[en] The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code is being developed at the INEL under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. NRC. As The current time, the SCDAP/RELAP5/MOD3.1 code is the result of merging the RELAP5/MOD3 and SCDAP models. The code models the coupled behavior of the reactor coolant system, core, fission product released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. Major purpose of the report is to provide information about the characteristics of SCDAP/RELAP5/MOD3.1 core T/H models for an integrated severe accident computer code being developed under the mid/long-term project. This report analyzes the overall code structure which consists of the input processor, transient controller, and plot file handler. The basic governing equations to simulate the thermohydraulics of the primary system are also described. As the focus is currently concentrated in the core, core nodalization parameters of the intact geometry and the phenomenological subroutines for the damaged core are summarized for the future usage. In addition, the numerical approach for the heat conduction model is investigated along with heat convection model. These studies could provide a foundation for input preparation and model improvement. (author). 6 refs., 3 tabs., 4 figs

[en] The main purpose of this report is to introduce essential features and functions of a severe accident risk information management system, SARD (Severe Accident Risk Database Management System) version 1.0, which has been developed in Korea Atomic Energy Research Institute, and database management and data retrieval procedures through the system. The present database management system has powerful capabilities that can store automatically and manage systematically the plant-specific severe accident analysis results for core damage sequences leading to severe accidents, and search intelligently the related severe accident risk information. For that purpose, the present database system mainly takes into account the plant-specific severe accident sequences obtained from the Level 2 Probabilistic Safety Assessments (PSAs), base case analysis results for various severe accident sequences (such as code responses and summary for key-event timings), and related sensitivity analysis results for key input parameters/models employed in the severe accident codes. Accordingly, the present database system can be effectively applied in supporting the Level 2 PSA of similar plants, for fast prediction and intelligent retrieval of the required severe accident risk information for the specific plant whose information was previously stored in the database system, and development of plant-specific severe accident management strategies

[en] This report contains analysis methodologies and calculation results of loss of offsite sequences for the severe accident analysis database system. The Korean standard nuclear power plant has been selected as a reference plant. Based on the probabilistic safety analysis of the corresponding plant, Twelve accident scenarios, which was predicted to have more than 10-10 /ry occurrence frequency, have been analyzed as base cases for the loss of offsite sequence database. The functions of the severe accident analysis database system will be to make a diagnosis of the accident by some input information from the plant symptoms, to search a corresponding scenario, and finally to provide the user phenomenological information based on the pre-analyzed results. The MAAP 4.06 calculation results of loss of offsite sequence in this report will be utilized as input data of the severe accident analysis database system. This report updates and complements a previously published Technical Report

[en] This report contains analysis methodologies and calculation results of station blackout sequences for the severe accident analysis database system. The Korean standard nuclear power plant has been selected as a reference plant. Based on the probabilistic safety analysis of the corresponding plant. Eight accident scenarios, which was predicted to have more than 10^-10/ry occurrence frequency have been analyzed as base cases for the station blackout sequence database. Furthermore, the sensitivity studies for operational plant systems and for phenomenological models of the analysis computer code have been performed. The functions of the severe accident analysis database system will be to make a diagnosis of the accident by some input information from the plant symptoms, to search a corresponding scenario, and finally to provide the user phenomenological information based on the pre-analyzed results. The MAAP 4.06 calculation results of station blackout sequence in this report will be utilized as input data of the severe accident analysis database system

[en] In a recent international meeting (SNL, MCAP-2003) which has introduced current MELCOR (severe accident analysis computer code) development activities, existing MELCOR core release models failed to predict ISP46 test (international fuel behavior experiments) results well and the latest ORNL-Booth model was suggested to be an optimum model. The need for this kind of model update has been proposed also in the domestic FPT-1 test analysis. Therefore, this research analyzes releases of selected representative volatile and non-volatile radionuclides during conservative high and low pressure sequences in the APR1400 plant using three core release models (CORSOR, CORSOR-M, CORSOR-Booth) in the latest MELCOR 1.8.5 version and the ORNL-Booth update model. As a research strategy, the difference in release fractions among existing and update models is compared and the uncertainty range is then evaluated. The MELCOR fission product core release calculations, which are based on the CORSOR models developed by Battelle Memorial Institute, are performed using various CORSOR empirical release correlations. These correlations assume that fission products are released from the fuel-cladding gap when a failure temperature criterion exceeds or intact geometry is lost and their release rates are based on fuel temperatures. In the analysis, the option of the fuel component surface-to-volume ratio in the CORSOR and CORSOR-M models and the option of the high and low burn-up in the CORSOR-Booth model are considered together. In addition, the update model simulates the effect of Molybdate (Cs₂MoO₄) compound, which is known to be created from the combination of Cs and Mo species during core release, by modifying vapor pressure data. As the results, the CORSOR/CORSOR-M release rate is high for volatile radionuclides, and the CORSOR release rate is high for non-volatile radionuclides with insufficient consistency. As the uncertainty range for the release rate expands from several times (volatile radionuclides) to more than maximum 10,000 times (non-volatile radionuclides), user's careful choice for the core release models is needed. In the trend, the updated ORNL-Booth model is similar with the CORSOR-M model for volatile and mid-volatile radionuclides while it is similar with the CORSOR-Booth model for non-volatile radionuclides. From these, the ORNL-Booth model is recommended for volatile and mid-volatile radionuclides based on the results of the ISP-46 international test and this research, though an optimum model is not fixed until now. On the contrary, no specific model is recommended for non-volatile radionuclides due to insufficient and inconsistent results. This research is the first domestic results evaluating the core release models in the commercial nuclear plants other than in the experiments and is expected to be used as a reference for the source term uncertainty analysis in the domestic nuclear plants

[en] MELCOR executes in two parts. The first is a MELGEN program, in which most of the input is specified, processed, and checked. The second part of MELCOR is the MELCOR program itself, which advances the program through time based on the database generated by MELGEN and any additional MELCOR input. In particular, MELCOR execution involves two steps: (1) a setup mode in MEXSET, during which the database is read from the restart file and any additional input is processed, and (2) a run mode in MEXRUN, which advances the simulation through time, updating the time-dependent portion of the database each cycle. MELGEN and MELCOR share a structured and modular architecture that facilitates the incorporation of additional or altenative phenomenological modes. This structure consists of four primary levels: executive level, database manager routine level, package level, and utility level. MELCOR is composed of 24 different packages, each of which models a different portion of the accident phenomenology or program control. To identify the relation of the MELCOR subroutines with the packages, first two or three letters of the package's name are duplicated in the name of the subroutines. The same rule applies to the naming of the common block. Data flows and the specific subroutines in the MELGEN and MELCOR are analyzed by their functions according to the hierarchy of four levels for model improvement and replacement during the integral code development project

[en] This report contains analysis methodologies and calculation results of loss of offsite sequences for the severe accident analysis database system. The Korean standard nuclear power plant has been selected as a reference plant. Based on the probabilistic safety analysis of the corresponding plant, twelve accident scenarios, which was predicted to have more than 10^-10 /ry occurrence frequency, have been analyzed as base cases for the loss of offsite sequence database. The functions of the severe accident analysis database system will be to make a diagnosis of the accident by some input information from the plant symptoms, to search a corresponding scenario, and finally to provide the user phenomenological information based on the pre-analyzed results. The MAAP calculation results of loss of offsite sequence in this report will be utilized as input data of the severe accident analysis database system

[en] An ISAAC computer code, which was developed for a Level-2 PSA during 1995, has developed mainly with fundamental models for CANDU-specific severe accident progression and also the accident-analyzing experiences are limited to Level-2 PSA purposes. Hence the system nodalization model, decay model and volatile fission product release model, which are known to affect fission product behavior directly or indirectly, are evaluated to both enhance understanding for basic models and accumulate accident-analyzing experiences. As a research strategy, sensitivity studies of model parameters and sensitivity coefficients are performed. According to the results from core nodalization sensitivity study, an original 3x3 nodalization (per loop) method which groups horizontal fuel channels into 12 representative channels, is evaluated to be sufficient for an optimal scheme because detailed nodalization methods have no large effect on fuel thermal-hydraulic behavior, total accident progression and fission product behavior. As ANSI/ANS standard model for decay heat prediction after reactor trip has no needs for further model evaluation due to both wide application on accident analysis codes and good comparison results with the ORIGEN code, ISAAC calculational results of decay heat are used as they are. In addition, fission product revaporization in a containment which is caused by the embedded decay heat, is demonstrated. The results for the volatile fission product release model are analyzed. 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 shows mitigated conservative results

[en] As an ISAAC computer code, which was developed for a Level-2 PSA during 1995, has mainly fundamental models for CANDU-specific severe accident progression and also the accident-analyzing experiences are limited to Level-2 PSA purposes, the core heat transfer model, break flow model, emergency core cooling system model and containment dousing spray /local air cooler model are evaluated to enhance understanding for basic models and to accumulate accident-analyzing experiences. Sensitivity studies using model parameters and sensitivity coefficients are performed. According to the results from AECL experiments and code analyses for core heat transfer model, it was found that one representative fuel rod for the actual 37 fuel rods did not cause serious temperature discrepancies during the severe accident progression. The results from emergency core cooling system model, shows a good comparison with the FSAR. As the results of the evaluation, it was found that local air coolers could control containment pressure whether dousing spray is operating or not, and their operation does not cause containment failure. Regarding the dousing system, it could control containment pressure as long as it is operating and its operating time depends on containment conditions. For a large LOCA sequence without local air coolers, spray works only for 1.2 hours and delays containment failure by 13 hours compared to the no spray case. According to the test results, the ISAAC models for local air coolers show a consistent trend for steam removal. As ISAAC could model local air coolers only at two locations at present, future work is planning to generalize the locations for local air coolers

[en] In this report, the effect of external vessel cooling on debris coolability and vessel integrity for the KNGR were examined from the two typical pressure range of high(170 bar) and low(5 bar)case using the lower plenum model in MELCOR1.8.4. As the conditions of these calculations, 80 ton of debris was relocated simultaneously into the lower vessel head and the debris relocation temperature from the core region was 2700 K. The decay heat has been assumed to be that of one hour after reactor shutdown. The creep failure of the vessel wall was simulated with 1-D model, which can consider the rapid temperature gradient over the wall thickness during the ex-vessel cooling. From the calculation results, both the coolant temperature and the total amount of coolant mass injected into the cavity are known to be the important factors in determining the time period to keep the external vessel cool. Therefore, a long-term strategy to keep the coolant temperature subcooled throughout the transient is suggested to sustain or prolong the effect of external vessel cooling. Also, it is expected that to keep the primary side at low pressure and to perform the ex-vessel flooding be the essential conditions to sustain the vessel integrity. From MELCOR, the penetration failure always occurs after relocation regardless of the RCS pressure or availability of the external vessel cooling. Therefore, It is expected that the improvement of the model for the penetration tube failure will be necessary