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[en] The air ingress accident is one of the postulated design basis accidents, a guillotine-type break of the main pipes connecting to the reactor vessel, which is still unclear if the present HTGRs can maintain a passive safe function for this type of break. In order to analyze the air ingress accident, we developed a multi-dimensional multi-component mixture analysis code (GAMMA) and investigated chemical reaction and thermo-fluid behaviors related to the accident. GAMMA includes the models to address the important physical phenomena: multi-component molecular diffusion, bulk and surface chemical reactions, and heat transport by conduction, convection and thermal radiation. Since the period of the transient is very long, about one month, the efficient analysis tool is necessary in order to get a solution numerically stable and computationally fast. Therefore, for fast code run, we adopt the Implicit Continuous Eulerian (ICE) technique which reduces a 10N x10N matrix to an NxN pressure matrix. A concerned complex system can be configured by the linkage of a 1-D calculation module and a 2/3-D calculation module. In order to verify and validate the GAMMA code, we assessed various experiments and benchmark problems on the chemical reaction and heat removal behaviors: molecular diffusion tests, graphite oxidation tests, air-ingress tests, pebble-bed heat removal tests, and reactor cavity cooling system performance tests. The calculation results of the GAMMA code are in a high level of agreement with the experimental data as well as those of the other analysis codes used for the conventional HTGRs. From the air ingress analyses for the reference gas cooled reactors, PBMR and GT-MHR, significant rise in fuel temperature is observed for an assumed large air volume but the peak fuel temperature is predicted below the fuel failure criterion (1600 .deg. C). Sensitivity analysis on the air volumes in a vault and the onset timings of natural convection shows that the major parameters affecting on the severity of air ingress are the air concentration in a vault, the natural convection flow rate, and the bottom reflector temperature. In particular, if the air supply into the core is limited by isolating the reactor cavity immediately following the break, the consequence of air ingress would be mitigated enough to maintain the fuel and internal structure integrity during the accident
[en] Tritium is generated by a ternary fission reaction in TRISO coated fuel particles as well as by neutron absorption reactions with 6Li, 10B, 3He, etc, which are present in core graphite of VHTR and helium coolant, respectively. Since VHTR uses helium gas as its coolant, tritium can be readily diffused to other components in the system. In addition, because of tritium's higher diffusivity even in unbroken materials, the behavior in the VHTR system needs to be treated more importantly in the licensing point of view for its construction. In particular, the tritium activity concentration in the product hydrogen and in the components of the hydrogen plant should be adequately evaluated in order to confirm that regulatory requirements can be satisfied. In this respect, KAERI has developed tritium behavior analysis code for VHTR, which contributes to put a basis for the quantitative analysis of tritium behavior in a variety of environments of the VHTR system
[en] This report describes the preliminary thermalhydraulic analysis of HTR-10 steady state full power initial core to provide a benchmark calculation of VHTGR(Very High-Temperature Gas-Cooled Reactors) safety analysis code of GAMMA(GAs Multicomponent Mixture Analysis). The input data of GAMMA code are produced for the models of fluid block, wall block, radiation heat transfer and each component material properties in HTR-10 reactor. The temperature and flow distributions of HTR-10 steady state 10 MWth full power initial core are calculated by GAMMA code with boundary conditions of total reactor inlet flow rate of 4.32 kg/s, inlet temperature of 250 .deg. C, inlet pressure of 3 MPa, outlet pressure of 2.992 MPa and the fixed temperature at RCCS water cooling tube of 50 .deg C. The calculation results are compared with the measured solid material temperatures at 22 fixed instrumentation positions in HTR-10. The wall temperature distribution in pebble bed core shows that the minimum temperature of 358 .deg. C is located at upper core, a higher temperature zone than 829 .deg. C is located at the inner region of 0.45 m radius at the bottom of core centre, and the maximum wall temperature is 897 .deg. C. The wall temperatures linearly decreases at radially and axially farther side from the bottom of core centre. The maximum temperature of RPV is 230 .deg. C, and the maximum values of fuel average temperature and TRISO centreline temperature are 907 .deg. C and 929 .deg. C, respectively and they are much lower than the fuel temperature limitation of 1230 .deg. C. The comparsion between the GAMMA code predictions and the measured temperature data shows that the calculation results are very close to the measured values in top and side reflector region, but a great difference is appeared in bottom reflector region. Some measured data are abnormally high in bottom reflector region, and so the confirmation of data is necessary in future. Fifteen of twenty two data have a relative error less than ±20% and six of them only show good predictions less than ±10% error. For more accurate calculation of GAMMA code in future, the application of cone-shaped core model, the consideration of heat generation and fluid block in fuel discharging tube and the application of the improved geometrical model in bottom reflector zone are suggested
[en] The probability of unexpected Station Blackout after Fukushima is becoming the challenging issue to provide electric power for a long period of time to minimize the damages to the property and the person. High Temperature Gas Cooled Reactor have been studied by considering the passive safety, the system simplicity and the high reliability between 100kWt ∼ 20MWt. 1MWth movable nuclear reactor(MAGMA) among them is selected as a future R and D objective with regard to fuel cycle, power generation and high reliability. This study is in phase to develop the concept of small-scale reactor to provide emergent power. Therefore, further study should be conducted to specify the MAGMA with related to system design, analysis, component test and integral demonstration test
[en] Tritium is a radioactive isotope of hydrogen with the half life of 12.32 years. Effect of radiation from tritium on health is relatively small and hazard only if it is taken into the body, because tritium decays by emitting a low energy(18.6 keV) beta particle with no gamma radiation. However, it has been proven that an internal exposure by the tritium through inhalation or drinking water can cause serious damage to human bodies and that the effectiveness of tritiated water (HTO or T2O) is 25000 times the effectivity of tritium (HT or T2). In a Very High Temperature Gas Cooled Reactor (VHTR), neutron interactions with boron in control rods, impurities in the graphite core and 3He isotopes in helium coolant gas as well as ternary fissions produce tritium under reactor operating conditions. Compared to the pressurized heavy or light water reactors, tritium in VHTR circulates and diffuses more readily because of the gaseous coolant. Since tritium can penetrate the structure barriers with its high diffusion coefficients even in intact material conditions, tritium should be treated as an important part of the regulation of a VHTR. Recently, the code developments to evaluate quantitatively tritium behavior in high temperature gas cooled reactors has been performed; for examples, TRITGO code (General Atomics), THYTAN code (JAEA), TPAC code (Idaho National Laboratory), etc. Efforts to develop the mechanistic models and to increase the reliability of the analysis results have also been made in world-leading countries in the field of gas cooled nuclear reactors such as USA and Japan. In KAERI, a tritium behavior analysis code (TRIBAC) for a VHTR is under development, which will be verified and validated in this series of studies
[en] Predicting radioactive fission product (FP) behaviors in the reactor coolant system and the containment of a nuclear power plant (NPP) is one of the major concerns in the field of reactor safety, since the amount of radioactive FP released into the environment during the postulated accident sequences is one of the major regulatory issues. Radioactive FPs circulating in the primary coolant loop and released into the containment are basically in the form of gas or aerosol. In this study, a multi-component and multi-sectional analysis module for aerosol fission products has been developed based on the MAEROS model, and the aerosol transport model has been developed and verified against an analytic solution. The deposition of aerosol FPs to the surrounding structural surfaces is modeled with recent research achievements. The developed aerosol analysis model has been successfully validated against the STORM SR-11 experimental data, which is International Standard Problem No. 40. Future studies include the development of the resuspension, growth, and chemical reaction models of aerosol fission products.
[en] GAMMA-FP (GAs Multicomponent Mixture Analysis-Fission Products module), consists of gaseous and aerosol fission product analysis modules. The aerosol FP module adopts a multi-component and multi-sectional aerosol analysis model that has been developed based on the MAEROS model. For the first work of FP module development, the MAEROS model has been implemented and examined against some analytic solutions and experimental data by Yoo et al. An aerosol transport model was developed and implemented in the GAMMA-FP code, and verified. In this study, the aerosol deposition model in the GAMMA-FP code was improved by adopting recent achievements, and was validated against an experimental data available. The aerosol deposition model in the GAMMA-FP code has been improved and successfully validated against the STORM SR-11 deposition test. The simulation with the improved deposition model predicted the matched results with the experimental data well. For future studies, the aerosol deposition model by flow irregularities will be implemented and validated against the TRANSAT bend effect test
[en] One of the unique features of a Very High Temperature Gas Cooled Reactor (VHTR) is Vented Low Pressure Containment (VLPC) containing two separate vent paths where both have two gravity operated relief valves in a series. Because VLPC strategy allows the release of a relatively small amount of radioactive fission products(FP) into the environment during the blowdown phase, behavior analyses of the fission products circulating in the primary coolant loop and in the containment are major consideration factors for safety evaluation. For thermal-fluid analysis of a Very High Temperature Gas Cooled Reactor (VHTR), the GAMMA(GAs Multicomponent Mixture Analysis)+ code is under development. The multi-component analysis modules of aerosol fission products has been developed based on the MAEROS model and the aerosol transport model of the CONTAIN code. Furthermore, Yoo et al. developed and incorporated gaseous fission product transport phenomena into the GAMMA+ code. In this study, the phase change of fission products between condensed and vapor forms are modeled by thermochemical equilibrium, and the thermochemical and physical data of 288 species and 26 elements are implemented as a form of look-up table in the FP module of the GAMMA+ code
[en] Sponsored by OECD/NEA, a benchmark study on the prismatic coupled neutronics/thermal fluids transient for the MHTGR-350 core was initiated in 2012. The benchmark consists of three phases, i.e., steady-state (Phase 1), transient-state (Phase 2) and depletion problems (Phase 3). Phase 1 has three exercises. Exercise 1 deals with neutronics stand-alone calculation whereas Exercise 2 focuses on thermo-fluid stand-alone calculation. The coupled simulation is defined in Exercise 3. Korea Atomic Energy Research Institute (KAERI) is participating in Phase 1 Exercise 2 using the GAMMA+ code. The present paper summarizes the GAMMA+ model and results of Phase 1 Exercise 2 which contains the steady-state thermo-fluid simulation of the MHTGR-350 core. In this work, the GAMMA+ modeling and calculation results were presented for Phase 1 Exercise 2 of the OECD/NEA MHTGR-350 benchmark. Four cases were considered with two different grid size models. The results of the GAMMA+ calculations were found to be reasonable. The results were submitted to the benchmark organizer. The international comparison is on-going and the final OECD/NEA publication will be issued later.
[en] The accurate prediction of local hot spot during normal operation is important to ensure core thermal margin in a very high temperature gas-cooled reactor because of production of its high temperature output. The active cooling of the reactor core determining local hot spot is strongly affected by core bypass flows through the inter-column gaps between graphite blocks and the cross gaps between two stacked fuel blocks. The bypass gap sizes vary during core life cycle by the thermal expansion at the elevated temperature and the shrinkage/swelling by fast neutron irradiation. This study is to investigate the impacts of the variation of bypass gaps during core life cycle as well as core restraint mechanism on the amount of bypass flow and thus maximum fuel temperature. The core thermo fluid analysis is performed using the GAMMA+ code for the PMR200 block-core design. For the analysis not only are some modeling features, developed for solid conduction and bypass flow, are implemented into the GAMMA+ code but also non-uniform bypass gap distribution taken from a tool calculating the thermal expansion and the shrinkage/swell of graphite during core life cycle under the design options with and without core restraint mechanism is used