Monte Carlo (MC) burnup calculation method, implemented through coupling neutron transport and point depletion solvers, is widely used in design and analysis of nuclear reactor. Burnup calculation is generally solved by dividing reactor lifetime into steps and modeling geometry into numbers of burnup areas where neutron flux and one group effective cross sections are treated as constant during each burnup step. Such constant approximation for neutron flux and effective cross section will lead to obvious error unless using fairly short step. To yield accuracy and efficiency improvement, coupling schemes have been researched in series of MC codes. In this study, four coupling schemes, beginning of step approximation, predictor-corrector methods by correcting nuclide density and flux-cross section as well as high order predictor-corrector with sub-step method were researched and implemented in RMC. Verification and comparison were performed by adopting assembly problem from VERA international benchmark. Results illustrate that high order coupled with sub-step method is with notable accuracy compared to beginning of step approximation and traditional predictor-corrector, especially for calculation in which step length is fairly long.

ThO 2 has been considered as a possible replacement for UO 2 fuel for future generation of nuclear reactors, and thorium-based mixed oxide (Th-MOX) fuel performance in a light water reactor was investigated due to better neutronics properties and proliferation resistance compared to conventional UO 2 fuel. In this study, the thermal, mechanical properties of Th 0.923 U 0.077 O 2 and Th 0.923 Pu 0.077 O 2 fuel were reviewed with updated properties and compared with UO 2 fuel, and the corresponding fuel performance in a light water reactor under normal operation conditions were also analyzed and compared by using CAMPUS code. The Th 0.923 U 0.077 O 2 fuel were found to decrease the fuel centerline temperature, while Th 0.923 Pu 0.077 O 2 fuel was found to have a bit higher fuel centerline temperature than UO 2 fuel at the beginning of fuel burnup, and then much lower fuel centerline than UO 2 fuel at high fuel burnup. The Th 0.923 U 0.077 O 2 fuel was found to have lowest fuel centerline temperature, fission gas release and plenum pressure. While the Th 0.923 Pu 0.077 O 2 fuel was found to have earliest gap closure time with much less fission gas release and much lower plenum pressure compared to UO 2 fuel. So the fuel performance could be expected to be improved by applying Th 0.923 U 0.077 O 2 and Th 0.923 Pu 0.077 O 2 fuel.

Deep penetration problems exist widely in reactor applications, such as SPRR300 (Swimming Pool Research Reactor 300), a light water moderated, enriched uranium fueled research reactor in China. Deterministic transport theory is intrinsically suitable for deep penetration. But there exist some problems when it’s applied in SPRR-300research reactors. First, the reactor core is complicated for geometry description in deterministic theory codes. Monte Carlo method has advantages in complex geometry modeling. And it uses continuous energy cross sections which are independent with specific reactor types and research objections. But usually it’s difficult to converge well enough to deal with deep penetration problems, even though there are a number of variance reduction techniques. Based on the advantages and disadvantages of Monte Carlo and Deterministic method, we proposed a coupled neutron transport calculation method for deep penetration. It combines advantages of these two methods. Firstly, we use Monte Carlo code to finish fine modeling and do the whole reactor core calculation. Domestically developed Monte Carlo code JMCT is used to do the neutron transport calculation. Then homogenized group constants in each mesh are calculated from JMCT output by a self-developed script. Afterwards, we do the whole reactor calculation with deterministic theory code TORT. It directly uses group constants generated by Monte Carlo code. Finally, we can get the deep penetration calculation results from TORT output. Verification is carried out by comparing the group constants of benchmark problem, and by comparing k eff calculated by this method with continuous energy Monte Carlo method. Benchmark calculation is conducted with OECD/NEA slab benchmark problem. The comparison shows that group constants generated by this study are in good agreement with results from published references. Then above group constants are applied to 3-dimensional discrete ordinates deterministic theory transport code TORT. But k eff calculated by TORT is a little lower than that calculated by Monte Carlo code JMCT. To minimize other influence factors, different Sn/Pn order, and different mesh size in TORT has been tried. Unfortunately the k eff difference between these two methods remains. Even though the k eff results in this benchmark are less than k eff calculated by continuous energy MC method, Benchmark results show that all the group constants generated by this method are in good agreement with existing references. So it can be expected that after further verification and validation, this coupled method can be effectively applied to the deep penetration problem in such kind of research reactors.

In recent years there has been an increasing demand from nuclear research, industry, safety, and regulation bodies for best estimate predictions of Light Water Reactors (LWR’s) performances to be provided with their confidence bounds. From a neutronic point of view, among the different sources of uncertainty the main challenge is represented by the one related to the accuracy of the nuclear data libraries used in the transport calculations. The assessment of nuclear data uncertainties and their impact on the main reactor parameters plays a fundamental role not only for criticality safety but also in burnup analyses. In facts, the accurate prediction of nuclear parameters in burnup calculations strongly affects the management of spent nuclear fuel, the core design, as well as the economy and safety of nuclear reactors. In this paper a study related to the impact of the nuclear data uncertainties on the evolution in time of the criticality and the nuclide concentrations in burnup calculations is presented. The analysis has been performed by using a statistical sampling methodology in which all the uncertain parameters are handled as random dependent variables by a sampling procedure. The probability distributions of the uncertain input parameters are used to generate random variations of these input quantities starting from a covariance library in a 56-group energy structure. Calculations have been performed by means of the SCALE 6.2.2 code and ENDF/B-VII.1 nuclear data. The method has been tested on a PWR pin cell model representative of the TMI-1 15 × 15 assembly as defined in an international benchmark exercise. In the first part of the paper the methodology and the neutronics modelling of the problem are presented.

The main objective of this work is to produce an optimal modeling for our aged Planar-HPGe detector using Monte Carlo method (MC). That optimization included the analysis of the germanium dead (inactive) layer thickness for our old detection system (planar-HPGe detector). DL is one of the important parameters needed in order to obtain the smallest discrepancy between simulated and experimental measurements of detector efficiency. Also, precise determination of 235 U enrichment for UO 2 samples which is necessary for purposes of nuclear materials verification in the field of nuclear safeguards. The thickness of Germanium dead layer (DL) can be vary by time as it is not well known due to the existence of a transition zone where photons are strongly attenuated and absorbed, that cannot contribute to the total photon energy absorption which causes a significant decrease in efficiency. Therefore, using data provided by manufacturers since long years (manufacture date) in the detector simulation model is not convenient. As a result, some strong discrepancies appear between measured and simulated efficiency, in addition to that non-accurate results for 235 U enrichment determination. The Monte Carlo method applied to overcome this difficulty was to vary the thickness of dead layer step by step in simulation, a good agreement (minimum deviation) between estimated and experimental efficiency was reached when a suitable germanium dead layer thickness was chosen. Calculations and measurements were performed for radioactive nuclear material samples in the form of UO 2 powder with different sizes and enrichments at different locations, under different gamma-lines emitted after a-decay of the 235U nuclei. Results indicated that a good agreement between simulated and measured efficiencies is obtained using a value for the germanium dead layer thickness approximately (2.45 mm) six in comparison with (0.389 mm) provided by the detector manufacturer.

The efficient solution of the neutron diffusion equation for large scale whole core calculations is of paramount importance; especially if the detailed pin-level power distribution and reaction rates are required. For heterogeneous whole core calculations finite element based techniques have been one approach to modelling the detailed pin geometry in whole core calculations. A new approach, pioneered in the last few years, is isogeometric analysis (IGA) methods which enable the exact geometry of fuel pins to be modelled. In order to efficiently solve elliptic partial differential equations (PDEs), such as the neutron diffusion equation, typically multi-grid or multi-level iterative solution techniques are used such as the algebraic multi-grid method. However, using IGA methods it is possible to develop true geometric multi-grid techniques which are potentially much more efficient than the standard algebraic multi-grid methods. In this paper we explore the use of IGA methods to develop a scalable, multi-level, iterative algorithm which is then used to solve the neutron diffusion equation over several geometries. This multilevel solution algorithm utilises a single patch multi-grid framework suggested by Hofreither and Takacs that takes advantage of the tensor product construction to provide scalability with respect to spatial, and polynomial refinement. Furthermore, a two-level balancing Neumann-Neumann solver is used to extend the solver to multiple patches in a scalable way. It is seen that the number of iterations required depends on the mapping between the unit square and the physical geometry, as well as the material coefficients.

The UK Department of Business, Energy and Industrial Strategy (BEIS) recently launched an R&D programme in Digital Reactor Design, incorporating the development of a Nuclear Virtual Engineering Capability with an integrated Modelling and Simulation programme. A key challenge of nuclear reactor design and analysis is the system complexity, which arises from a wide range of multi-physics phenomena being important across multiple length scales. This project constitutes the first step towards developing an integrated nuclear digital environment (INDE) linking together models across physical domains and incorporating real world data across all stages of the nuclear lifecycle. Simulation case studies will be developed within the INDE framework, delivering an enhanced modelling capability while ensuring the framework has immediate application. For these case studies have been specified that are relevant to design and operation phases for AGR and PWR type reactors. The AGR case considers the through-life structural performance of graphite bricks. This involves modelling of multi-scale, multi-physics phenomena in the support of reactor operations. The PWR case study is based on core multiphysics modelling, with potential relevance to operating and future PWRs, and in particular in the design of SMRs.

Longevity of sensors and portable devices are severely limited by temperature, chemical instability, and electrolyte leakage issues associated with conventional electrochemical batteries. Batteries undergo self-discharge and permanent loss of capacity at high temperatures, exhibit lower voltage and capacity at low temperatures, and leak electrolyte shortening operating lives, corrosion of nearby electronics, and potential safety hazards in the form of burns and poisoning. Instabilities in lithium types often short resulting explosions, fire, and venting of hazardous gasses. Betavoltaics do not have these problems and can operate in a wide temperature range without permanent degradation and loss of capacity, will not explode, and be made safe. Betavoltaic technology is maturing with advances in radioisotope sources, semiconductor materials, and developments in low-power applications, and has been identified by Department of Defense as a disruptive technology that is needed and should be pursued. This study presents experimental and modeling research on the leading betavoltaic technology, recent developments and proposed advancements. The next generation 100 microwatt betavoltaic is introduced along with projected voltage, current, and power characteristics.

In order to improve nuclear reactor’s performance and safety, a new-type fuel, Inert Matrix Dispersion Pellet (IMDP) with greatly enhanced thermal conductivity was studied. In this paper, the pellet was developed by Spark Plasma Sintering Technique (SPS), which brings an improved thermal conductivity compared with normal method and exhibits fully dense SiC matrix with a higher TRISO volume in kernel of pellet. Thermal-physical properties of the IMDP from room temperature to 1400 °C were investigated. The thermal conductivity improved 1100% at room temperature and 974% under 1000°C when TRISO volume is 40%, which behave much better than traditional UO 2 fuel. The thermal diffusivity reduced and heat capacity increased at different TRISO volumes, linearly implied stable geometry at high temperature. From CT test we can see an intact structure of uniform TRISO granules. For comparison, modeling thermal conductivity is analyzed by FEM, which shows lower than measured data, indicating an optimized technology of this new method. Advantages of SPS will be discussed as well.

Uncertainty quantification is an indispensable analysis for nuclear reactor simulation as it provides a rigorous approach by which the credibility of the predictions can be assessed. Focusing on propagation of multi-group cross-sections, the major challenge lies in the enormous size of the uncertainty space. Earlier work has explored the use of the physics-guided coverage mapping (PCM) methodology to assess the quality of the assumptions typically employed to reduce the size of the uncertainty space. A reduced order modeling (ROM) approach has been further developed to identify the active degrees of freedom (DOFs) of the uncertainty space, comprising all the cross-section few-group parameters required in core-wide simulation. In the current work, a sensitivity study, based on the PCM and ROM results, is applied to identify a suitable compressed representation of the uncertainty space to render feasible the quantification and prioritization of the various sources of uncertainties. While the proposed developments are general to any reactor physics computational sequence, the proposed approach is customized to the TRITON-NESTLE computational sequence, simulating the BWR lattice model and the core model, which will serve as a demonstrative tool for the implementation of the algorithms.

In nuclear plant piping, when high-temperature water penetrates from main pipe, a thermal stratification may be formed in a branch pipe with a closed end. The penetrating flow of high-temperature water is called as a cavity flow. If a thermal stratification interface is formed in an elbow of vertical-horizontal branch pipes, thermal fatigue may occur due to periodic temperature fluctuations. Because high cycle thermal fatigue may have significant influence on structure integrity of pipe, it is important to evaluate the position which is called penetration depth of a cavity flow in which a thermal stratification interface is formed. In the present evaluation guideline which was formulated by the Japan Society of Mechanical Engineers, applicable range is limited only to 50 mm diameter of vertical-horizontal branch pipe. Therefore, it needs to expand the applicable diameter range of vertical-horizontal branch pipe. In this research, 200 mm diameter piping test is conducted to confirm characteristic phenomena, and to build evaluation method of penetration depth in large diameter piping. This paper presents experimental results and consideration of thermal hydraulics phenomena about thermal stratification. The horizontal piping test and the vertical-horizontal branch piping test are conducted. Pipes are made of acrylic resin for visualization of water flow. The water temperature is less than 60°C. The horizontal piping test reproduces natural circulation flow by installing a heater which imitated heating from cavity flow, and a chiller which imitated heat dissipation from a blockage valve. Temperature profiles are obtained by using thermocouples and optical fibers. Moreover, water flow in piping is visualized by injected ink. From the visualization and the temperature measuring result in the horizontal piping test, flow characteristic of the natural circulation region of large diameter piping is clarified The vertical-horizontal branch piping test reproduces cavity flow penetration from main pipe. Temperature profiles are obtained at cross-section direction and axial direction. Moreover, the amount of mass transfer at the thermal stratification which is formed between a cavity flow and natural circulation is evaluated experimentally by using a lithium tracer method. The second paper[Ref.2] explains modeling of evaluation method for cavity flow penetration depth in large diameter piping.

This study recaps the experimental effort to characterize a transverse plane flow field in a tightly-packed, 61-pin, wire-wrapped, hexagonal fuel bundle prototypical for a sodium fast reactor. The motive was to produce high spatiotemporal experimental data for computational fluid dynamics turbulence model validation. The matched-index-of-refraction and laser-based optical measurement techniques were utilized on an isothermal experimental flow facility. Measurements were performed on a transverse plane perpendicular to the axial flow. Fluid flow in the three types of subchannels (corner, exterior, and interior) were quantified. All measurements have been performed at a bundle-averaged Reynolds number of approximately 10,400. Results include flow statistics such as ensemble-averaged velocity, root-mean-square fluctuating velocity, and Reynolds stress. Of interest was the flow behavior around the restriction caused by the wire spacer and hexagonal duct wall, where recirculation regions formed. Regions of maximum and minimum momentum transfer coincided with regions of maximum and minimum fluctuations. These regions highlight locations of maximum and minimum cooling of the fuel pins. The experimental data will be used to benchmark computational fluid dynamics simulations of the sodium fast reactor fuel bundle using Reynolds-averaged Navier Stokes and large-eddy simulation turbulence modeling methods.

The direct and indirect hydro-dynamic loads in the suppression pool of the BWR fleet were developed well before 2000. These loads were based on scaled experiments and numerical solutions using one dimensional models. The analysis was cumbersome and conservatisms were added in multiple steps. As these loads spread throughout the containment, as global vibrations, they are generally a part of all structural verification inside the containment. This fact has made it hard and unpractical to challenge and revise these loads, as a change could lead to significant re-work. A consequence of this is that loads in the suppression pools have seldom been revisited, regardless if they cause local or global vibrations. This is problematic when new equipment is needed. The design of this equipment suffers from the very conservative loads. Experience has shown that these loads can be challenged and refined using updated techniques, leading to significantly lower loads. This was realized during the modernization of Swedish plant Oskarshamn Unit 2 with Mark II containment, where loads following pool swell proved to be particularly challenging and it was decided to investigate possibilities to reduce the conservative loads. Due to the large scale of the condensation pool coupled with the transient and small scaled condensation of steam at the drywell vent pipe nozzle full CFD resolution is not feasible. Instead lumped models in GOTHIC was used to increase resolution from the 1D approach of normal containment analysis, to a resolution that can account for the features of the condensation pool. This showed that the pool swell was less uniform than initially thought, leading to fewer objects affected as well as lower loads on objects that suffered from these loads. A full CFD analysis was then used to resolve phenomena working on even shorter time scales leading to a complete rework of all local loads. The loads addressed using updated codes and modelling techniques was pool swell impact (PSI), pool swell drag (PSD) and local drag loads due to pressure relief valve opening, LV/SRV. The current work shows that using updated modelling techniques and aligning results with previous analysis and documentation, it is possible to reduce loads for some events in the suppression pool without violating safety for the power plant. The results from the GOTHIC model shows that the cylindrical shape of the pool will create an uneven velocity distribution radially at pool swell resulting in much smaller loads at the outer boundary. CFD analysis of the LV/SRV event shows that the loads are reduced in comparison with previous methodology and this is mainly because of shadow effects. The calculations also verifies assumptions used in the previous methodology.

In nuclear plant piping, when the high-temperature water penetrates from main pipe, a thermal stratification interface may be formed in branch pipe with a closed end. The penetrating flow of high-temperature water is called a cavity flow. If a thermal stratification interface is formed in an elbow of vertical-horizontal branch pipes, thermal fatigue may occur by a periodic temperature fluctuation. Because high cycle thermal fatigue may have significant influence on structure integrity of pipe, it is important to evaluate the position which is called penetration depth of a cavity flow in which a thermal stratification interface is formed. In present evaluation guideline which was formulated by the Japan Society of Mechanical Engineers, applicable piping is limited only to 50 mm diameter of vertical-horizontal branch pipe. Therefore, it needs to expand the applicable diameter range of vertical-horizontal branch pipe. In this research, 200 mm diameter piping test is conducted to confirm characteristic phenomena, and to build evaluation method of penetration depth in large diameter piping. This paper presents about modeling of evaluation method for cavity flow penetration depth in large diameter piping. First, the mass transport model at the interface of layer in natural circulation region is arranged with Reynolds number and Richardson number based on visualization and the temperature measuring result. Second, the mass transport model at a thermal stratification interface between a cavity flow domain and a natural circulation domain is built based on time variation of lithium tracer concentration. The evaluation model of cavity flow penetration depth applicable to large diameter piping is proposed by combining new mass transportation models and existing evaluation guideline. In conclusion, predictive accuracy of cavity flow penetration depth in 200 mm diameter piping became the same order as the evaluation result of 50 mm diameter piping by using present evaluation guideline.

When a severe accident of a nuclear reactor occurs, decommissioning work becomes important task. In the decommissioning work of a boiling water reactor after the severe accident, estimation of the sedimentation place of the molten debris is important. However, the technique to estimate exactly the sedimentation place has not been enough developed. Therefore, the detailed and phenomenological numerical simulation code named “JUPITER” is under development for predicting the molten core behavior in JAEA (Japan Atomic Energy Agency). The comparison between experimental and numerical results is necessary to clarify the validity of the numerical analysis code. The study provides the experimental data to examine the numerical simulation code. As a basic study to examine the numerical simulation code, a liquid film flowing in a modeling flow channel was studied by using water. The flow was visualized, and the flow data were obtained by image processing.

MELCOR is a fully-integrated, system-level computer code developed by Sandia National Laboratories (SNL) for the Nuclear Regulatory Commission (NRC) with the primary objective of modeling the progression of severe accidents in light water nuclear power plants [1,2,3]. Since the project began in 1982, MELCOR has undergone continuous development to address emerging issues, process new experimental information, and create a repository of knowledge on severe accident phenomena. This paper summarizes model development specifically developed for non-LWR applications such as high temperature gas reactors (HTGR), sodium fast reactors (SFR) and molten salt reactors (MSR). Beginning in 2008, active development work began on HTGR modeling in MELCOR. Models were developed for helium gas thermodynamics, oxidation of graphite, thermal hydraulics and heat transfer for both prismatic and pebble bed designs, cavity cooling systems, fuel failure and fission product release, graphite dust generation, and aerosol transport, deposition, and resuspension. In 2013, work commenced on the development of modeling capabilities for sodium fast reactors. This development included the addition of sodium as a working fluid as well as the addition of models for simulating containment fires (both spray and pool) as well as sodium atmospheric chemistry. Validation of these new models has been completed and code-to-code comparisons with the CONTAIN/LMR code has been performed. Work continues as development of sodium concrete interaction models is now underway. In 2017, work began on adding capabilities for molten salt reactors. A new equation of state for FLIBE coolant has been successfully tested in MELCOR and is now undergoing validation against experiments. The alternate working fluid model has also been extended to permit both water and one alternate condensable working fluid in the same input model.

In this paper, a thermal-hydraulics analysis code under non-inertial system was developed by establishing dynamic simulation models of typical ocean conditions and modifying RELAP5 code. Based on the modified code, the effects of the inclination, fluctuation and rolling conditions on the natural circulation characteristics of an integrated PWR were studied. The dynamic input and output interface model of RELAP5 was further established, and RELAP5 was coupled with a two-group three-dimensional neutron kinetics code to realize the improvement of nuclear feedback module in thermal hydraulic code. And the distribution of core flow and power under different motions was analyzed. The results indicate that, the uneven distribution of coolant flow increases with the increasing inclination angle, this trend leads to an uneven distribution of primary coolant flow. The flow fluctuation has a 180° phase difference under rolling conditions, and the reactor power and coolant flow oscillation increases with the increasing rolling period and amplitude. In the case of heaving motion, the peaks of the oscillation amplitude of the flow and power lying in the hottest channel as the additional forces on the fluid of each channel are spatially uniform. The code developed in this paper has the functions of modelling ocean conditions and three-dimensional coupled neutronics/thermal-hydraulics, and can be used as a simulation tool for Floating PWR.

The hydrodynamic loads are very important for the discharge piping design because of the fast opening of the valves and the presence of liquid in the upstream loop seals. Therefore, this work implemented a force balance formulation into the RELAP/SCDAPSIM code to determine the hydrodynamic loads directly. The force balance method, which equates resultant force transmitted from fluid to structural elements as the sum of all pressure and frictional tractions acting on the wetted surface of the element. Then an assessment was performed with RELAP/SCDAPSIM code for EPRI/CE Test No.917, which was a simulation of a hot water loop seal discharge followed by steam discharge through a pressurizer safety valve. In the RELAP/SCDAPSIM calculations, the input data such as valve opening characteristics, initial fluid conditions, system Nodalization, and system component modeling, were based on original RELAP5/MOD1 and RELAP5/MOD3 calculations performed by EPRI. The calculation results were compared with the experimental data as well as the RELAP5/MOD1 and RELAP5/MOD3 calculation to assess the applicability of the RELAP/SCDAPSIM for the hydrodynamic loads calculation. It was demonstrated from the comparisons that the RELAP/SCDAPSIM code with force balance calculation yields very similar hydrodynamic loads to the experimental data as well as the RELAP5/MOD1 and RELAP5/MOD3 calculation results. Therefore, it is recommended that the new force balance calculation implemented in RELAP/SCDAPSIM can be used to calculate the hydrodynamic loads for the hot water loop seal discharge followed by steam discharge transient.

Molten corium-concrete interaction (MCCI) is an important ex-vessel phenomenon that could happen during the late phase of a hypothetical severe accident in a light water reactor. When the molten corium, which is generally comprised of UO 2 , ZrO 2 and metals such as zircalloy and stainless steel, is discharged into a dry reactor cavity, a stratified molten pool configuration with two immiscible oxidic and metallic phases can be expected to form and lead to MCCI. Compared to a homogenous oxidic molten pool configuration, the metallic phase in the stratified molten pool might influence the crust formation on the corium-concrete interface and consequently cause different concrete ablation behavior to evaluate MCCI progression concerning containment failure. In terms of this issue, past experimental studies, such as COMET-L, VULCANO VBS and MOCKA test series, have been carried out to investigate the influence of such oxidic and metallic stratified pool configuration on MCCI. The experimental results have shown that the metallic phase can have a significant impact on the axial and radial ablation kinetics that could influence the ablation patterns of reactor pit. As regards numerical studies, past numerical modeling of MCCI was generally based on Eulerian methods and simplified empirical approach to simulate solid/liquid phase change and evolving of corium/crust/concrete interface. Such modeling might be efficient but have shown deficiencies and inadequacies due to its Eulerian and empirical nature, which has suggested a necessity to seek for a more mechanistic approach for modeling of MCCI. In this sense, Moving Particle Semi-implicit (MPS) method is considered suitable for MCCI analysis for its advantages of tracking interfaces and modeling phase change accurately as a Lagrangian particle method. In the present study, a three-dimensional (3-D) numerical study has been performed to simulate COMET-L3 test carried out by KIT with a stratified molten pool configuration of simulant materials with improved MPS method. Solid/liquid phase change was simulated with types of solid and liquid particles with thermal and physical properties including temperature and solid fraction, which enabled tracking of the solid/liquid status of each particle to achieve accurate free surface and corium/crust/concrete interface capturing. The heat transfer between corium/crust/concrete was modeled with heat conduction between particles. Moreover, the potential influence of the siliceous aggregates was also investigated by setting up two different case studies since there was previous study indicating that siliceous aggregates in siliceous concrete might contribute to different axial and radial concrete ablation rates. The simulation results have indicated that metal melt as corium in MCCI can have completely different characteristics regarding concrete ablation pattern from that of oxidic corium, which needs to be taken into consideration when assessing the containment melt-through time in severe accident management.

In this work, a so-called mini-core consisting of a 3 × 3 array of 17 × 17 pressurized water reactor (PWR) fuel assemblies (FA) is considered with the aim of identifying the most conservative window size for hot channel analysis of bowed fuel assemblies. Overall, five different mini-core configurations are analyzed: one is the reference case, i.e. without FA displacement and four different cases with diagonal and parallel FA displacements. Rod power maps for these mini-cores were exported from neutronic calculations with CASMO-SIMULATE codes. Subchannel modelling with COBRA-TF code of all five mini-cores allows one to identify the rod position with a minimum departure from the nucleate boiling ratio (DNBR) and to construct input decks with different rod window sizes around the previously identified rod position. Overall, eight different window sizes are considered: 3 × 3, 5 × 5, 7 × 7, 9 × 9, 11 × 11, 13 × 13, 15 × 15 and 17 × 17. Results of subchannel analysis for a mini-core and different subchannel window configurations are compared with the help of DNBR parameter, which is the ratio between the critical heat flux (CHF) and the actual local heat flux on a rod. An assessment of three different CHF models is applied in this work: Groeneveld CHF look-up table (LUT), W3 CHF correlation, and Doroschuk CHF LUT. The general conclusion of this work is that for deformed core configurations, an appropriate rod window size needs to be determined to adequately capture the local flow redistribution. For large displacements (the largest displacement considered in this work is 10 mm), the DNBR ratio can drop to one. DNBRs obtained with the W3 CHF correlation give the most conservative results.