In-vessel Retention Research Articles

Numerical Study on the Effect of Jet Cross Section Shape on the Corium Jet Breakup Behavior With Lattice Boltzmann Method

In a core meltdown accident in light water reactors, molten corium may drop into the lower plenum of the pressure vessel and interact with water, which is called fuel–coolant interaction (FCI). The behavior of the corium jet breakup in water during FCIs is important for the in-vessel retention strategy and has been extensively studied. While in previous studies, the jet cross-section shapes are naturally assumed to be circular, which is actually not always the case, in this study, the breakup processes of the corium jets with four different elliptical cross-section shapes and three different penetration velocities are simulated with color-gradient lattice Boltzmann method. The effect of the cross-section shape on the hydrodynamic breakup behavior of the corium jet is analyzed in detail. It is found that the effect of the cross-section shape on the jet penetration depth is very limited. With the increase in the aspect ratio under the same penetration velocity, the jet breakup length decreases gradually. In general, the dimensionless corium surface area increases with the increase in the aspect ratio for the jets under the same penetration velocity.

Transient heat transfer analysis of double-layer molten pool under severe accident

ABSTRACT Reactor safety is crucial to the development of nuclear energy. In a severe accident, after the core of a nuclear reactor melts and collapses, the use of in-vessel retention external reactor vessel cooling (IVR-ERVC) technology can effectively alleviate the heat flux at the lower head and ensure the intactness of the lower head. The different densities of the melt result in stratification in the molten pool (neglect thermochemical effects and two layers are immiscible). The molten pool structure is investigated based on single-, double-, and triple-layer molten pool structures. Based on a third-generation nuclear reactor, the ANSYS commercial software is used to establish a three-dimensional geometric model of the double-layer melt. A large eddy simulation is performed to comprehensively analyze the turbulent heat transfer characteristics of the molten pool with a bulk heat source. Wall-Modeled Large eddy simulation (WMLES), Volume of Fluid (VOF), and phase change models were used in this work, for example, nature convection, temperature field, and heat flux distribution were all investigated. The temperature field, velocity field, and heat transfer of the vessel outer wall are compared for two different boundary conditions of top radiation and top cooling on the light metal layer. The transient heat transfer characteristics of the molten pool under IVR-ERVC conditions are investigated to provide a basis for further investigations.

Chemical Interactions of Molten Steel with (U-Zr-O) Corium: Main Outcomes of the CORDEB Program

One of the strategies to mitigate the consequences of a severe accident is to stop the progression of molten materials (corium) within the vessel by external cooling of the vessel. This strategy is called in-vessel retention (IVR). The safety demonstration of IVR requires the accurate evaluation of the thermal load imposed on the internal surface of the vessel by the corium debris. During the progression of the accident, solid debris may be generated initially in the lower plenum because of fragmentation in water and, after melting of those debris, a (possibly stratified) molten pool would be formed. The consequences of the thermal and chemical interactions of the corium with the vessel are difficult to predict because there are still uncertainties on the application of existing results to models for reactor cases. These include scaling effects of course, but also the influence of transient processes occurring simultaneously: oxidation, dissolution, melt progression and/or stratification. In order to be able to predict the heat fluxes through the RPV wall, it is necessary to establish an overall view of the possible transient evolutions of corium and determine which are the most probable ’steady-state’ corium configurations after debris melting. The CORDEB program studies corium phenomena taking place at different stages of molten pool formation in the reactor vessel. It may be considered as an investigation of corium configurations that were not considered in previous programs such as the OECD RASPLAV and MASCA programs. All experiments within the program are carried out in the Rasplav tests facility of NITI, using the technology of induction melting in a cold crucible. This program is co-funded by IRSN, EDF, AREVA and CEA in the frame of the CPSIN agreement. This paper presents the main conclusions drawn from CORDEB tests, both for phenomenological knowledge and for transposition to reactor situations. Quantitative elements are provided for a clearer picture of kinetic effects. CORDEB tests have shown the importance of chemical interactions between molten steel and solid sub-oxidized (U-Zr- O) corium which lead to penetration of steel inside the solid oxide (crust or debris) and to the partial dissolution of the solid. As a consequence, the crust surrounding a molten pool cannot be considered as a barrier to mass transfer. CORDEB tests also made it possible to assess the oxidation kinetics of a molten pool: in the presence of surface crust, this is a rather slow process and it does not have a significant effect on the deterioration of accident scenario. The gaseous atmosphere above the pool plays an important role in several processes, in particular in the evaporation of molten steel.

Modeling of IVR and Sensitivity Analysis of Cavity Injection System of a 1000MWe Advanced PWR

In-Vessel Retention (IVR) is one of the primary technique to mitigate severe accidents. This enables retaining the corium inside the Reactor Pressure Vessel (RPV) through external cooling. However, it is important to assess the validity and success of this technique, though actual validation is impossible. In this work, we have developed a model of selected case study and used MATLAB 2018b to simulate it. Developed model has been tested using two sources; constant temperature source and constant heat flux. These two sources were applied at the internal surface of RPV separately and heat transfer through RPV lower plenum was monitored. Main purpose of this analysis is to obtain temperature profiles along the RPV lower plenum external surface. Results show that the technique is successful because external surface temperature is lower than melting point of RPV in both the cases discussed. Sensitivity analysis showed that external temperature can be further decreased by increasing heat transfer, which can be changed by thickness of RPV and sub-cooling of inlet water. However, both parameters have their own limitations. Results obtained through MATLAB simulations are also comparable to MELCOR1.8.5 code simulation results and both found in agreement.

Numerical evaluation of geometric variations in internally heated melt pool

In Vessel Retention (IVR) is considered as one of the best safety strategy for a hypothetical severe accident in light water nuclear reactors (LWR). Since the strategy involves heat transfer between the corium melt pool and surrounding, it is highly impacted by the natural convection behaviour of such internally heated melt pool. Extensive experiments and numerical study has been performed worldwide to understand the behaviour of corium melt pool. COPRA (COrium Pool Research Apparatus) is one such experiment which uses molten salts of non-eutectic binary mixture (20 mol% NaNO3–80 mol% KNO3) and eutectic binary mixture (50 mol% NaNO3–50 mol% KNO3) as simulant material. Geometric variation in COPRA experiment is investigated to get a better understanding of the natural convection behaviour of the melt pool. Although the influence of geometry is reported in literature, the increase in geometry beyond the conventional size is yet to be explored. Eutectic salt (50 mol% NaNO3–50 mol% KNO3) is used as a simulant material for this study. A commercial CFD code ANSYS FLUENT 20.2 is employed for the numerical modelling of the corium pool. K-epsilon (k-ε) turbulence model and solidification-melting model are used in this simulation. Experimental data is used for validating this model. The geometric effect is investigated by assuming the same height and radius for the pool. Also, the ratio of heating power and cooling length is maintained for all the variations of geometry. The temperature field, liquid fraction contour, and velocity distribution of the melt pool are calculated. This paper provides an extension to the existing data thus further enhances the understanding of the thermal–hydraulic characteristics of the corium melt pool.

Core and safety design for France–Japan common concept on sodium-cooled fast reactor

France (CEA and FRAMATOME) and Japan (JAEA, MFBR, and MHI) teams have carried out collaborative works to have common technical views regarding a sodium-cooled fast reactor (SFR) concept. This paper mainly describes the capabilities of ASTRID 600 to demonstrate SFR technologies of both countries. Japan has studied the feasibility of an enhanced high burnup low-void effect (CFV) core and fuel using oxide dispersion-strengthened steel cladding in ASTRID 600, considering Japan’s target for core performance. The neutronics design of the core has satisfied most required design targets and conditions. Regarding passive shutdown capabilities, Japan team has performed a preliminary numerical analysis for ASTRID 600 using a complementary safety device (CSD), called a self-actuated shutdown system (SASS), one of the safest approaches in Japan. Japan’s team used the SASS in place of the hydraulically suspended absorber rod, called RBH, one of the safest approaches of France, to investigate the potential of the SASS as a design measure common to the countries. The preliminary analysis has shown that the SASS can satisfy the countries’ main requirements. This study has also revealed that the mitigation measures of ASTRID 600 against a severe accident are promising to achieve in-vessel retention for both countries.

Experimental research of heat transfer in heavy metallic layer under different top boundary conditions

The External Reactor Vessel Cooling (ERVC) is crucial to realizing In-Vessel Retention (IVR) strategy when reactor core degradation happens; consequently, fission products would be limited in Reactor Pressure Vessel (RPV). Because of the fission products in the lower head of RPV, vessel integrity is threatened. Due to the immiscibility of oxide and metal materials, the molten pool in the lower head of RPV would turn to a stratified configuration. Two-layer (light metallic layer atop of oxidic layer) and three-layer (light metallic layer, oxidic layer, heavy metallic layer) configurations have been proposed. The possible heavy metallic layer has a different shape, top boundary conditions, and decay heat compared with the other two layers. Thereby, the objective of the heavy metallic layer experiment is to study its heat transfer characteristics, and two possible top boundary conditions are performed. The heating cables simulating the decay heat are placed in a hemispheric test section (diameter = 2.4 m). Water is used as the experimental simulant with a height of 0.3 m. The experimental results demonstrate that increasing the top cooling rate, the ratio of top cooling power to sidewall cooling power, and the ratio of upward heat flux to sideward heat flux increase. In addition, they share almost the same trend. The central melt temperature rises with the height increment, and the maximum normalized temperature reaches about 1.3 for both top cooling and insulated boundary conditions. Thermal stratification is found in the experiment, and the possible natural convection is limited in the heavy metallic layer. Correlations of UCLA, Mini-ACOPO, and RASPLAV reasonably predict the heavy metallic layer's normalized sideward heat flux distribution.

Numerical Simulation and Validation of Debris Bed Self-Leveling Behavior with Mixed-Density Particles Using CFD-DEM Coupling Algorithm

For the Japan Sodium-cooled Fast Reactor (JSFR), should the hypothesized core disruptive accident (CDA) happened, the in-vessel retention (IVR) will be the main target to achieve. In the heat-removal phase of the CDA, the debris bed will be piled up on the debris catcher. The capability of stable cooling and avoiding recriticality on the debris bed will be the main issues for achieving IVR. Previous studies have shown that the homogeneous debris bed can attain stable cooling and eliminate the probability of recriticality. Besides, self-leveling, which is a mechanism redistributing and flattening the debris bed by the natural circulation or vaporization from surrounding coolant, can further suppress the debris bed’s thickness to below the coolable thickness. However, in the real situation, the debris bed is composed of mixed-density debris particles. Hence, when these mixed-density debris particles start to redistribute due to self-leveling, the debris bed will form a heterogeneous density distribution. Under this scenario, the capability of coolability and the probability of recriticality could deviate from the previous study. Therefore, it is necessary to obtain a verified coupled model between the computational fluid dynamics (CFD) and the discrete element method (DEM) to track the mixed-density debris particles’ movement under the phenomenon of self-leveling. In this paper, first, the experiments simulating self-leveling on the mixed-density particle bed are performed. Afterward, the random heavy particle movement’s experimental data are extracted and transformed into the statistics form as the benchmark materials. Finally, the CFD-DEM model is validated via a series of sensitivity studies. The verified CFD-DEM can be expected to simulate the self-leveling behavior on the mixed-density debris bed and the real reactor case.

Influence of crust formation on the heat load to a reactor vessel under an in-vessel retention condition

The heat load imposed on a reactor vessel was measured with and without the crust layer under an IVR-ERVC condition. Mass transfer experiments were performed based on heat and mass transfer analogy to achieve high RaʹH of order 1015 with compact test rigs. Geometric configuration and thermal boundary condition of the oxide pool due to crust formation were modeled. The results showed larger Nuup and lower Nudn than existing studies, which is due to the large Pr of the current experimental condition. The angular heat flux to the vessel increased monotonously with the angle regardless of the crust formation. Once the crust layer is formed, the thermal boundary condition near the vessel becomes melting temperature of the crust, which reduces the system temperature difference. The decreased buoyancy results in the decrease in both upward and downward heat transfers but the variation of the crust thickness did not show notable effect.

Monitoring of tritium and impurities in the first wall of fusion devices using a LIBS based diagnostic

Laser-induced breakdown spectroscopy (LIBS) is one of the most promising methods for quantitative in-situ determination of fuel retention in plasma-facing components (PFCs) of magnetically confined fusion devices like ITER and JET. In this article, the current state of understanding in LIBS development for fusion applications will be presented, based on a complete review of existing results and complemented with newly obtained data. The work has been performed as part of a research programme, set up in the EUROfusion Consortium, to address the main requirements for ITER: (a) quantification of fuel from relevant surfaces with high sensitivity, (b) the technical demonstration to perform LIBS with a remote handling system and (c) accurate detection of fuel at ambient pressures relevant for ITER. For the first goal, the elemental composition of ITER-like deposits and proxies to them, including deuterium (D) or helium (He) containing W–Be, W, W–Al and Be–O–C coatings, was successfully determined with a typical depth resolution ranging from 50 up to 250 nm per laser pulse. Deuterium was used as a substitute for tritium (T) and in the LIBS experiments deuterium surface densities below 1016 D/cm2 could be measured with an accuracy of ∼30%, confirming the required high sensitivity for fuel-retention investigations. The performance of different LIBS configurations was explored, comprising LIBS systems based on single pulse (pulse durations: ps–ns) and double pulse lasers with different pulse durations. For the second goal, a remote handling application was demonstrated inside the Frascati-Tokamak-Upgrade (FTU), where a compact, remotely controlled LIBS system was mounted on a multipurpose deployer providing an in-vessel retention monitor system. During a shutdown phase, LIBS was performed at atmospheric pressure, for measuring the composition and fuel content of different area of the stainless-steel FTU first wall, and the titanium zirconium molybdenum alloy tiles of the toroidal limiter. These achievements underline the capability of a LIBS-based retention monitor, which complies with the requirements for JET and ITER operating in DT with a beryllium wall and a tungsten divertor. Concerning the capabilities of LIBS at pressure conditions relevant for ITER, quantitative determination of the composition of PFC materials at ambient pressures up to 100 mbar of N2, the D content could be determined with an accuracy of 25%, while for atmospheric pressure conditions, an accuracy of about 50% was found when using single-pulse lasers. To improve the LIBS performance in atmospheric pressure conditions, a novel approach is proposed for quantitative determination of the retained T and the D/T ratio. This scenario is based on measuring the LIBS plume emission at two different time delays after each laser pulse. On virtue of application of a double pulse LIBS system, for LIBS application at N2 atmospheric pressure the distinguishability of the spectra from H isotopes could be significantly improved, but further systematic research is required.

Experimental and numerical study on multiaxial creep behavior of 16MND5 steel at 700℃

In-vessel retention (IVR) is an effective strategy of ensuring the structural integrity of RPV under severe accidents. In the IVR condition, due to the geometric discontinuity of pressure vessel, multiaxial creep is the main form of failure of RPV. Moreover, 16MND5 steel undergoes a phase transition from bainite to austenite. In order to study the multiaxial creep failure mechanism, creep tests on different notched bar are carried out at 700℃. The influence of grain boundary precipitation on creep behaviors was studied by micro-structural observation. Furthermore, based on the Kachanov-Robotnov continuous damage model (CDM), the term of activation energy Q is adopted in the newly-developed model as the correction parameter that is used to study the damage evolution and multiaxial creep behaviors of 16NND5 steel. Finally, the good agreement is achieved between theoretical prediction and experimental results.

Evaluation of nanofluids for enhancing critical heat flux for in-vessel retention during sever reactor accidents

ABSTRACT In-vessel retention (IVR) of a molten core through external cooling of the reactor vessel is a severe accident management strategy at nuclear power plants. The enhancement of critical heat flux (CHF) is one of the measures to improve the success probability of IVR. A nanofluid, in which nanoparticles are dispersed in a liquid, is known to reliably enhance CHF, mainly under pool boiling conditions. However, to apply nanofluids to IVR, it is necessary to clarify the effect of CHF enhancement under convection boiling similar to that in an actual IVR event. Therefore, CHF tests were conducted using a rectangular test section 50 mm wide × 50 mm deep × 600 mm long under conditions simulating severe accidents. From the CHF data obtained for various nanofluid parameters under typical severe accident conditions, each parameter’s effect on CHF enhancement was clarified. It was found that CHF was improved by 50% or more by a nanofluid as compared with deionized water. Furthermore, the CHF of nanofluid obtained in the tests simulating the actual severe accident conditions increased from 30% to 50% compared with that of deionized water, suggesting that CHF can be enhanced by nanofluids even in an actual plant.

Mechanistic critical heat flux prediction for in-vessel retention conditions

In-vessel retention (IVR) is considered a feasible technique to keep reactor pressure vessel (RPV) integrity in a severe reactor accident. For light water reactor (LWR), the effectiveness of this strategy relies soundly on the critical heat flux (CHF) distribution over the external surface of the lower plenum of RPV, whose orientation varies gradually from downward-facing horizontal to vertical. The CHF prediction capability of the liquid sublayer dryout model is efficient for high mass flux in vertical flow boiling conditions. This paper focuses on how to adapt the model to the changed orientation of the heating surface. Bubble departure diameter (dB) and net vapor generation point (NVG), the starting point for the void fraction developing in a heating channel, is one of the important key points in the CHF prediction. Therefore, to assess the predictive potential of CHF under IVR, experimental research was performed to measure bubble departure diameter and NVG for a changing heating surface orientation from downward-facing horizontal to vertical with a mechanistic model basing on the force balance. A modified liquid sublayer dryout model was then proposed where the channel orientation effect is considered to measure the bubble departure diameter (vapor blanket diameter) using the improved force balance model. The NVG is modified according to the departure diameter. The predicted departure diameter and subcooling at NVG show good consistency with the experimental data, and the modified liquid sublayer dryout model can predict the CHF data with an average relative error of 18.36% in IVR.

Critical heat flux in a CANDU end shield – Influence of shielding ball diameter

Experiments were performed to measure the critical heat flux (CHF) on a vertical surface abutting a coarse packed bed of spherical particles. This geometry is representative of a CANDU reactor calandria tubesheet facing the end shield cavity during the in-vessel retention (IVR) phase of a severe accident. Deionized light water was used as the working fluid. Low carbon steel shielding balls with diameters ranging from 6.4 to 12.7 mm were used, allowing for the development of an empirical correlation of CHF as a function of shielding ball diameter. Previously published data is used to develop a more comprehensive empirical correlation accounting for the impacts of both shielding ball diameter and heating surface height. Tests using borosilicate shielding balls demonstrated that the dependence of CHF on shielding ball thermal conductivity is insignificant. The deposition of iron oxide particles transported from shielding balls to the heating surface is verified to increase CHF non-trivially. The results presented in this paper improve the state of the knowledge base permitting quantitative prediction of CHF in the CANDU end shield, refining our ability to assess the feasibility of IVR. The findings clarify the mechanisms governing CHF in this scenario, permitting identification of potential future research directions.

Effect of different conditions on heat transfer characteristics of molten Zr-Stainless steel pool with water injection

External Reactor Vessel Cooling (ERVC) could be accompanied by the in-vessel water injection to alleviate the thermal focusing effect of the light metallic layer and enhance the In-Vessel Retention (IVR) strategy. Through the wATer injectiOn on molten Zirconium-Stainless steel Metallic pool experiment (ATOM), the effects of different temperatures, mass flow rates of water injection, and different mass percentages of Zr on the heat transfer characteristics after water injection have been investigated. The mass flow rate of water injection ranges from 0.05 kg/s to 0.10 kg/s, and the mass percentage of Zr ranges from 13% to 60%. Eight tests were carried out, and the maximum upward heat flux removed from the melt surface ranged from around 800 to 1400 kW/m2. A higher water injection mass flow rate and a lower Zr mass percentage could lead to a higher steam generation rate. After water injection, the experimental melt temperature and concentration factor decrease. By applying the ATOM heat transfer coefficient as the input data in the nuclear reactor case, the calculated results indicate that the thermal focusing effect inducing by the decrease of the light metallic layer’s height could be highly mitigated by water injection. As a result, the integrity of the RPV could be enhanced. During the whole experiment for all the tests, no vapor explosion and melt ejection were found.

Cold sprayed superhydrophilic porous metallic coating for enhancing the critical heat flux of the pressurized water-cooled reactor vessel in nuclear power plants

In the nuclear power plants, external reactor vessel cooling (ERVC) by water is an effective strategy to ensure in-vessel retention (IVR) of the molten fuel in the pressurized water-cooled reactors (PWRs) to avoid the release of nuclear fission products under severe accidents. The effectiveness of IVR - ERVC, namely the heat taken away by the ERVC highly depends on the critical heat flux (CHF) of the outer wall of lower hemisphere the Reactor Vessel. Although porous coating has been shown significantly enhanced CHF experimentally and theoretically, flexible coating process is still in great demand for practical applications. In this work, cold spray (CS) is employed to deposit porous coatings by controlling the particle impact velocities of Ti-based alloy powder sprayed on a SA508 Gr3 steel substrate, which is one of the frequently used material for fabricating the PWRs. Effect of the spraying parameters on porosity, adhesion strength and wetting ability of the coating was examined. The boiling heat transfer CHF of the coating was correlated with microstructure features of the coatings. The microstructural observations show that the porosity of the coating can reach up to 30% by the controlling the impact velocity of the spraying particle. The porous microstructure makes the coating surface superhydrophilic and leads to a remarkable increment of up to 50.3% (499 kW/m2) in the CHF as compared to the bare SA508 Gr3 steel surface. Although the adhesion strength of the coating decreases with the porosity of the coating, cracking, delamination or peeling-off were not observed on the coating with highest porosity even after 4 cycles of CHF measurement and 100 cycles of water quenching thermal shock from 300 °C suggesting its high mechanical robustness.

CHF enhancement of downward-facing saturated pool boiling on the SCGS-modified surfaces with multi-scale conical pin fin structures

Enhancing the critical heat flux (CHF) of external reactor vessel cooling (ERVC) to achieve the strategy of in-vessel retention (IVR) has been deemed as an essential requirement for the security of modern commercial PWR nuclear power plants. A new type of multi-scale conical pin fin structures had been designed to meet this need. These conical structures of four different axial heights (1.0, 1.5, 2.0 & 2.5mm) and similar radial geometric parameters were fabricated on four heating surfaces of copper substrates by supersonic cold gas spray technique (SCGS), and the conical pin fin structures were combined with a thin layer of micro porous structures. Transient-state (quenching) and steady-state (heating) pool boiling experiments had been respectively conducted on 6 different downward-facing surfaces under different inclination angles including 5°, 30°, 45°, 60°, 75° & 90° (in vertical), which included above-mentioned four pin fin surfaces, a plain surface and a coated surface covered with a 1mm-thick layer of uniform porous coating fabricated by SCGS with copper particles. The experimental results showed that the CHF increased with the inclination angles. The CHF of all the SCGS-modified surfaces were significantly enhanced compared with those of the plain surface. The maximum CHF value and the maximum CHF enhancement belonged to the 2.0mm-height pin fin surface whose CHF values were more than 1.7 times of those of the plain surface, and up to 2.1 times in 90°. The 1.0mm pin fin surface had the greatest performance in boiling heat transfer coefficient (HTC) and its CHF-enhancing performance also greatly exceeded that of 1.0mm uniform porous coating surface. Remarkable performances of multi-scale conical pin fin structures had been experimentally confirmed, hence the safety margin of IVR-ERVC will be greatly enhanced after adopting them. In addition, the CHF enhancement mechanism of the multi-scale conical pin fin structures was analyzed.

Sensitivity Analysis of Multi-Layer Molten Pool Model of PWR Lower Head

The molten pool model of the lower head is an important model to evaluate the effectiveness of In Vessel Retention (IVR), which has been widely used in the safety evaluation of typical pressurized water reactor (PWR). The distribution and transfer process of the composition and heat of the melt in the pool of the traditional two-layer pool model and the three-layer pool model proposed in recent years are simulated, which is with the characteristics of complex relationship and strong nonlinearity. In order to provide a support for the optimization of the molten pool delamination model and the mitigation strategy of serious accidents, both global sensitivity analysis library (SALib) and IVR analysis code (CISER V2.0) developed by Nuclear Power Institute of China are used to analyze the sensitivity of four molten pool multilayer models and obtain the influence degree of the main input parameters on the key result parameters of each model. The sensitivity analysis results reflected a typical characteristics of the molten pool model. The radius of the lower head has a significant impact on the four molten pool multilayer models, while the input parameters that have significant influence on the key result parameters are basically the same in the Salay & Fichot model and the two-layer molten pool model, the initial mass of molten material has the greatest impact on the Esmaili model, and the density of the molten material has the greatest impact on the Seiler model.

Численный анализ естественной конвекции кориума в условиях внутрикорпусной локализации с учетом переменного тепловыделения

Численный анализ естественной конвекции кориума в условиях внутрикорпусной локализации с учетом переменного тепловыделения

Numerical study on thermal-hydraulics of external reactor vessel cooling in high-power reactor using MARS-KS1.5 code: CFD-aided estimation of natural circulation flow rate

This paper presents a numerical investigation of two-phase natural circulation flows established when external reactor vessel cooling is applied to a severe accident of the APR1400 reactor for the in-vessel retention of the core melt. The coolability limit due to external reactor vessel cooling is associated with the natural circulation flow rate around the lower head of the reactor vessel. For an elaborate prediction of the natural circulation flow rate using a thermal-hydraulic system code, MARS-KS1.5, a three-dimensional computational fluid dynamics (CFD) simulation is conducted to estimate the flow rate and pressure distribution of a liquid-state coolant at the brink of significant void generation. The CFD calculation results are used to determine the loss coefficient at major flow junctions, where substantial pressure losses are expected, in the nodalization scheme of the MARS-KS code such that the single-phase flow rate is the same as that predicted via CFD simulations. Subsequently, the MARS-KS analysis is performed for the two-phase natural circulation regime, and the transient behavior of the main thermal-hydraulic variables is investigated.