the accident progression accurately using computer codes. Furthermore, it is important to attain the capability to analyze the advanced nuclear reactor design for the severe accident prevention and mitigation. Accident analyses are also investigated how much AFW is effective to mitigate severe accident progresses. Nominal base case of SBO without AFW, time interval between feedwater stop and reactor vessel failure is 12,740 seconds. When 2, 4 and 8 hours of AFW operation considered to mitigate the accident progression, the reactor vessel failure is delayed for 20,415 seconds, 22,633 seconds and 26,508 seconds

[en] The ISAAC fission product release calculation is based on the detailed FPRAT models developed by Jaycor and the release of these materials from the core is governed by the release rate of fission products from the fuel matrix, the ability of the gas flow from the core to carry these materials to the rest of the primary system, and the saturated vapor pressure of fission product species given by chemical thermodynamic equilibrium. For volatile fission product release calculation, either Cubicciotti steam oxidation correlation or the NUREG-0772 correlation is used as user's options. In this study, sensitivity analyses are made for these volatile fission product release models. As the results, in case of early release, the IDCOR model with an in-vessel Te release option shows the most conservative results and for the late release case, NUREG-0772 model shows the most conservative results. Considering both early and late release, the IDCOR model with an in-vessel Te bound option is evaluated to show mitigated conservative results

[en] This report presents the results from six MELCOR calculations of Korean Next Generation Reactor (KNGR) LOCA that is one of the representative low-pressure sequences, the first is the base case calculation, and the other five include the sensitivity calculations. The MELCOR code is one of the currently available computer codes for the integral analysis of nuclear power plant severe accident progressions, which was developed at SNL for U.S. NRC. As the base case calculation for LOCA, this study assumes that the cold leg break of 0.5 ft2 under the fail-to-operation of all safety injection pumps (SIPs) and containment spray pumps (CSPs) and thus the safety injection is only made by the passive system of four SITs. Each of the five sensitivity parameters are then characterized by (a) the reduction of flow discharge coefficient via the break, (b) the operation of one CSP, (c) the hot leg break of 0.5 ft2, (d) the cold leg break of 0.6xDEG (Double-ended-guillotine), and (e) the cold leg break of 0.1 ft2. The other conditions of the sensitivity calculation are the same as those of the base case. As the response parameters for the resultant severe accident progressions, this study considers the timing of key phenomenological events expected in the reactor pressure vessel and thermo-chemical behavior of the molten core materials which enter containment. More specifically, they include (a) thermal-hydraulics of the core and reactor coolant system (RCS), (b) impact of the safety injection on the severe accident progressions, (c) distribution of thermal behavior of core material in the core and lower plenum, (d) molten corium-concrete interaction in the reactor cavity, and (e) containment pressure buildup before and after the reactor lower head failure.As the results of the aforementioned sensitivity calculations, it has been quantitatively identified that the variation of flow rate via the break is closely related with the timing of severe accident phenomenological events and pressure buildup rate in the containment before the lower head breach

[en] The modification of fthe MELCOR code, the integrated severe accident analysis program, has been performed for the heat transfer model between ex-vessel molten corium and overlying water pool. This model impacts on the corium-concrete interaction and the containment pressure behavior which are considered to be very important during severe accidents. Since the existing model do not consider debris particulation and water penetration in the ex-vessel debris cooling, the predicted heat flux is low compared to the measured value from the large scale experiments using real reactor materials. A dryout heat flux model has been employed in determining the heat removal from a debris bed by water penetration. Sensitivity analyses for debris particulate sizes and porosities are also performed and compared to MACE experiments

[en] Severe accident management strategies for Wolsong 2,3,4 Nuclear Power Plants are presented. The priorities of these strategies and entry conditions are also presented. The priorities are following; Injection into steam generator, Injection into PHTS, Injection into Calandria Vessel, Injection into Caladria Vault, Control Containment State, Control Fission Product Release, Control Hydrogen Concentration in the Containment. The criteria for the priority of strategy are the entry timing of each strategy and minimization of fission product release to environment. These severe accident management strategies and their priorities can be used to develop severe accident management guidance, which is part of severe accidenr management program, for Wolsong NPPs

[en] This paper introduce the desktop severe accident graphic simulator module (VMAAP) which is a window-based severe accident simulator using MAAP as its engine. The VMAAP is one of the submodules in SAMEX system (Severe Accident Management Support Expert System) which is a decision support system for use in a severe accident management following an incident at a nuclear power plant. The SAMEX system consists of four major modules as sub-systems: (a) Severe accident risk data base module (SARDB): stores the data of integrated severe accident analysis code results like MAAP and MELCOR for hundreds of high frequency scenarios for the reference plant; (b) Risk-informed severe accident risk data base management module (RI-SARD): provides a platform to identify the initiating event, determine plant status and equipment availability, diagnoses the status of the reactor core, reactor vessel and containment building, and predicts the plant behaviors; (c) Severe accident management simulator module (VMAAP): runs the MAAP4 code with user friendly graphic interface for input deck and output display; (d) On-line severe accident management guidance module (On-line SAMG); provides available accident management strategies with an electronic format. The role of VMAAP in SAMEX can be described as followings. SARDB contains the most of high frequency scenarios based on a level 2 probabilistic safety analysis. Therefore, there is good chance that a real accident sequence is similar to one of the data base cases. In such a case, RI-SARD can predict an accident progression by a scenario-base or symptom-base search depends on the available plant parameter information. Nevertheless, there still may be deviations or variations between the actual scenario and the data base scenario. The deviations can be decreased by using a real-time graphic accident simulator, VMAAP.. VMAAP is a MAAP4-based severe accident simulation model for OPR-1000 plant. It can simulate spectrum of physical processes occurring during accident including core heatup, cladding oxidation and hydrogen generation, core melt progression, vessel failure, fission product release, transport and deposition, and containment failure. Output results are displayed in user friendly graphical format by using text-based (numerical) output of MAAP program.. Window-based simulator of VMAAP is designed to provide graphical displays of the results during the transient simulation so that the users can easily follow the plant dynamics. Figure 1 through 4 show an example of VMAAP graphic display for the reactor coolant system, reactor vessel, containment building, and plotting of important parameters. VMAAP is able to simulate various scenarios very easily and quickly from the input deck of the scenario database of the SARDB. Since hundreds of input decks for severe core damage scenarios are available in SARDB, the simulation for a user-defined scenario can be performed very quickly by using a sub-module of VMAAP Input-editor which is a window-based MAAPspecific input deck generation program. VMAAP consists of following sub-modules: - System menu and tool bar - Project view - Event summary - Interactive control - Parameter help view - Input editor - Reactor vessel view - Reactor coolant system view - Containment building view The plant model used in VMAAP module is oriented to severe accident phenomena and thus it can simulate the in-vessel and ex-vessel behavior for a severe accident. Even though it may not be compatible with the desire to have a best-estimate analysis of an ongoing event, it can be a supporting or supplementary measure to understand the trends of accident progression