External Reactor Vessel Cooling (ERVC) is a key measure to realize molten corium In-Vessel Retention (IVR). The uncertainty of severe accident in core leads to the uncertainty of molten pool parameters in lower head, which may threat the reliability of successful implementation of IVR. In this study, an uncertainty analysis is employed to ensure the reliability of ERVC technology to IVR with the final two-layer molten pool configuration of AP1000. 600 simulations are performed to determine the effect of five key parameters. As long as parameters are within high probability density range, lower head can remain intact, and safety margin is greater than 15%. For influence level and correlation of parameters, the order on melt layer is: Decay heat > Fe mass > UO2 mass > Zr oxidation fraction > Emissivity of metal layer. The results can provide guidance for subsequent research direction and optimal design of mitigation measures.

In this study, the transient behavior of two-layer corium pool was studied based on COPRA facility. The transient change of two-layer corium pool caused by power change and the effect of stratification and stratified plate thickness on heat transfer characteristics were investigated. Then maximum heat flux ratios (MHFRs) of two-layer corium pools was proposed and investigated using CFD code FLUENT. The heat flux ratio is the ratio of heat flux to qCHF of vessel wall. The maximum of heat flux ratios at all angles is the MHFR of the corium pool. The MHFRs of two-layer corium pool under radiation top surface condition and isothermal top surface condition were obtained. By comparing the values and positions of the MHFRs, the degree of approaching boiling crisis of local reactor pressure vessel wall can be quantitatively analyzed. The results show that MHFR appears at metallic layer wall under radiation top surface condition and at upper part of the oxide layer wall under isothermal top surface condition. The results of this research can provide references for IVR (In-Vessel Retention) safety analysis and ERVC (External-Reactor Vessel Cooling) methods such as Nano-coating measure.

Since the accident at Fukushima, one major goal of reactor safety research has been the development of more accident tolerant technologies that can mitigate or delay fuel degradation during a Beyond Design Basis Accident (BDBA). One major effort has been focused on increasing the capability of the fuel to be more tolerant of damage during an accident, i.e., Accident Tolerant Fuel (ATF) materials. In this work, we present the development of a generic BWR plant model, the modification of MELCOR to model ATF materials and the use of ATF materials (specifically FeCrAl alloy) as a coating on Zircaloy cladding or as a substitute material for cladding and fuel assembly canister material and its effect on severe accident progression, specifically, a Station Blackout accident. The analysis indicates that significant fuel degradation via fuel heat-up, clad oxidation, and hydrogen generation was delayed up to an hour if FeCrAl alloy was used as a clad and canister material. And, combined with the passive safety systems (i.e., the Reactor Core Isolation Cooling system, RCIC), the extended operation of these systems delayed fuel degradation further. However, an adverse effect should be emphasized for the monolithic FeCrAl design-it generated more hydrogen than the designs based on the Zircaloy due to the high reaction rate at a high temperature of FeCrAl. The design of the FeCrAl-coated-Zircaloy avoids this defect. Therefore, it is a promising choice to combine some of the beneficial traits of both materials.

In recent years, micro-reactor concepts have attracted increasing attention in the nuclear industry due to the market demand for flexible, reliable, and sustainable power and heat on-site for industrial or federal installations or remote communities. To help demonstrate and validate these innovative reactor concepts, the Micro-reactor AGile Non-nuclear Experimental Test-bed (MAGNET) is being constructed at Idaho National Laboratory (INL) with an initial focus on the thermal and structural performance of heat pipe cooled micro-reactors. At this time, the preliminary design parameters of the MAGNET facility have been specified. In this work, a simulation using the System Analysis Module (SAM) code is performed for the prototypical 37-heat-pipes test article to predict its experimental facility performance and associated uncertainties. We first carry out a benchmark demonstration of our modeling method with an example provided by Argonne National Laboratory (ANL). Then, we predict the thermal performance of the MAGNET facility under steady-state operation. Moreover, several sensitivity parameters are analyzed to investigate their impact on facility thermal performance. This MAGNET experiment simulation provides valuable information for researchers to validate the facility's initial design and steady-state operation.

The deposition of protective coatings on nuclear fuel cladding has been considered as a near-term Accident Tolerant Fuel (ATF) concept that will reduce the high-temperature oxidation rate and enhance accident tolerance of the cladding while providing additional benefits during normal and transient. In this study, the performance of the proposed ATF concept of Cr-coated-Zircaloy is assessed using a generic Boiling Water Reactor MELCOR plant model considering a Short-term Station Blackout (STSBO) scenario. Simulation results indicate that the use of Crcoated-Zircaloy as cladding and canister material mitigates the core degradation process as compared to the traditional Zircaloy cladding and canister design. The onset of fuel degradation and collapse is delayed by over thirty minutes, and the extent of fuel degradation is reduced. Specifically, the gross in-vessel hydrogen generation decreased by almost a factor of three. Although the eutectic reaction between Cr-coating and Zircaloy could cause an early failure of the coating, the improvement in the delay of fuel degradation is still notable. Additionally, a thicker coating is found helpful to obtain additional coping time and to decrease hydrogen generation. In addition to the eutectic formation that may compromise Cr-coated Zr, a different failure mode is identified for the Cr-coated-Zr when compared to Zircaloy; i.e., a complete melt of base material leads to component collapse before the coating is oxidized and consumed. These findings can help the industry focus on productive areas of research and development for accident-tolerant fuel concepts and enhancement of core safety margins.

After the Fukushima nuclear accident, great attention is being paid to the late-phase behavior of the severe accident in the reactor, especially the molten corium-concrete interaction (MCCI). The MCCI process is very complex, involving physical and chemical reactions, heat and mass transfer processes. Many factors will affect the final result, with a high degree of uncertainty and there are still many unknowns based on current knowledge. In this paper, aimed at the MCCI phenomena in the containment cavity during a hypothetical accident in the nuclear power plant, an MCCI analysis code MOQUICO was developed by coupling the models of molten corium-concrete heat transfer, the concrete pyrolysis, and the corium cooling characteristics. The code MOQUICO was validated using the CCI-2, CCI-3 and SURC-2 tests with limestone-common sand concrete (LCS), siliceous concrete (SIL), and basaltic concrete (BAS), respectively. The simulant axial and radial ablation depths, upward heat flux and melt temperature agreed well with the experimental measurements. Afterwards, the code was used to simulate the typical PWR and BWR nuclear power plants. The ablation depth and gas production were analyzed for both simulations. Furthermore, the corium cooling was studied with eight different water injection moments for PWR and sensitivity analysis for decay heat and concrete type was conducted for BWR. The analysis proved that the developed code is capable of simulating MCCI and related phenomena of Light Water Reactor (LWR). (C) 2020 Elsevier Ltd. All rights reserved.

During core meltdown accidents, the heat flux distribution along the wall of lower plenum was closely related to the potential of vessel failure. Concerning with the natural convection heat transfer behavior in stratified corium pools, several two-layer tests were performed on COPRA facility recently. The test vessel (a 1/4 circular slice with an inner radius of 2.2 m and a width of 0.2 m) is divided into two parts by a plate made of 304 stainless steel. The lower part, heated by 18 electrical heating rods, is applied to simulate the oxide pool. The upper part without heat generation is assumed to be a metallic layer. During the experiment, different heating powers and boundary conditions were employed to obtain the temperature distributions of oxide pool and metal layer, as well as the heat flux through the vessel wall. Prototypic Ra' range, varied from 1.88 x 10(16) to 5.39 x 10(16), was successfully achieved in the oxide pool. Despite the smaller downward Nu(dn) numbers, the upward Nu(up) numbers within this study were well in accordance with the reported data. It should be mentioned that the results of water are unrepresentative for the reactor situation since no crust was formed in the bottom and top of oxide pool. However, the current study will still attract interest by the qualitative analysis of top cooling influence on natural convection phenomenon and provide valuable information to other researchers who concern with the crust formation and focusing effect on stratified corium pools behavior.

During a core meltdown accident, the significant decay heat within the corium pool will continually impose thermal load on the vessel wall and challenge the integrity of reactor pressure vessel (RPV). External Reactor Vessel Cooling (ERVC) is considered to be an effective strategy to reduce the potential of vessel failure. After long-term cooling, the initially uniform melt pool can reach a final steady state with stable two-layer or three-layer configuration. Based on the COPRA facility, a series of tests were performed to study the natural convection heat transfer phenomenon of stratified corium pools. The test section was a full-scale 2D slice, simulating the hemispherical lower head of the HPR1000 reactor design. Direct electrical heating was applied in the lower part of vessel to simulate the homogenously heated ceramic pool, while no heat was generated in the upper part to model the metallic layer. A binary mixture of 20 mol% NaNO3-80 mol% KNO3 was selected as simulant material due to the similar phase-change characteristics with the prototypical melt. During the experiment, effects of heating power, top boundary condition, corium pool height, decay heat distribution and simulant material on corium pool heat transfer behavior were analyzed. The results implied that the top cooling and decay heat distribution slightly influenced the temperature and heat flux of ceramic pool, but strongly impacted the characteristics of metallic layer. The comparison between the dimensionless parameters of water tests and salt tests also demonstrated the key role of simulant material. Furthermore, the downward heat transfer Nu(dn) and the upward heat transfer Nu, obtained from COPRA two-layer experiment were in accordance with the previous data or empirical correlations. Note that using the same simulant material for oxide layer and metal layer fails to repeat the boundary conditions in the reactor case, however, the current study pays more attention to the qualitative analysis of different influence on natural convection heat transfer behavior. More valuable information about the crust characteristics and stratification effect can be obtained by the following CFD simulation of these tests. (C) 2019 Elsevier Ltd. All rights reserved.

The Large Eddy Simulation (LES) methods were employed to analyze the turbulent heat transfer characteristics in detail for volumetrically heated corium pools. The main components and property parameters were calculated for the one-layer and two-layer configurations in the HPR1000 reactor. The numerical simulations were performed based on the WMLES turbulence model, phase change model, as well as the VOF model to describe thermal behaviors in the corium pool such as natural convection flow, pool temperature field and heat flux distribution. Then based on the Kelvin-Helmholtz (KH) instability theory, the calculated velocity vector difference between the two-layer flows at the interface was lower than the KH instability critical value. The high Rayleigh number natural convection in the bottom oxide pool was still not able to entrain the top metal layer. The molten salt (20 mol% NaNO3-80 mol% KNO3) and the Lead-Bismuth eutectic (44.5 wt% Pb-55.5 wt% Bi) were chosen as the appropriate simulant materials for the two-layer corium pool. The simulation results presented similar thermal characteristics compared with those from the prototypical reactor two-layer case, which could provide guidance for further research.

The COPRA experiments were performed to study the natural convection heat transfer behavior in a large-scale homogeneous melt pool inside the reactor pressure vessel lower plenum. The test section consists of a two-dimensional 1/4 circular slice with an inner radius of 2.2 m. A non-eutectic binary mixture 20%NaNO3-80%KNO3 was selected as melt simulant in the previous tests and the Rayleigh number of the melt pool reached up to 10(16). In this paper, the working fluid was a eutectic binary mixture 50%NaNO3-50%KNO3. The melt pool temperature, heat flux distribution and crust thickness were obtained in the experiments with different heating powers. Results from the eutectic molten salt tests can be applied for posttest calculations and comparative analyses.

In-vessel retention (IVR) has been proposed as an effective mitigation strategy during severe accidents in pressurized water reactors (PWRs). To help understand better the application of IVR, the large-scale COPRA facility was used for experimental research and the eutectic salt, i.e. the binary mixture of 50 mol% NaNO3 - 50 mol% KNO3 was selected as the simulant material for corium. Besides, numerical studies were done to capture the detailed flow field information. The experiment studied the effects of a sudden change in boundary condition and the influences of power changes on the transient behaviors and the steady-state characteristics of melt pool. Experimental results show that a sudden introduction of sidewall cooling and the melt pouring are the most threatening factors to the reactor vessel safety, but which can be guaranteed as long as the vessel is kept undamaged during melt pouring phases. Also, power changing won't lead to rapid variation of sideward heat flux, so the correlations for steady state heat transfer calculation are basically qualified to evaluate the reactor safety in analysis of long time cooling. Besides, the power changing only exerts obvious impacts on the crust growth on the curved sidewall inner surface with polar angle below 30 degrees, contrary to which the sideward heat flux is clearly influenced by different power densities only above the angle of 30 degrees. Finally, the numerical simulation results show that vortexes could form near the region where the crust formed, and lead to a more homogeneous temperature distribution; generally, the convection happens between the lower part of melt pool and cooling boundary, with flow velocity at a magnitude of 0.01 m/s.

During a severe accident in light water reactors, the core may melt and relocate into the lower plenum of the reactor pressure vessel. The decay heat will threaten the structural and thermal integrity of the reactor vessel if there is no effective cooling mechanism. Natural convection plays an important role in determining the thermal-hydraulic behaviors inside the debris pool, which is directly relevant to the problem of retention of molten corium inside the lower plenum. The COPRA (COrium Pool Research Apparatus) experiments were performed to study the natural convection heat transfer behavior in an internally heated melt pool with high Rayleigh numbers. The COPRA test facility is a two-dimensional 1/4 circular slice vessel with an inner radius of 2.2 m to simulate the lower plenum of reactor vessel for the Chinese advanced PWR in a full scale. The inner width of the slice is 20 cm and the curved vessel wall has a thickness of 30 mm. 20 electrical heating rods, each with a diameter of 16 mm but different lengths according to their locations, are uniformly distributed in the vessel to simulate homogenous internal decay heat. They can provide a maximum of 30 kW power to the melt pool. The outside of the curved wall is enclosed with a regulated external cooling path to keep the boundary temperature nearly isothermal. The top surface of the pool can be maintained insulated with an adiabatic lid. 79 thermocouples are installed in the melt pool to measure the melt pool temperature field and 48 in the curved wall to obtain local heat flux distribution along the curved wall. In the first series of experiments, water was employed as the simulant material. Due to the full scale geometry, the Rayleigh number within the pool could reach up to 1016, matching that in the prototypical situation for PWR. The heat transfer characteristic with volumetrically internal heat was investigated by using the full scale facility COPRA. Relations of pool temperature and heat flux distribution, as well as Nuch-Rai were developed. The results have been compared with the results and correlations from other experiments. (C) 2015 Elsevier Ltd. All rights reserved.

Based on test facility named COPRA , the natural convection heat transfer characteristics in corium pool inside the reactor pressure vessel lower plenum during severe accident were studied . The test apparatus was a two-dimensional 1/4 circular slice structure with an inner radius of 2.2 m to simulate the lower plenum of reactor pressure vessel at 1:1 scale for the Chinese independently-designed GEN-Ⅲ PWR .A non-eutectic binary mixture of 20% NaNO3-80% KNO3 (in mole fraction ) compositions was selected as the simulant material .The Rayleigh number within the pool could reach up to 1016 ,matching those in the prototypical situation for PWR .The influences of relocation position ,pool height ,power density and relocation times on pool temperature and heat flux distribution were studied in the experiments .The result shows that the Nu of the downward heat transfer from COPRA experiment is lower than those from other experiments within the same range of Rayleigh number .

Large scale COPRA experiments were performed to study the natural convection heat transfer in corium pools.Both water and a non-eutectic binary mixture of 20%NaNO3-80%KNO3 (in mole fraction) compositions were employed as the simulant material.The Rayleigh number within the pool could reach up to 1016,matching those in the prototypical situation for PWR.The comparisons of the pool temperature and heat flux distributions from water and molten salt experiments show that the crust formation along the inside curved wall is of significant impact to the heat transfer phenomenon in the melt pool.Heat transfer correlations were obtained from COPRA experiment and the heat flux distribution compared well with other experiments.The downward heat transfer towards the curved wall is lower than those from other experiments within the same range of Rayleigh number.

Large scale COPRA experiments were performed to study the natural convection heat transfer in corium pools inside the reactor pressure vessel lower plenum during severe accidents. The test facility is designed to simulate the lower plenum of reactor vessel at 1:1 scale for the Chinese advanced PWR, which is a two-dimensional 1/4 circular slice structure with an inner radius of 2.2 m. A non-eutectic binary mixture of 20 mol%NaNO3-80 mol%KNO3 compositions was selected as the simulant material. Twenty custom-designed heating rods were located in the facility to provide homogenous internal heating. Due to the full scale geometry, The Rayleigh numbers within the corium pools could reach to 1.188 x 10(15)-1.784 x 10(16), matching those in the prototypical situation during severe accident. The pool temperature field and heat flux distribution, as well as the crust thickness distribution from tests with different pool heights and heating powers were obtained and compared in the experiments. Influences of the relocation position and upper cooling on the heat transfer characteristics were also studied in the experiments. Relations of pool temperature and heat flux distribution, as well as Nu(dn)-Ra' relation were developed. Comparison with previous experiments showed that the downward heat transfer Nu(dn), prediction from COPRA experiments were in good agreement with SIMECO and LIVE salt experimental results. (C) 2015 Elsevier Ltd. All rights reserved.

During a severe accident in light water reactors, the melt of the core may relocate into the lower head of the reactor pressure vessel. Natural convection plays an important role in determining the thermal load from the debris pool, which is directly relevant to the problem of retention of molten corium inside the lower plenum. This paper presents the description of COPRA (COrium Pool Research Apparatus) experiments and results on natural convection heat transfer in an internally heated melt pool. The test apparatus is a two-dimensional 1/4 circular slice structure with an inner radius of 2.2 m to simulate the lower plenum of reactor vessel at 1:1 scale for the Chinese advanced PWR. A non-eutectic binary mixture of 20mol%NaNO3-80mol%KNO3 compositions is selected as the simulant material. The melt is preheated to about 350 in a custom-designed heating furnace before being transferred to the facility. 20 electrical heating rods, each with a diameter of 16 mm but different lengths according to their locations, are uniformly distributed in the vessel to simulate the homogenous internal decay heat. The top surface of the pool can be maintained insulated with an adiabatic lid. The outside of the curved wall is enclosed with a regulated external cooling path to keep the boundary temperature nearly isothermal. Due to the full scale geometry, the Rayleigh numbers within the pool could reach up to 10, matching those in the prototypical situation for PWR. The results have been compared with the correlations from other experiments.

During a severe accident in nuclear reactors, the core may melt and relocate into the lower head. The decay power in the accumulated melt pool will threaten the integrity of the reactor vessel if there is no effective cooling mechanism. Natural convection plays an important role in determining the heat flux distribution from the debris pool, which is directly relevant to the problem of retention of molten corium inside the lower plenum. Many experimental programs on natural convection heat transfer in corium pools have been performed under different conditions with multifarious research directions, thus a review work on this topic is necessary. Based on three different kinds of geometrical structures, this paper presents a comprehensive literature review of experimental methods and results. Influences of geometry, simulant material, heating method and layer stratification are also discussed. Conclusions and further research directions on natural convection heat transfer in corium pools are proposed.

To meet the domestic demand of software autonomous scheduling, Xi'an Jiaotong University had developed MIDAC (a Module In-vessel degraded severe accident Analysis Code) which can not only analyze the processes of in-vessel severe accident, but also provide the related results of each one. This code was composed of five modules: the early behavior module, the core degradation module, the debris bed module, the melting materials IVR module and the connecting module. In this paper, the basic mathematic-physical models of those modules were briefly introduced. Then, the CPR1000 station blackout scenario severe accident was set as an example to calculate the primary system thermal-hydraulic transient, the core degradation and the debris behavior, as well as the creep rupture of heat structures in primary loop pressure boundary. At last, the accuracy of MIDAC was verified in partial comparison with SCDAP/RELAP5 in primary system thermal-hydraulic transient analyzing part.