Coupled Neutron Research Articles

Pareto front of sodium void worth and breeding ratio in metal-fueled sodium fast reactor with axially heterogeneous core design by coupling neutronics and genetic algorithm

For sodium-cooled fast reactors (SFRs), achieving low sodium void worth (SVW) and flexible breeding ratios (BR) is crucial. This study establishes a method coupling neutronics simulations with genetic algorithms (GA) and applies it to a 3000 MWth metal-fueled SFR to optimize the SVW, BR, and the power distribution of axially heterogeneous cores. The Pareto front, i.e. optimal front, between SVW and BR is determined, quantitatively clarifying the trade-off relationship between these two parameters. Axially heterogeneous cores with equal inner and outer fissile heights can achieve SVW adjustments from −71 to 3148 pcm and BR adjustments from 1.26 to 1.62. Designs with unequal fissile heights slightly expand the feasible region, but the impact on the optimal front for equal-height designs is limited. Through the analysis of the spatial distribution of the sodium void effect, it is shown that the bond sodium model significantly influences the SVW in low sodium void cores, while its impact on high breeding cores is minimal. The results indicate that the power distribution can be optimized by adjusting the enrichment of subregions within the constraints of SVW and BR. For low sodium void cores, retaining a certain amount of lower fertile material is necessary to flatten the power distribution. While the criticality search during depletion calculations affects the parameters, its impact on identifying the optimal front is limited.

From ionic clusters dynamics to network constraints in ionic polymer solutions.

Physical networks formed by ionizable polymers with ionic clusters as crosslinks are controlled by coupled dynamics that transcend from ionic clusters through chain motion to macroscopic response. Here, the coupled dynamics, across length scales, from the ionic clusters to the networks in toluene swollen polystyrene sulfonate networks, were directly correlated, as the electrostatic environment of the physical crosslinks was altered. The multiscale insight is attained by coupling neutron spin echo measurements with molecular dynamics simulations, carried out to times typical of relaxation of polymers in solutions. The experimental dynamic structure factor is in outstanding agreement with the one calculated from computer simulations, as the networks are perturbed by elevating the temperature and changing the electrostatic environment. In toluene, the long-lived clusters remain stable over hundreds of ns across a broad temperature range, while the polymer network remains dynamic. Though the size of the clusters changes as the dielectric constant of the solvent is modified through the addition of ethanol, they remain stable but morph, enhancing the polymer chain dynamics.

Coupled neutronics and thermal-hydraulics using TRUST-NK for high performance computing in molten salt reactor simulation

Molten Salt Reactors display a strong coupling between neutronics and thermal-hydraulics for which various simulation tools have been developed. In this paper we describe TRUST-NK, a neutronics code based on TRUST, CEA’s High Performance Computing (HPC) platform. TRUST-NK provides a multi-group diffusion solver with transport of delayed neutron precursors, for both steady-state and transient simulations. Being based on TRUST allows TRUST-NK to take advantage of the platform’s HPC capabilities and to easily couple to CFD solvers provided by other derivatives of TRUST. After a brief description of the code, we present an example of application to a simplified MSR model.

Coupled neutronics, thermal-hydraulics, and fuel performance analysis of dispersion plate-type fuel assembly in a cohesive way

The nuclear reactor is a multi-physics and multiscale coupling system. The conditions of high temperature and high neutron flux in the reactor pose a significant challenge to the thermodynamic performance of the fuel. To evaluate the plate-type fuel's performance, a coupling calculation of neutron physics code OpenMC, fuel performance analysis code BEEs-Plates, and nuclear reactor system analysis code NUSAC has been implemented based on the open-source MOOSE platform. A novel and efficient method for mesh grids generation and data mapping between external codes and MOOSE was proposed to ensure accurate data exchange. This method can simplify the mesh grids generation process of external codes and accurately transfer data across different mesh grids. Finally, the coupling calculation of the JRR-3 M research reactor fuel assembly was carried out based on the above coupling code, and the neutronics parameters, thermal–hydraulic parameters, and fuel performance-related parameters of the fuel assembly were analyzed. The calculation results show that the maximum fuel burnup will reach 16.91% FIMA after 240 days. With the increase of fuel burnup, the peak power of the fuel assembly is significantly shifted towards both ends of the height direction. The VonMises stress, volume stress, creep strain, and displacement of the fuel will be considerable. Finally, the impact of the coupling calculation on the accuracy of the computational efficiency is preliminarily investigated.

Coupled neutronics–thermal-hydraulics analyses of unprotected transients in LBE-cooled Compact High-Temperature Reactor

Coupled neutronics–thermal-hydraulics analyses of unprotected transients in LBE-cooled Compact High-Temperature Reactor

Solid-state battery measurement

Coupling neutron reflectometry with electrochemistry could accelerate the understanding of the interphase between lithium metal and solid electrolytes in solid-state batteries.

Nuclear analysis method based on current coupling neutron response reconstruction method

ABSTRACT The current coupling neutron response reconstruction method has been developed, based on the already developed direct response matrix method. In the current coupling neutron response reconstruction method, by setting fuel rod neutron production rates as new unknown values, the need to calculate the response matrix is eliminated, and the calculation time can be reduced compared to the direct response matrix method. Furthermore, by evaluating the neutron fluxes and the cross sections using the reconstruction method of homogenized cross sections which has already been developed, the neutron effective multiplication factor is directly evaluated from the reaction rate ratio, and convergence can be improved. For the purpose of verification, calculations for heterogeneous systems were performed. The results obtained using the current coupling neutron response reconstruction method agreed well with those obtained using direct calculations by the Monte Carlo method, and they were almost the same as those obtained using the direct response matrix method. These good agreements with the direct calculations using the Monte Carlo method meant that the current coupling neutron response reconstruction method accurately reflected the effects of intra- and inter-assembly heterogeneities. In terms of calculation time, the current coupling neutron response reconstruction method was 33 to 60 times faster than the direct response matrix method by eliminating the need for a response matrix and improving convergence.

The NILO-CMFD Method for Iteratively Solving Coupled Neutron Transport–Thermal Hydraulics Problems

The coarse-mesh finite difference (CMFD) method is commonly used to accelerate the iterative convergence of single-physics neutron transport problems. For multiphysics problems, the neutron cross sections depend on the temperature and density, both of which depend on the fission heat source; the resulting nonlinear feedback can significantly degrade the performance of CMFD and even cause instability. In this paper, we propose, for a class of one-dimensional (1-D) model multiphysics problems, a new nonlinearly implicit low-order (NILO) CMFD (NILO-CMFD) acceleration method to improve the performance of CMFD-based methods for solving loosely coupled multiphysics problems. Our numerical testing and Fourier analysis show that for the 1-D model problems, the new NILO-CMFD method achieves the same rapid convergence rate that CMFD achieves for single-physics problems.

Optimal Radial Build and Transmutation Properties of a Fusion-Based Transmutation Reactor with Molten Salt Coolants

The optimal shape of a fusion-based transmutation reactor with a molten salt coolant was determined by plasma physics, technology, and neutronic requirements. System parameters such as neutron multiplication, power density, shielding, and tritium breeding, were calculated in a self-consistent manner by coupling neutron transport analysis with conventional tokamak systems analysis. The plasma physics and engineering levels were similar to those used in the International Thermonuclear Experimental Reactor. The influence of aspect ratio of the tokamak and fusion power on the radial build, and the transmutation properties associated with two molten salt options, FLiBe and FliNaBe, were investigated. Being compared with a transmutation reactor with a small aspect ratio, a transmutation reactor with large aspect ratio was smaller in size and had a larger maximum fusion power. This type of reactor also revealed increased tritium-breeding capability and a smaller initial transuranic (TRU) inventory with a slightly lower burn-up rate. The burn-up rate for molten salt using either FLiBe or FLiNaBe was similar, but the initial TRU inventory and the tritium-breeding capability were smaller with FLiNaBe compared with FLiBe.

Coupled neutronics and thermal-hydraulics transient simulation of a gas-cooled reactor in the aircraft nuclear propulsion system

Aircraft nuclear propulsion (ANP) has many historical explorations, but also features with high power density and high operating temperature. These challenges lead to numerical investigations for the transient behaviors of gas-cooled reactors to be used for ANP system. This prismatic-block type reactor is featured by a large number of internal discrete cooling holes with high aspect ratio. When using conjugate heat transfer (CHT) model in CFD analyses for the entire solid core of such type of reactors, the refined mesh resolution near the interface between the fuel matrix and the flow channels leads to high computational costs. Therefore, we proposed a new porous heat transfer (PHT) model with Newton’s law of cooling to predict three-dimensional distributions of temperature in the solid and fluid domain of the reactor core. The verification of the PHT model is conducted in the presented study firstly, and shows that the accuracy of this model is deeply influenced by the prediction of convective heat transfer coefficients. Sequentially, a neutronics (N) and thermal-hydraulics (TH) transient coupling code is developed in the open-source platform OpenFOAM. The point kinetic model is employed to evaluate the dependency of heat power on the time. Doppler broadening effect is considered in the coupling process by delivering fuel average temperatures. Then coupled N/TH simulations are carried out to analyze the gas-cooled reactor during the start-up process and the blockage accident. The obtained results show that the wall temperature is always lower than the design limit even in a rapid start-up process, indicating that the reactor is safe at extreme operations. Moreover, simulation of the blockage accident shows that if several flow channels are totally occupied by debris, the fuel temperature increment is less than 300 K, and the position of the peak fuel temperature transforms from the core center to the zone near the blockage.

The NILO-CMFD Method for Iteratively Solving Coupled Neutron Transport-Thermal Hydraulics Problems

The NILO-CMFD Method for Iteratively Solving Coupled Neutron Transport-Thermal Hydraulics Problems

A Code-Agnostic Driver Application for Coupled Neutronics and Thermal-Hydraulic Simulations

While the literature has numerous examples of Monte Carlo (MC) and computational fluid dynamics (CFD) coupling, most are hardwired codes intended primarily for research rather than as stand-alone, general-purpose applications. In this work, we describe an open source application, the Exascale Nuclear Reactor Investigative COde (ENRICO), which enables coupled neutronic and thermal-hydraulic simulations between multiple codes that can be chosen at run time (as opposed to a coupling between two specific codes). The application has been designed such that the control flow logic, domain mapping, nonlinear fixed-point iteration, solution transfers, and convergence checks are all agnostic to the underlying physics solvers used. Special emphasis has also been placed on enabling efficient execution on distributed-memory computing environments. The transfer of solution fields between solvers is performed in memory rather than through filesystem input/output. Additionally, solvers can be configured to run on overlapping or disjoint sets of processes. To date, coupling with the OpenMC and Shift MC codes, the Nek5000 CFD code, and a simplified heat diffusion and subchannel solver has been implemented in ENRICO. We present results for coupled simulations of a single light water reactor fuel assembly based on the NuScale reactor using various combinations of the physics solvers. For this problem, the coupled simulations are shown to converge in about four Picard iterations. A comparison of the heat source and temperature distributions computed by ENRICO using OpenMC coupled with Nek5000 and Shift coupled with Nek5000 illustrates remarkable agreement between the codes.

Detailed Simulation of the Nominal Flow and Temperature Conditions in a Pre-Konvoi PWR Using Coupled CFD and Neutron Kinetics

The aim of the numerical study was the detection of possible vortices in the upper part of the core of a Pre-Konvoi Pressurized Water Reactor (PWR) which could lead to temperature cycling. In addition, the practical application of this Computational Fluid Dynamic (CFD) simulation exists in the full 3D analysis of the coolant flow behavior in the reactor pressure vessel of a nuclear PWR. It also helps to improve the design of future reactor types. Therefore, a CFD simulation of the flow conditions was carried out based on a complex 3D model. The geometry of the model includes the entire Reactor Pressure Vessel (RPV) plus all relevant internals. The core is modelled using the porous body approach, the different pressure losses along and transverse to the main flow direction were considered. The spacer-grid levels were taken into account to the extent that in these areas no cross-flow is possible. The calculation was carried out for nominal operating conditions, i.e., for full load operation. Furthermore, a prototypical End of Cycle (EOC) power distribution was assumed. For this, a power distribution was applied as obtained from a stationary full-core calculation with the 3D neutron kinetics code DYN3D. In order to be able to adequately reproduce flow vortexes, the calculation was performed transiently with suitable Detached Eddy Simulations (DES) turbulence models. The calculation showed fluctuating transverse flow in the upper part of the core, starting at the 8th spacer grid but also revealed that no large dominant vortices exists in this region. It seems that the core acts as a rectifier attenuating large-scale vortices. The analyses included several spacer grid levels in the core and showed that in some areas of the core cross-section an upward increasingly directed transversal flow to the outlet nozzle occurs. In other areas of the core cross-section, on the other hand, there is nearly any cross-flow. However, the following limitations of the model apply: In the model all fuel elements are treated identical and cross flows due to different axial pressure losses for different FA types cannot be displayed. The complex structure of the FAs (eg. flow vanes in spacer grids) could also influence the formation of large-scale vortices. Also, the possible influence of two-phase flows was not considered.

Stability analysis of predictor-corrector schemes for coupling neutronics and depletion

Numerical instability resulting from coupling isotopic depletion and neutronics is widely reported, but little work has been done to interrogate the phenomenon theoretically. This work extends the pioneering analysis developed by Densmore et al., who examined the stability of a monoenergetic neutron diffusion problem coupled with a simple depletion system through the explicit Euler scheme. Their Fourier stability analysis is applied to the predictor-corrector scheme and its iterated and implicit variants. The results reproduce behaviour described in previous reports of numerical instability and its solution, and develop a future avenue for diagnosing the stability of burn-up problems.

Development and testing of a coupled SuperMC and SUBCHANFLOW code for LWR simulations

Coupled neutronics and thermal-hydraulics simulations have evolved in the numerical modelling of light water reactors (LWRs) recently. To meet the request of large-scale coupling analyses, this study conducts the steady-state coupling simulations by using the Monte Carlo neutronics code, SuperMC, and the sub-channel code, SUBCHANFLOW. An external coupling mode is adopted in the coupling code, which provides a flexible selection of thermal-hydraulic codes. An on-the-fly Doppler broadening method is used in the coupling code to decrease its complexity. Based on the innovative CAD functions of SuperMC, two different methods are developed to create the data mapping between SuperMC and SUBCHANFLOW. To accelerate the convergence speed, the stochastic approximation method is applied to the coupling iterations. The coupling code is tested with a PWR single pin case, a BWR 3 × 3 cluster case, and a PWR full-assembly case. The results show a good agreement with the reference results calculated by other similar coupling codes. In conclusion, this study demonstrates the ability of SuperMC to perform large-scale steady-state coupled neutronics and thermal-hydraulics simulations of LWRs.

Simulation of a Transient Process in VVER-1200 by Means of the Athlet/BIPR-VVER Coupled Neutronics and Thermohydraulic Code

The results of a simulation, performed using the code ATHLET/BIPR-VVER (Russia, Germany), of tests on connecting a single main circulation pump (MCP) unit with three operating units during operation of VVER-1200 at 26% normal power level are presented. A comparison of the calculation and experiment showed the accuracy of the computational model to be satisfactory. Modeling is sufficient from the stand-point of calculating the macroparameters of a reactor facility.

Coupled Neutronics–Thermal-Hydraulic Simulation of BEAVRS Cycle 1 Depletion by the MCS/CTF Code System

The coupled neutronics–thermal-hydraulic simulation of the Benchmark for Evaluation and Validation of Reactor Simulations (BEAVRS) Cycle 1 depletion has been performed by the Monte Carlo–based multiphysics coupling code system MCS/CTF. MCS/CTF is a cyclewise pi-card iteration-based inner-coupling code system that couples the subchannel thermal-hydraulic code CTF as a thermal-hydraulic solver in the Monte Carlo neutron transport code MCS. MCS has been developed by the Computational Reactor Physics and Experiment Lab group at the Ulsan National Institute of Science and Technology for the full-core analysis of large-scale commercial light water reactors with high fidelity at the engineering level. With the high-fidelity performance of MCS, the quarter-core pinwise depletion simulation for the BEAVRS Cycle 1 benchmark has been conducted with thermal-hydraulic feedback including fuel temperature, coolant temperature, and coolant density. Moreover, the MCS internal one-dimensional thermal-hydraulic solver TH1D (MCS/TH1D) has been utilized as the reference. On one hand, the simulated results of the criticality boron concentration and axially integrated assemblywise detector signals were compared with measured data. On the other hand, the comparisons of power, fuel temperature, coolant temperature, and density are also presented in this paper.

Elemental analysis of the geothermic microecology of the Los Azufres (Mexico) volcanic complex

Three geothermal microecologies, constituted of fumaroles, hydrothermal springs and the native volcanic superstratum at the Los Azufres (Mexico) volcanic complex, have been analysed for the concentrations of Cd, Hg, Pb, Th and U in the tissues and rhizospheric soils of the extremophilic plants that colonise the fumaroles, as well as in the sediments and volcanic stratum, using Polarised Energy Dispersive X-ray Fluorescence spectrometry and supplementarily, by Inductively Coupled Plasma Mass Spectrometry and Neutron Activation Analysis. The plants are found to sequester high concentrations of these heavy elements in their tissues and are more effective at their immobilization in the rhizosphere than are the hydrothermal chemistries for the sediment. Inferences are drawn regarding their roles in, in the biogeochemical recycling of these heavy elements in the Paleozoic era whose ecology these geothermal niches ‘mimic’, as well as their potential for phytoremediation.

Coupled Neutron and Gamma Heating Calculation Based on VARIANT Transport Solutions

For accurate assessment of nuclear heating in fast reactors, a new coupled neutron and gamma heating calculation scheme has been developed based on VARIANT nodal transport solutions of neutron and gamma flux distributions. The MC2-3 code was extended to generate multigroup neutron and gamma cross sections and kinetic energy release in materials (KERMA) factors, and a utility program CURVE was developed to reconstruct detailed pin and duct wall powers from VARIANT output files. The improved heating calculation scheme has been verified against MCNP6 Monte Carlo reference solutions for the Advanced Burner Test Reactor (ABTR) and Experimental Breeder Reactor II (EBR-II) benchmark problems. Compared to the existing coupled heating calculation method based on DIF3D diffusion theory solutions, the new heating calculation scheme utilizes more accurate gamma cross sections and KERMA factors, accounts for the transport effects, and eliminates the approximations in the existing pin power reconstruction scheme. As a result, it produces more accurate assembly and pin power distributions. For both the ABTR and EBR-II problems, the maximum assembly power error was ~1% in fuel assemblies and ~2% in instrumented structure assemblies, and the maximum error in pin segment powers in an axial node of fuel assembly was ~4%. In the blankets of the EBR-II problem, the maximum error in pin segment powers was increased to ~8%, mainly due to the lower power level and the relatively large error in the nodal power of the VARIANT solution.

Coupled Three-Dimensional Neutronics and Thermal-Hydraulics Analysis for SCWR Core Typical Transients

Transient performance of China supercritical water-cooled reactor (SCWR) with the rated electric power of 1000 MWel (CSR1000) core during some typical transients, such as control rod (CR) ejection and uncontrolled CR withdrawal, is analyzed and evaluated with the coupled three-dimensional neutronics and thermal-hydraulics SCWR transient analysis code. The 3D transient analysis shows that the maximum cladding surface temperature (MCST) retains lower than safety criteria 1260 °C during the process of CR ejection accident, and the MCST retains lower than safety criteria 850 °C during the process of uncontrolled CR withdrawal transient. The safety of CSR1000 core can be ensured during the typical transients under the salient fuel temperature and water density reactivity feedback and the essential reactor protection system.