Neutronic Parameters Research Articles

Neutronic design performance of 100 MWe MSR with thorium-enriched uranium and thorium-plutonium-minor actinide fuel

Molten Salt Reactors offer fuel flexibility, allowing the use of different starting fuels, including U-233, U-235, and Pu-239. U-233 is not a fissile nuclide that is commercially accessible. However, it can be produced through the utilization of the thorium cycle. In this paper, the neutronic performance of MSR with a power output of 100 MWe is analyzed for Th-U(enriched) and Th-Pu-MA as a starting fuel. The fuel composition is varied to show the neutronic parameter change and obtain the reactor criticality for five years of operation time. The calculation used SRAC Program and JENDL 4.0 as nuclear data libraries. It was found that the reactor with a Th-U fuel case can be operated in critical condition with a minimal fuel loading of UF4 of 7.30 mol%. Similarly, in the Th-Pu-MA fuel case, a minimal amount of PuF4 is at least 3.87 mol%. The Th-U fuel case exhibits a higher initial keff and reactivity swing compared to the Th-Pu-MA fuel case. At the end of the reactor cycle, the Th-U fuel produces several plutonium isotopes, such as Pu-238, Pu-239, Pu-240, Pu-241, and Pu-242. In contrast, the density of plutonium isotopes in the Th-Pu-MA fuel decreases during operation such as Pu-239 and Pu-241. Therefore, using Th-Pu-MA fuel is a viable approach to reducing spent nuclear fuel, as it reduces the number of plutonium isotopes.

Implications of using nanoparticles on the performance and safety of nuclear systems

Nanotechnology, and particularly nanoparticles (NPs), command significant attention across diverse fields such as medicine, chemistry, and engineering. NPs are integrated into a base material through various techniques, such as dispersion, coating, and encapsulation. This review intends to offer an evaluation of the implementation of NPs in nuclear energy and radiation protection. Particular focus is given to nanofluids - suspensions of NPs in a base fluid - and their impact on the thermal–hydraulic and neutronic attributes of nuclear reactor cores. Discussion centers on how substituting water with nanofluids as a coolant can modify these characteristics. Computational modeling and experimental methods of nanofluids are examined, highlighting an enhanced heat transfer capacity under both regular and emergency reactor conditions. The influence of nanofluids on neutronic parameters, aiming for optimized operation and safety of nuclear reactors, is also explored. Furthermore, the potential of NPs for radiation shielding is considered, underlining their capability to attenuate radiation effects. Examples are drawn from the addition of NPs to containment structures in nuclear facilities and their use in radiotherapy, as well as enhancing radiation shielding in imaging techniques. Finally, the review encompasses the effects of nanofluids on surface oxidation and corrosion. The importance of these factors lies in their substantial impact on heat transport properties and potential for corrosive damage.

Characterization of neutron spectrum parameters in PFTS irradiation channel of KAMINI reactor and validation of Westcott formalism using k0 IM-NAA

Characterization of neutron spectrum parameters in PFTS irradiation channel of KAMINI reactor and validation of Westcott formalism using k0 IM-NAA

Effect of cross-sections data on calculated static neutronic parameters of PWR MOX/UO2 core transient benchmark case using NODAL3 code

This paper describes the effect of cross-section data generated by several codes on calculated neutronic parameters. The Pressurized Water Reactor Mixed Oxide and Uranium Oxide (PWR MOX/UO2) Core Transient Benchmark case was chosen because it has been used widely to validate neutronic codes. The cross-section data in this study will be generated by SRAC, Serpent, and HELIOS codes. The NODAL3 code will be used to calculate neutronic parameters from each cross-section. The neutronic parameters calculated by NODAL3 are the effective multiplication factor (keff), control rod worth, critical boron concentration, and power distribution under Hot Zero Power (HZP) conditions. The Power-Weighted Error (PWE) and Error-Weighted Error (EWE), as a measure of the relative error in fuel assembly power, are less than 5 %, indicating that the calculation is consistent with DeCART as a reference. The difference in calculated radial power peaking factor for all three cross-sections to reference data reaches 6.284 % (G-3), 8.438 % (G-3), and 10.998 % (C-7), respectively, for SRAC, Serpent, and HELIOS. The axial power distribution calculated by NODAL3 at the top and bottom of the reactor core has a relative error that peaked at 16.60 %, 13.86 %, and 10.20 %, respectively, for cross-sections provided by SRAC, Serpent, and HELIOS. Further improvements are needed for NODAL3 by applying various discontinuity factors to improve its performance.

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.

Modeling and characterization of beryllium reflector elements under irradiation conditions

This study aims at modeling the beryllium reflector poisoning under neutron irradiation conditions and calculating the impact of beryllium poisoning on the core parameters of ETRR-2 research reactor. The CITVAP code was used to calculate the neutron flux and parameters of ETRR-2 core with beryllium reflector elements. The neutron flux in each reflector element was calculated to solve the modeling equations for the atomic densities of lithium-6 (6Li), tritium-3 (3H), and helium-3 (3He) using the BERYL program. The results are discussed based on CITVAP calculations of the core excess reactivity and cycle length Full Power Days (FPD). Possible solutions to minimize the degradation due to beryllium poisoning are also discussed and compared based on calculations.

Comparative analysis of fuel burnup calculations of fourth-generation European fast reactors

Abstract The sustainability of nuclear energy implies the continuous research and development of fast reactor technology. This work represents a comparative analysis of fuel burnup calculations of three commercial-level fast spectrum concepts which are under research and development in Europe. The investigated designs are the European Sodium-cooled Fast Reactor, the European Lead-cooled Fast Reactor, and the European Gas-cooled Fast Reactor. MCNPX transport code was used to design three representative fuel assemblies of these fourth-generation concepts to analyze and compare their neutronic and safety parameters. The neutronic and safety analysis in this work includes the evolution of infinite multiplication factor and fissile inventory, neutron yield, the average energy of neutrons causing fission, neutron energy spectrum, neutron generation time, Doppler reactivity effect, effective delayed neutron fraction, and coolant void reactivity coefficient. The fuel burnup results showed that the ESFR is superior regarding the reactivity swing and breeding of fissile 239Pu. On the other hand, analysis of safety parameters results showed that they depend mainly on fuel composition rather than on other design specifications.

Neutronic and thermal hydraulic analysis of a PWR fuel assembly using uranium mononitride

The nitride fuel is currently considered as a material to replacement of uranium dioxide in Light Water Reactors (LWR), mainly due to its higher thermal conductivity when compared with the oxide fuels and similar melting point. This behavior may contribute to larger heat conduction in the fuel assemblies and contributes to increase the safety margins to fuel melting during operation. In this way, the present work aims to contribute to studies on the use of uranium mononitride (UN) in LWRs. Neutronic and thermal hydraulic analysis were performed for a typical assembly of Pressurized Water Reactor (PWR) using uranium dioxide (UO2) and uranium mononitride (UN) fuels. The SCALE 6.0 (Standardized Computer Analyses for Licensing Evaluation – version 6.0) and STHIRP (Thermal Hydraulic Simulation of Research Reactors) codes were used to calculate the neutronic and thermal hydraulic parameters, respectively. The goal is to verify the possible use of UN in a PWR. Comparing the evaluated fuels, the largest thermal conductivity of UN contributes to largest heat conduction in the pellet zone and to lowest fuel temperature. However, the UN presents highest production of transuranic elements and considerable 14C production into UN during the burnup. Results of neutronic and thermal hydraulic parameters behavior for both fuels are presented in details in this work as well as carefully analyzed.

Mechanic-elastic properties and radiation attenuation efficiency of TeO2/WO3/K2O composite glass systems for nuclear and medical application

WO3 effects on neutron and ionizing radiation defending factors of ternary tellurite-based glass blocks with molecular formula 80TeO2 –(20-x)WO3 – xK2O; x = 0–20 mol% (denoted as TKW-glass) has been reported via Phy-X theoretical calculations and Geant4 simulation code. Correlations between shielding factors and kinetics properties of the investigated glasses at different photon energy have been examined. The highest values of mass (MAC) attenuation coefficient were noted at 15 keV of the examined TKW-glass materials with the values of 38.408, 44.388, 49.855, 54.872, 59.492 cm2/g for TKW-0, TKW-5, TKW-10, TKW-15, and TKW-20, respectively. Generally, these values of the TKW-glasses obey the sequence: (TKW-0)MAC< (TKW-5)MAC < (TKW-10)MAC < (TKW-15)MAC < (TKW-20)MAC. The highest mean free path (MFP) values of TKW-glasses were registered at 15 MeV with the values of 6.101, 5.591, 5.097, 4.647, and 4.302 cm for TKW-0, TKW-5, TKW-10, TKW-15, and TKW-20, respectively. The two parameters half value layer (HVL) and MFP follow the pattern: (TKW-0)HVL, MFP > (TKW-5)HVL,MFP > (TKW-10)HVL,MFP > (TKW-15)HVL,MFP > (TKW-20)HVL,MFP. The maximum values of effective atomic number (EAN) took place at gamma energy of 15 keV corresponding to 44.35, 48.86, 52.63, 55.83, and 58.58 for TKW-0, TKW-5, TKW-10, TKW-15, and TKW-20, respectively. The trend of the buildup factors was similar for all of the glass specimens. The fast neutron removal cross-section (ΣR) enhanced as WO3 content increased in the specimens. Thus, the peaked value of ΣR is 0.1059 cm−1 was noted in the TKW-20 sample. Mechanical properties, neutron and γ-rays protection parameters were observed to improve with enhanced WO3 mol% in the TKW-glasses. The current results bear their utilization for neutron and gamma protection purposes.

Design and optimization of dual-cooled fuel assembly in a 12×12 configuration for NuScale SMR based on neutronic-thermohydraulic parameters using the combined ANN-GA approach

The Small Modular Reactors (SMRs) use conventional solid fuels. To increase the safety of SMRs, Dual-Cooled Fuel (DCF) can be used. The DCF with two heat transfer surfaces improves heat removal from the fuel. For the first time, this study investigated the use of DCF in a new fuel assembly configuration for the NuScale reactor. Important neutronic and thermohydraulic (N–Th) parameters are obtained. Finally, the optimal DCF rod in a 12 × 12 configuration for the NuScale reactor has been presented. For this purpose, initially, a DCF assembly in a 12 × 12 configuration was designed for the reactor. The fuel mass, enrichment of 235U, and dimensions of fuel assemblies and the core are kept fixed. Then, 32 types of fuel rods with different internal radii were proposed for this assembly. Each of these fuels was placed separately in the assembly and simulated by neutronic codes under cold-clean and full power conditions. Also, the designed fuels were simulated using Computational fluid dynamics (CFD) software, and important thermohydraulic parameters were obtained. Also, to select the optimal fuel rod, the Artificial Neural Network (ANN) and Genetic Algorithm (GA) coupled method is used. To investigate the optimal DCF effects on N–Th parameters in the NuScale reactor, the reference reactor was also simulated and compared with the results of optimal DCF. As one of the most important results of this study, the fuel rod with an internal radius of 0.413423 cm was introduced as the optimal fuel. Also, optimal fuel reduces the temperature and pressure drop, and increases the Departure from Nucleate Boiling Ratio (DNBR).

Machine learning approaches to equilibrium burnup analysis for Molten Salt Reactor

During the operation of a liquid-fueled molten salt reactor (MSR), the deep depletion to an equilibrium state can be attained with online refueling and reprocessing techniques. However, the equilibrium burnup simulation of an MSR using current depletion codes requires considerable computational cost and time, and even fails when the equilibrium conditions are unsatisfied. To efficiently evaluate the specific characteristics of the equilibrium burnup in MSRs, a fast-filtering model for equilibrium state and a fast prediction model for equilibrium neutronic properties are developed based on the machine learning (ML) technique. A database of 4860 cases including the neutronic information of equilibrium and non-equilibrium of a fuel assembly with variable geometric parameters in an MSR is first generated, and then the neutronic properties of 232Th-233U, 232Th-EU (enriched uranium) and 232Th-Pu fuel cycles are evaluated. Based on the MSR database, fifteen different ML classification algorithms are implemented to establish the fast-filtering model for equilibrium burnup state. Moreover, a model for the fast prediction of equilibrium neutronic parameters of neutron flux (Φ) and breeding ratio (BR) is developed with fifteen ML regression algorithms. Considering the various performance metrics for measuring the predicted performances of ML models in the classification and regression, the LightGBM (LGBM) model is the most favorable for filtering the burnup state and predicting the neutronic parameters in MSRs. With the test data, the LGBM model can obtain the accuracy of about 0.98 for classification and the relative error of about 1.10% for regression.

Steady-state analysis of ROSTOV-2 benchmark problem using SERPENT-2 and PARCS codes

Recently, a new benchmark problem has been introduced by the OECD/NEA on the Rostov-2 reactor to address the verification and validation of novel High-Fidelity Multi-Physics tools. In line with the first phase of this benchmark, the current research was dedicated to the steady state reactor physics analysis, using “traditional multi-physics” codes for further and future comparisons. A set of homogenized two-group cross-section libraries were developed using SERPENT-2 Monte Carlo code. Some sophisticated techniques have been employed for detailed consideration of radial as well as axial reflectors, spacer grids and assembly discontinuity factor evaluation. Since PARCS code requires the group constants in PMAXS format, the generated libraries were re-formatted using the GenPMAXS interface. The PMAXS library verification, as well as the steady state analysis have been conducted in three operational states of the reactor using PARCS nodal diffusion code, independent SERPENT-2 Monte Carlo simulation and some available reference data. A set of criticality calculations have been conducted to evaluate the total reactivity worth of different control rod groups, the group #10 integral reactivity curve, fuel temperature coefficient and the critical boron concentration in Hot Zero Power (HZP) and Hot Full Power (HFP) states. The maximum reactivity difference between PARCS and SERPENT-2 simulations was less than 100 pcm in all criticality calculations. In HFP state, the relative power distribution in radial and axial directions were calculated and compared. The maximum Power Peaking Factors (PPFs) and the location of the “hot assembly” from PARCS and SERPENT-2 are in good agreement with each other. Finally, the burn-up calculation was conducted with PARCS code to evaluate neutronic parameters at some specific burn-up values. The calculated reactor physics parameters including the critical boron concentration, the power peaking factors, the power and burn-up distributions at the Beginning of Experiment (BOE) were consistent with available reference data.

Neutron multiplication and reactor kinetics parameters in stochastic media based on randomized noise functions

The modelling of stochastic media is of considerable interest for criticality safety and reactor physics applications. Novel methods for modelling highly disordered, random heterogeneous mixtures based on randomized Weierstrass functions and Perlin noise have been implemented in the MONK® Monte Carlo code using Woodcock tracking. Perlin noise in particular is shown to be well suited to the application, efficiently producing random geometries with much greater physical realism than the artificial nature of alternative methods. The effect of geometry randomization on neutron multiplication and reactor kinetics parameters in a simple benchmark configuration has been investigated, and results are compared with previous research based on tessellation algorithms. While the results are found to be broadly consistent with previous work, the new methods presented here offer significant benefits in terms of increased physical realism, conservation of material volume fractions, computational efficiency and ease of use.

Neutronic investigation of enriched gadolinium, natural gadolinium and Zirconium-erbium as burnable poisons in the advanced PWR assemblies by MCNPX code

In Pressurized Water Reactors, burnable absorbers are implemented for controlling the excess reactivity and avoiding the high power peaking factor. A feasibility study was carried out for investigating the effect of enriched gadolinium (157Gd2O3), natural gadolinium (152 - 160Gd2O3) and Zirconium-Erbium (ZrEr2) as burnable absorbers on the safety parameters of advanced PWR assembly. Three poisoned assemblies are modeled and simulated by the well-known MCNPX code. To hold a fair comparison between these assemblies, the author adjusted the concentration of fissile materials and burnable absorbers in each assembly to provide approximately the same initial reactivity. The MCNPX simulation showed that the enriched gadolinium model achieved a number of advantages over Zirconium-Erbium as: 1) prolonging the fuel cycle length by 1.04%, 2) increasing the negativity of moderator temperature coefficient and fuel temperature coefficient by 10.3% and 9.3% at the BOL and this contributes to core safety, 3) minimizing residual reactivity penalty at the higher burnup steps and 4) more hardening of neutron spectrum at the BOL. On the other hand, the Zirconium – Erbium layer presented a number of benefits over enriched gadolinium as 1) reducing the peaking factor by 1.3%, 2.6% and 2% at 20, 30 and 50 GWd/ton, 2) increasing the reactivity worth by 1.6% at the BOL, 3) increasing the power flattening by 7.6% at the EOL, and this is considered an enormous economic benefit. Therefore, the modeling of enriched gadolinium and zirconium erbium in this study is suggested for improving the neutronic and safety parameters of PWR cores.

Neutronic Parameter Analysis of Plate-Type Fueled TRIGA 2000 Reactor by MCNPX

A novel simulation to calculate the neutronic parameters of the TRIGA 2000 reactor using plate-type fuel has been performed. The plate fuel used was produced by the Indonesian Nuclear Industry (PT INUKI) with U3Si2-Al material. Neutronic parameters based on INUKI’s plate-type fuel dimension and the current TRIGA’s configuration were simulated using MCNPX. The simulation was performed by modeling the complete reactor’s configuration on a fresh fuel core state. We obtained the kinetic parameter values from the simulation, i.e., delayed neutron fraction of 8.11×10‑3, a prompt neutron lifetime of 2.0551×10‑4 s, and an average neutron generation time of 1.87×10‑4 s. The excess reactivity of the reactor was 9.02 %Δk/k, while reactivity in the one-stuck-rod state was below ‑0.5 $ with an average value of ‑3.40 %Δk/k (‑4.19 $). The average thermal neutron flux peak occurred at the central irradiation position with the value of 3.0×1013 to 3.1×1013 n/(cm2 s). The reactor has a power peaking factor of 1.379 in the control rod position of 0 % on D3 fuel. The reactor had a negative feedback reactivity coefficient, except for the moderator coefficient. These results suggest that the current configuration of plate-type fuel met the nuclear reactor neutronic safety standards.

Role of TiO2 addition on recycling of TV screen waste glasses: Experimental and theoretical studies on structure and radiation attenuation properties

Using the powder metallurgy, TiO2-doped TV screen glass ceramic samples were prepared with TiO2 concentrations of 0 (S-TV), 10 (S-TV/Ti10), and 20 (S-TV/Ti20) wt%. The TV screen glass was collected from a damaged TV's CRT. The fabricated materials were investigated for their physical, structural, and radiation-shielding attributes. The chemical composition of the prepared glasses was analysed with the aid of the X-ray fluorescence (XRF) method. The crystalline phase and microstructure of the glass ceramics were studied through the X-ray diffraction (XRD) method and scanning electron microscope (SEM), respectively. The density and Vickers hardness of the sample were determined by the Archimedes method and a micro hardness tester, respectively. The absorption parameters of electrons, protons, alpha particles, carbon ions, fast neutrons, and thermal neutron were calculated from various software and analytic expressions. The XRD spectra of the samples showed the presence of distinct crystalline peaks associated with TiO2, SiO2, and Ba2SiO4. The SEM images revealed that crystalline structures nucleated and increased with coarser crystalline grains when the amount of TiO2 was increased. The densities of the glass ceramics were 2.525, 2.546, and 2.285 g/cm3 for the S-TV, S-TV/Ti10, and S-TV/Ti20 samples, respectively while the corresponding Vickers hardness values were 495, 540, and 578 HV. The projected range of electrons, protons, alpha particles, and heavy carbon ions were within the limits 2.78–32182.18 μm, 0.215–1400 μm, 0.154–130.54 μm, and 0.053–14.10 μm in S-TV, 2.76–31991.36 μm, 0.205–1320 μm, 0.147–122.05 μm, and 0.049–13.11 μm in S-TV/Ti10, and 2.84–32824.95 μm, 0.212–1360 μm, 0.151–125.55 μm, and 0.051–13.45 μm in S-TV/Ti20, respectively. Among the three samples, S-TV was the best thermal neutron absorber, while S-TV/Ti10 is most effective in slowing down fast neutrons. The optimum TiO2 doping content for enhancing the charged particle and fast neutron moderating abilities of TV screens is 10 wt%.

Application of SCALE code for reactor physics analysis of the as-built Nigeria miniature neutron source reactor operating with low enriched uranium fuel

This study presents a broad reactor physics analysis including, neutronic, burnup, and decay heat of the ‘as-built’ NIRR-1 LEU core performed using SCALE 6.2.3 code package. First, verification of neutronic parameters of the NIRR-1 was performed by modeling the reactor ‘as-built’ LEU core. Thereafter validation of the model was done by comparing calculated result with measured data during the commissioning of the reactor. The results showed that the estimated neutronic parameters using KENO-VI code of SCALE 6.2.3 code package are in close agreement with the measured data. Hence, KENO-VI can be used for detailed neutronic assessment of MNSRs. The total decay heat at the end of core life after operation for 216 EFPD was estimated to be 1620.15 W and was majorly due to the contribution from fission products which was estimated to be 1608.10 W. This value explains why MNSRs are cooled through natural convection of the coolant during operation and at shutdown conditions. Additionally, the burnup of the core was determined to be 0.53 wt% which is within the expected burnup for a typical MNSRs (less than1 wt%) with a core lifetime of about 200 EFPD (∼10 years). Furthermore, the mass of 239Pu buildup after 216 EFPD was far less than the IAEA category III nuclear material and indicates a huge advantage of research reactor conversion in reducing proliferation risk.

Whole core analysis of the single-fluid double-zone heavy water-moderated molten salt reactor (SD-HWMSR)

Thermal-spectrum molten salt reactor (MSR) concepts usually adopt graphite as a neutron moderator. Although graphite as a moderator has many advantages, it also has drawbacks, including a relatively short lifespan (it has to be replaced), positive temperature feedback, and loss of impermeability due to expansion. Replacing graphite with heavy water in MSRs can effectively solve the problems introduced by graphite. The SD-HWMSR is a Single-fluid Double-zone Heavy Water-moderated Molten Salt Reactor. Optimization of SD-HWMSR’s fuel channel pitch and radius has been conducted in this work. As a result, the SD-HWMSR with a high breeding ratio (1.07465 ± 0.00060), low initial 233U loading (1.43 t), and negative temperature and void reactivity coefficients is put forward. The SERPENT-2 is used to analyze the neutronics parameters of the reactor design. The current work investigates the change in the multiplication factor, breeding ratio, and the accumulation of significant nuclides in the core. The suitable 232Th and 233U refill rates needed to maintain criticality and enable analysis of the whole core of SD-HMMSR are thoroughly determined in this study. Uranium-233 and thorium-232 both have a maximum refill rate of 3.49 and 3.20 kg/d, respectively. While the average refill rate for 233U and 232Th during the 60 years of operation is 1.90 and 2.35 kg/d, respectively. The net production of 233U rises with time, and by the end of the 60 years, it is around 1.98 t. The doubling time for the SD-HWMSR is 31 years, which is consistent with previous results.

Results and lessons learned from the Generation IV SCWR-FQT comprehensive Monte Carlo computational benchmark

A joint European Canadian Chinese development of a supercritical water-cooled small modular reactor technology has been in progress since September 2020 in the framework of a Horizon 2020 project called ECC-SMART. A specific work package has been dedicated to studying the design- and safety-related neutronic parameters and reactor physics behavior of the SCW-SMR to support the pre-conceptual design process. Three Monte Carlo codes, viz., MCNP, OpenMC, and Serpent, were selected for pre-conceptual design applications and code-to-code comparison within the Gen-IV SCWR-FQT reactor physics computational benchmark. The effective multiplication factor, the axial power distribution within the fuel, the axial three-group neutron flux distribution, and the spatial distribution of the energy deposition due to neutron and photon interactions were determined. In this paper, results and lessons learned from this study are presented, and useful considerations are summarized to provide guidance in obtaining consistent results among the three Monte Carlo codes.

Three-Dimensional Surrogate Model Based on Back-Propagation Neural Network for Key Neutronics Parameters Prediction in Molten Salt Reactor

The simulation and analysis of neutronics parameters in Molten Salt Reactors (MSRs) is fundamental for the design of the reactor core. However, high-fidelity neutron transport calculations of the MSR are time-consuming and require significant computational resources. Artificial neural networks (ANNs) have been used in various industries, and in recent years are increasingly introduced in the nuclear industry. Back-Propagation neural network (BPNN) is one type of ANN. A surrogate model based on BP neural network is developed to quickly predict two key neutronics parameters in reactor core: the effective multiplication factor (keff) and the three-dimensional channel-by-channel neutron flux distribution. The dataset samples are generated by modeling and simulating different operation states of the Molten Salt Reactor Experiment (MSRE) using the Monte Carlo code. Hyper-parameters optimization is performed to obtain the optimal surrogate model. The numerical results on the test dataset show good agreement between the surrogate model and the Monte Carlo code. Additionally, the surrogate model significantly reduces computation time compared to the Monte Carlo code and greatly enhances efficiency. The feasibility and advantages of the proposed surrogate model is demonstrated, which has important significance for real-time prediction and design optimization of the reactor core.