Neutronic Coupling Research Articles

New physics at a neutron beam dump

We find a new utility of neutrons, usually treated as an experimental nuisance causing unwanted background, in probing new physics signals. The new physics signals can either be radiated from neutrons (neutron bremsstrahlung) or appear through secondary particles from neutron-on-target interactions, dubbed “neutron beam dump.” As a concrete example, we take the FASER/FASER2 experiment as a “factory” of high-energy neutrons that interact with the iron dump. We find that if new physics particles had the same interaction strength to protons and neutrons, their production rate via neutron-initiated bremsstrahlung would be comparable to that via proton-initiated ones, in terms of the resulting flux and the range of couplings that can be probed. The neutron bremsstrahlung can be used, for instance, to probe dark gauge bosons with nonzero neutron coupling. In particular, we investigate protophobic gauge bosons and find that FASER/FASER2 can probe new parameter space. We also illustrate the possibility of neutron-induced secondary particles by considering axionlike particles with electron couplings. We conclude that the physics potential of FASER/FASER2 in terms of new physics searches can be greatly extended and improved with the inclusion of neutron interactions. Published by the American Physical Society 2024

Probing ultralight isospin-violating mediators at GW170817

Gravitational wave (GW) signals arising from binary neutron star mergers offer new, sensitive probes to ultralight mediators. Here we analyze the GW signals in the GW170817 event detected by the LIGO/Virgo collaboration to impose constraints on the ultralight isospin-violating mediator that has different couplings to protons and neutrons. Neutron stars, which primarily consist of neutrons, are the ideal places to probe the isospin-violating mediator. Such a mediator can significantly alter the dynamics of the binary neutron star mergers, through both the long-range Yukawa force and the new dipole radiation. We compute the gravitational waveform by taking into account the new physics effects due to the isospin-violating mediator and use the Bayesian inference to analyze the gravitational wave data in the GW170817 event. We find that although the current fifth force experiments (including MICROSCOPE and EW) often provide more stringent constraints than the GW170817 data, in the parameter space where the isospin-violating force is completely screened by the Earth (namely, the Earth is charge neutral under this force), the GW170817 data offer the leading constraints: the upper bound on the neutron coupling is fn ≲ 10−19 in the mediator mass range of ≃ (3 × 10−16, 5 × 10−14) eV.

Low-mass dark sector searches with deuteron photodisintegration

Recent years have seen much activity in searches for dark-sector messenger particles in the 10–100 MeV mass range, especially in view of a potential new light boson conjectured by the ATOMKI Collaboration, X17. Under the assumption that the messenger particle has definite parity and either zero or unit spin, quite stringent bounds already exist on its coupling to electrons and protons. Equally stringent bounds on the neutron coupling do not exist yet, but are nonetheless desirable. We explore how measurements of deuteron photodisintegration with a quasifree neutron can yield bounds on the neutron coupling, and compute projections for a potential measurement at the low-energy high-intensity electron scattering experiment MAGIX@MESA. The projected bounds are found to be competitive for an axial-vector or pseudoscalar scenario, but not for a vector or scalar scenario. Published by the American Physical Society 2024

A new method of fuel pebble temperature calculation based on neutronic and thermal coupling equivalent model

As one of the six reactor types in the fourth-generation nuclear energy systems, the Very High Temperature Gas-cooled Reactor (VHTR) possesses inherent safety features. One VHTR fuel pebble contains tens of thousands of randomly distributed TRISO fuel particles, displaying a double heterogeneity characteristic. Current research models employ the homogenized method for fuel pebble temperature calculation, which neglected the neutronic and thermal (N/T) coupling effect and double heterogeneity characteristic. This study proposes a fuel pebble coupled equivalent method (PCEM) for temperature calculation. The method is based on the effective thermal conductivity obtained by the fine structure fuel pebble N/T coupling method. The new method is evaluated on steady-state and transient VHTR multi-scale neutronic and thermal hydraulic (N/TH) coupling calculation. The results show that the traditional temperature calculation method is conservative. Under steady-state conditions, the maximum degree of conservatism of the fuel pebble average temperature and peak temperature are 0.6 K and 14.2 K respectively. For the sudden drop in coolant flow rate transient condition, the average temperature and peak temperature of the fuel pebble calculated by PCEM are smaller than the values calculated by traditional method at the same time. Due to the influence of the fuel Doppler effect, the transient response of the traditional calculation method is slower.

Development and application in multiscale and multiphysics methodologies in Spain: Present and future trends

In the field of reactor physics analysis, the coupling of multiple physics phenomena plays a crucial role in enhancing the accuracy of predictions and understanding complex systems. Multiphysics refers to the study and analysis of physical phenomena that involve multiple interconnected physical processes occurring simultaneously in a nuclear system to ensure essential aspects of reactor design, safety analysis, and optimization, among others. Multiphysics encompasses various domains of physics, such as Thermal-Hydraulics, Neutronics, Structural Mechanics, Radiation Transport, Chemistry, and Materials Science. On the other hand, the Multiscale approach involves combining high-fidelity models with lower-fidelity ones to accurately capture the relevant physical phenomena in a Nuclear Power Plant (NPP). Multiscale involves three main scales: system level, core level, and detailed analysis scale. This paper presents an overview of the main research performed by Spanish groups in the field of multiphysics and multiscale calculations, particularly on thermal–hydraulic analysis. This revision is focused mainly on two key aspects: thermal–hydraulic multiscale coupling and thermal–hydraulic / neutronic coupling. The paper also includes the expected future trends in this field.

Neutronic-coupled thermal hydraulic analysis of flow blockage in the hottest wire-wrapped LBE-cooled fuel assembly

Flow blockage in lead-cooled fast reactors is considered a realistic accident for fuel assemblies, whereas reactivity feedback and axial power profiles are considered in few studies due to the complexity and lack of neutronic modules in commercial computational fluid dynamic software. This study proposes a neutronic-coupled thermal–hydraulic model to analyze the coupling effect on flow blockage in a 19-pin wire-wrapped fuel assembly. Neutronic and thermal hydraulics were validated by comparing the reported benchmark and experimental data, respectively. Simulation results with and without neutronic coupling schemes were compared, and case studies of different reactivity insertions were conducted. The wall overheating of blockages with coupling schemes is more conservative than the constant heat source condition with the lower dimensionless temperature T¯. After a step change of cross-section, the response of Rw and T¯ shows synchronous behavior. For a larger reactivity insertion, it leads to higher wall heat flux, lower wall thermal resistance and lower dimensionless temperature for blockage regions.

Development of 3D transient neutronics and thermal-hydraulics coupling procedure and its application to a fuel pin analysis

This paper develops a three-dimensional fine-mesh coupled neutronics and thermal-hydraulics platform based on the open-source software OpenFOAM. For the neutronics, a three-dimensional multi-group multi-region neutron diffusion solver is developed for steady and transient problems with macroscopic cross-sections generated by the Monte Carlo code OpenMC. By modeling and calculating three typic neutronic benchmark problems, the results are compared with that of other programs to verify the correctness of this solver to solve neutron diffusion equations. The thermal-hydraulics is solved by the built-in solid-fluid conjugated heat transfer solver for steady or transient fluid flow and solid heat conduction between different regions. The unified solution of a coupled neutronic and thermal-hydraulic problem is implemented in the solver with the same time step size and the same set of meshes. The feasibility and rationality of this coupling procedure is verified through the simulation the 5×5 PWR assembly. Furthermore, the operator-splitting semi-implicit (OSSI) method and the fixed-point implicit (FPI) method have been applied to the coupling procedure. The performance of the above two neutronic and thermal-hydraulic coupling algorithms is evaluated by simulating the coolant inlet temperature decrease accident in a single fuel pin model of the pressurized water reactor. The numerical results indicate that the FPI coupling method outperforms the OSSI coupling method due to its accuracy and stability.

A multi-region algorithm for N-TH coupling calculation and its application to nuclear thermal propulsion reactor

Physical field mapping problem is generally faced in multi-physics coupling calculations, and multi-region (usually divided by different physical properties) coupling further increases the problem’s difficulty. The multi-region algorithm can achieve the solid–fluid conjugate heat transfer for the multi-region Neutronic and Thermal-Hydraulic (N-TH) coupling calculation. This paper proposed a multi-region algorithm for solving neutronic fields and further coupled it with multi-region heat transfer in OpenFOAM. The multi-region algorithm solves the neutron transport equation region-by-region and establishes a new coupling boundary condition to eliminate the obstacles to communication between adjacent regions. The multi-region algorithm is developed in OpenFOAM as a neutron transport solver and verified by the 2D-IAEA, 2D-TWIGL, and the 3D-TAKEDA benchmarks. The multi-region neutron transport solver and the original Conjugate Heat Transfer (CHT) solver are coupled to establish a multi-physics coupling platform. Hence, the multi-physics solver can provide pin-by-pin neutronics and thermal hydraulics results for multiple regions (including solid or fluid regions) based on the same set of grids. The geometric adaptability and parallel performance of the coupling platform are superior, which depends on the inherent characteristics of the OpenFOAM platform. The coupling platform is used to perform N-TH coupling calculations for the Nuclear Thermal Propulsion (NTP) reactor's Low Enriched Uranium (LEU) assemblies, demonstrating the coupling calculation capability of the multi-region algorithm for complex reactor geometry systems. The calculation results show intense N-TH coupling effects in the NTP assemblies, which can provide an essential reference for nuclear thermal propulsion reactor design.

Parametric study of effective thermal conductivity for VHTR fuel pebbles based on a neutronic and thermal coupling method

Pebble bed very high-temperature gas-cooled reactor (VHTR) has many economic and safety advantages, such as high safety and high outlet temperature. It uses fuel pebbles with Tristructural-isotropic (TRISO) fuel particles. The effective thermal conductivity of fuel pebbles is a key parameter to evaluate the peak temperature and average temperature of the fuel pebbles. The currently available effective thermal conductivity models do not consider the internal heat sources and the complex neutronic and thermal (N/T) coupling effects in the fuel pebbles. In this study, a three-dimensional fuel pebble heat transfer and a Monte Carlo neutronic coupled model were developed. The model has a random distribution of TRISO fuel particles and an accurate geometric model. The results show that traditional multi-scale heat conduction methods based on the existing effective thermal conductivity model, such as those based on the Chiew and Glandt models, are conservative in predicting the temperatures of VHTR fuel pebbles. The degree of conservatism is affected by fuel pebble power and packing fraction. Under the N/T coupling condition, the effective thermal conductivity is not much affected by the power of the fuel pebble but decreases with the increase of the fuel pebble boundary temperature. Geometric parameters affect the internal temperature distribution and peak temperature by changing the uniformity of the heat source distribution.

Improved design of LBE cooled solid reactor using 3D neutronics thermal-hydraulics coupling method

The goal of this study is to improve a micro reactor core into uniform power distribution, a smaller and safer core. Neutronic thermal-hydraulics coupling is an effective means to study the characteristics of microreactors. This paper performs an improvement work of lead–bismuth eutectic alloy (LBE) cooled solid core using the three-dimensional neutronics/thermal-hydraulics coupling method. Based on the Monte Carlo code RMC and the three-dimensional thermal-hydraulics code STAR-CCM+, the coupling code processes the power distribution, fuel temperature, coolant temperature, and coolant density. The improvement of the solid core is carried out in these aspects: reflector material, reflector dimensions, fuel materials, coolant channel dimensions, the volume fraction of fuel phase in dispersion fuel, and critical enrichment. This paper studies the effect of reflector material and dimensions on the solid core and its mechanism. After improvement, the power peaking factor is effectively reduced, while the size and weight of the solid core are also reduced than before due to the excellent ability of beryllium oxide (BeO), making it compact and lightweight. In addition, the core eigenvalues keff and thermal characteristics of different ceramic fuels, uranium dioxide (UO2), uranium nitride (UN), and uranium carbide (UC), are compared. UO2 was replaced by UN, which improves the thermal conductivity and reduces the maximum temperature of the solid core. At the same pressure difference, the mass flow rate of the coolant and temperature distribution of the fuel and the variation trend of neutron characteristics are studied when the coolant channel diameter changes. The results indicate that the coolant channel diameter of 1.775 cm has the best safety effect on the core. The variation of critical enrichment and thermal conductivity of the fuel at different volume fractions is compared, and 60 % is chosen as the volume fraction of the fuel phase. The final design of the shutdown rod and control drums is performed to ensure that the core is subcritical under all conditions. The improved core can operate at full power for five years and is characterized by miniaturization, high thermal conductivity, low volumetric swelling rate, flattened power distribution, and abundant reactivity control measures.

Optimization Design of Lbe Cooled Solid Reactor Using 3d Neutronic Thermal-Hydraulics Coupling Method

Optimization Design of Lbe Cooled Solid Reactor Using 3d Neutronic Thermal-Hydraulics Coupling Method

Neutronic and Thermal-Mechanical Coupling Schemes for Heat Pipe-Cooled Reactor Designs

Abstract Space fission power systems can enable ambitious solar-system and deep-space science missions. The heat pipe-cooled reactor is one of the most potential candidates for near-term space power supply, featuring safety, simplicity, reliability, and modularity. Heat pipe-cooled reactors are solid-state and high-temperature (up to 1500 K) reactors, where the thermal expansion is remarkable and the mechanical response significantly influences the neutronics and thermal analyses. Due to the considerable difference between heat pipe-cooled reactors and traditional water reactors in the structure and design concept, the coupling solutions for light water reactors cannot be directly applied to heat pipe-cooled reactor analyses. Therefore, a new coupling framework and program need to consider the coupling effects among neutronics, heat transfer, and mechanics. Based on the Monte Carlo program rmc and commercial finite element program ansysmechanicalansys parametric design language (APDL), this work introduces the three coupling fields of neutronics (N), thermal (T), and mechanics (M) for heat pipe cooled reactors. Besides, the finite element method and the Monte Carlo program use different meshes and geometry construction methods. Therefore, the spatial mapping and geometry reconstruction are also essential for the N/T-M coupling, which is discussed and established in detail. Furthermore, the N/T-M coupling methods are applied to the preliminary self-designed 10 kWel space heat pipe cooled reactor. Coupling shows that the thermal-mechanical feedback in the solid-state reactor has negative reactivity feedback (about −2000 pcm) while it has a deterioration in heat transfer due to the expansion in the gas gap.

A twin assemblies benchmark for coupling behavior and kinetics parameters calculations

Traditionally, stochastic analysis of reactor kinetics behavior had always been a prohibitively expensive problem, in terms of the necessary calculation time. Recent advances in computational methods, however, have made it possible for several widely used Monte Carlo codes to incorporate methods that enable the study and simulation of such topics, with one of them being reactor space–time kinetics or STK, which until recently was limited to deterministic codes only. The present paper uses the Transient Fission Matrix (TFM) model, a hybrid method that uses a system response obtained through Monte Carlo and stored in fission and time matrices as input for deterministic calculations. The result enables a view of the system’s behavior in terms of its neutronic coupling and its kinetic parameters, namely the neutron propagation probability, generation time, lifetime and delayed neutron fraction. The TFM method is applied to a coupled core configuration with the purpose of generating a numerical benchmark. The Serpent 2 Monte Carlo code is used for the stochastic part of the calculation. The geometry consists of two fuel assemblies placed in a light water tank, with a water blade of varying width between them. The process consists of adjusting the width (thus the separation) of the two assemblies and then extracting TFM, flux and fission rate results. The blade width ranges between 0 cm and 20 cm. The system behavior, as well as the coupling effects between the two assemblies, are then analyzed and discussed. By obtaining the system’s eigenvalues from the transient fission matrices, the degree of coupling can be assessed. As the assemblies move further apart, the system slowly transitions from a single core, to two only loosely coupled ones. This study enables to produce a benchmark to compare future calculations schemes against and predefine an innovative way of designing high dominant ratio configurations. An actual experimental program could be led in ad hoc zero power reactor able to perform them, such as the KUCA reactor of Kyoto University, where similar (in their philosophy) experiments were done in the 90’s.

Nanofluid application for heat transfer, safety, and natural circulation enhancement in the NuScale nuclear reactor as a small modular reactor using computational fluid dynamic (CFD) modeling via neutronic and thermal-hydraulics coupling

In recent years, “Nanofluid” and “Energy-Efficient Cooling” have been achieved growing attention for use in new nuclear reactor technology. Also, the assurance of reactor safety has become of top priority in nuclear energy development because of the significant role of nuclear power in providing worldwide electricity. The use of nanofluid coolant is an effective way to enhance safety margins and heat transfer performance. Studies also prove that nanofluids are promising cooling fluids for applying as a future coolant in nuclear reactors. The current paper investigates the nanofluid effects as a coolant on the velocity of bulk, pressure drop, heat transfer coefficient, power peaking factors, and minimum departure from nucleate boiling ratio in a NuScale reactor, a natural circulation pressurized water reactor (PWR). It also analyzes the thermal characteristics of Al2O3 nanofluid with specific attention to their boiling heat transfer behavior. The study first presents the NuScale designed core with the use of water and different concentrations of Al2O3 nanofluid (0.001–10% volume fractions and 10–90 nm particle sizes). Then it shows the simulated equivalent cell with surrounding coolant utilizing computational fluid dynamic code (CFD). The study also describes the outcomes of replacing water with nanofluid coolant on natural circulation parameters, heat transfer coefficient, and minimum departure from nucleate boiling ratio with changes in nanoparticle concentrations and sizes. The results demonstrate the potential of this innovative coolant to enhance the minimum departure from nucleate boiling ratio and heat transfer coefficient. Overall, the presence of Alumina nanoparticles in water improves thermal performance and safety. The investigation also shows that the nanoparticles do not change the power peaking factor and neutronic performance significantly.

Low-energy spectra of mirror mass-19 nuclei with a collective coupled-channel scattering model

The spectra of mass-19 nuclei are unusual and pose a challenge for theoretical models of their structure. Herein the Multi-Channel Algebraic Scattering (MCAS) method has been used with a collective (vibrational model) description of the low-excitation states of $$^{18}$$ O and $$^{18}$$ Ne to describe the spectra of four mass-19 nuclei, specifically the mirror pairs ( $$^{19}$$ O, $$^{19}$$ Na) and ( $$^{19}$$ F, $$^{19}$$ Ne). This coupled-channel approach allows for the effects of the Pauli principle to be included using non-local orthogonalizing pseudo-potentials. By the coupling of neutrons to the core nuclei, we obtain states in $$^{19}$$ O (a neutron coupled to $$^{18}$$ O) and $$^{19}$$ Ne (a neutron coupled to $$^{18}$$ Ne). Mirror symmetry and addition of the Coulomb interaction gives states in $$^{19}$$ F (a proton coupled to $$^{18}$$ O) and $$^{19}$$ Na (a proton coupled to $$^{18}$$ Ne). However, the same compound nuclei may be described as the coupling of more complicated nuclear clusters. We obtain the states in $$^{19}$$ F as the coupling of an $$\alpha $$ to $$^{15}$$ N, as well as the coupling of $$^3$$ H to $$^{16}$$ O. Similarly, states in $$^{19}$$ Ne have been obtained from the coupling of an $$\alpha $$ to $$^{15}$$ O and of the coupling of $${}^3$$ He to $$^{16}$$ O. Finally, as the method allows extraction of scattering matrices, it has been used to evaluate cross sections for the elastic scattering of low-energy neutrons from $$^{18}$$ O and of low-energy $$\alpha $$ particles from $$^{15}$$ O.

Pronghorn: A Multidimensional Coarse-Mesh Application for Advanced Reactor Thermal Hydraulics

This paper presents an overview of Pronghorn, a multiscale thermal-hydraulic (T/H) application developed by Idaho National Laboratory and the University of California, Berkeley. Pronghorn, built on the open-source finite element Multiphysics Object-Oriented Simulation Environment (MOOSE), leverages state-of-the-art physical models, numerical methods, and nonlinear solvers to deliver fast-running advanced reactor T/H simulation capabilities within a modern software engineering environment. This work summarizes the physical models, multiphysics and multiscale coupling, and numerical discretization in Pronghorn with emphasis on our initial target application to pebble bed reactors (PBRs). A diverse set of applications are shown to depressurized natural circulation in the SANA experiments, forced convection in the Pebble Bed Modular Reactor, three-dimensional (3-D)/one-dimensional coupling of Pronghorn and RELAP-7 systems T/H for loop analysis in the High Temperature Reactor Power Module, and forced convection in the Mark-1 Pebble Bed Fluoride-Salt-Cooled High-Temperature Reactor. A multiphysics coupling of Pronghorn, RELAP-7, and Griffin deterministic neutronics for a gas-cooled PBR demonstrates the capability of the MOOSE framework for reactor design calculations. These applications highlight the verification and validation underlying Pronghorn’s software development while emphasizing features that improve upon capabilities offered by legacy tools in areas such as 3-D unstructured meshing, physics modeling, and multiphysics coupling.

RMC/ANSYS MULTI-PHYSICS COUPLING SOLUTIONS FOR HEAT PIPE COOLED REACTORS ANALYSES

The heat pipe cooled reactor is a solid-state reactor using heat pipes to passively transfer heat generated from the reactor, which is a potential and near-term space nuclear power system. This paper introduces the coupling scheme between the continuous energy Reactor Monte Carlo (RMC) code and the finite element method commercial software ANSYS. Monte Carlo method has the advantages of flexible geometry modeling and continuous-energy nuclear cross sections. ANSYS Parametric Design Language (APDL) is used to determine the detailed temperature distributions and geometric deformation. The on-the-fly temperature treatment of cross sections was adopted in RMC code to solve the memory problems and to speed up simulations. This paper proposed a geometric updating strategy and reactivity feedback methods for the geometric deformation of the solid-state core. The neutronic and thermal-mechanical coupling platform is developed to analyze and further to optimize the heat pipe cooled reactor design. The present coupling codes analyze a 2D central cross-section model for MEGAPOWER heat pipe cooled reactor. The thermal-mechanical feedback reveals that the solid-state reactor has a negative reactivity feedback (~1.5 pcm/K) while it has a deterioration in heat transfer due to the expansion.

Hot and average fuel sub-channel thermal hydraulic study in a generation III+ IPWR based on neutronic simulation

The Integral Pressurized Water Reactors (IPWRs) as the innovative advanced and generation-III + reactors are under study and developments in a lot of countries. This paper is aimed at the thermal hydraulic study of the hot and average fuel sub-channel in a Generation III + IPWR by loose external coupling to the neutronic simulation. The power produced in fuel pins is calculated by the neutronic simulation via MCNPX2.6 then fuel and coolant temperature changes along fuel sub-channels evaluated by computational fluid dynamic thermal hydraulic calculation through an iterative coupling. The relative power densities along the fuel pin in hot and average fuel sub-channel are calculated in sixteen equal divisions. The highest centerline temperature of the hottest and the average fuel pin are calculated as 633 K (359.85 °C) and 596 K (322.85 °C), respectively. The coolant enters the sub-channel with a temperature of 557.15 K (284 °C) and leaves the hot sub-channel and the average sub-channel with a temperature of 596 K (322.85 °C) and 579 K (305.85 °C), respectively.It is shown that the spacer grids result in the enhancement of turbulence kinetic energy, convection heat transfer coefficient along the fuel sub-channels so that there is an increase in heat transfer coefficient about 40%. The local fuel pin temperature reduction in the place and downstream the space grids due to heat transfer coefficient enhancement is depicted via a graph through six iterations of neutronic and thermal hydraulic coupling calculations. Working in a low fuel temperature and keeping a significant gap below the melting point of fuel, make the IPWR as a safe type of generation –III + nuclear reactor.

Can neutron disappearance/reappearance experiments definitively rule out the existence of hidden braneworlds endowed with a copy of the Standard Model?

Many works, aiming to explain the origin of dark matter or dark energy, consider the existence of hidden (brane)worlds parallel to our own visible world — our usual Universe — in a multidimensional bulk. Hidden braneworlds allow for hidden copies of the Standard Model. For instance, atoms hidden in a hidden brane could exist as dark matter candidates. As a way to constrain such hypotheses, the possibility for neutron–hidden neutron swapping can be tested thanks to disappearance-reappearance experiments also known as passing-through-walls neutron experiments. The neutron-hidden neutron coupling [Formula: see text] can be constrained from those experiments. While [Formula: see text] could be arbitrarily small, previous works involving a [Formula: see text] bulk, with DGP branes, show that [Formula: see text] then possesses a value which is reachable experimentally. It is of crucial interest to know if a reachable value for [Formula: see text] is universal or not and to estimate its magnitude. Indeed, it would allow, in a near future, to reject definitively — or not — the existence of hidden braneworlds from experiments. In the present paper, we explore this issue by calculating [Formula: see text] for DGP branes, for [Formula: see text], [Formula: see text] and [Formula: see text] bulks. As a major result, no disappearance-reappearance experiment would definitively universally rules out the existence of hidden worlds endowed with their own copy of Standard Model particles, except for specific scenarios with conditions reachable in future experiments.

A New Precursor Integral Method for Solving Space-Dependent Kinetic Equations in Neutronic and Thermal-Hydraulic Coupling System

The accurate prediction of the neutronic and thermal-hydraulic coupling system transient behavior is important in nuclear reactor safety analysis, where a large-scale nonlinear coupling system with strong stiffness should be solved efficiently. In order to reduce the stiffness and huge computational cost in the coupling system, the high-performance numerical techniques for solving delayed neutron precursor equation are a key issue. In this work, a new precursor integral method with an exponential approximation is proposed and compared with widely used Taylor approximation-based precursor integral methods. The truncation errors of exponential approximation and Taylor approximation are analyzed and compared. Moreover, a time control technique is put forward which is based on flux exponential approximation. The procedure is tested in a 2D neutron kinetic benchmark and a simplified high-temperature gas-cooled reactor-pebble bed module (HTR-PM) multiphysics problem utilizing the efficient Jacobian-free Newton–Krylov method. Results show that selecting appropriate flux approximation in the precursor integral method can improve the efficiency and precision compared with the traditional method. The computation time is reduced to one-ninth in the HTR-PM model under the same accuracy when applying the exponential integral method with the time adaptive technique.