Neutron Analysis Research Articles

A novel approach for managing the excess reactivity at the beginning of the fuel cycle of VVER-1200

This work discusses various fuel pellet designs loaded with different types and concentrations of burnable absorbers to suppress the excess reactivity at the early stage of the fuel life in the VVER-1200 fuel assembly. MCNPX, V. 2.7 with the CINDER 90 fuel depletion code was utilized to analyze the neutronic characteristics of the proposed cases. Concentric shells (CS) of integral burnable absorbers are investigated in the fuel tube. Gadolinium and Erbium at different concentrations are investigated in the CS at different operation cycles to manage the excess reactivity. Gadolinium and boron as traditional burnable absorbers (BAs) are investigated in the standard fuel assembly to compare their results with the suggested cases. The infinity multiplication factor (kinf), the concentrations of the fuel composition, reactor-grade plutonium (rgPu), minor actinides (MAs) and some important fission products poisoned were studied with burnup. The radial power and thermal neutron flux distribution through one-six of the VVER-1200 assembly have been analyzed to distinguish the optimum BA and design for flattening the power profile. The neutronic analysis indicates that using Er2O3 in a concentric shell of a fuel pellet is an interesting method for reactivity management.

Neutronic analysis of alternative ceramic fuels in the next generation of VVER-1000/V446 reactor

One of the most important issues in nuclear fuel management is burnup increase, which impacts significantly on the reactor cycle length. Burnup enlargement specifically leads to reducing the cost of electricity generation and increasing efficiency. In the present study, the neutronic analysis of some alternative ceramic fuels includingUN, 15UN, and UChas been performed forthe VVER-1000/V446 reactor, comparing with the conventional UO2. Being single-ligand, the alternative fuels’ density of heavy metals such as uranium is greater than UO2’s one. Therefore, the absorption and fissionamounts in thesefuels will be higher than theones in UO2 fuel. By modellingVVER-1000/V446 reactor with MCNPX2.6 Monte Carlo code, we showed that the excess reactivity value for UC fuel (169.173 ∆K/K) increased compared to UO2 fuel (165.526). Due to the ultra-high burnup, the reactor cycle length is 398 days for 15UN fuel, while it is 270, 308, and 130 days for UO2, UC, and UN fuels, respectively, and has increased by 47%. Therefore, the use of alternative fuels has advantages like plant construction cost reduction,excess reactivity, discharge burn up, making them suitable substitution of conventional UO2 for the next generation of LWRs.

Multi-physics coupling analysis of test heat pipe reactor KRUSTY based on MOOSE framework

The heat pipe reactor utilizes heat pipes for passive heat transfer, eliminating the need for any moving parts. This feature significantly enhances its safety characteristics. Given the intricate nature of the heat pipe reactor system, a comprehensive multi-physics coupling simulation is necessary to study its safety features. In this study, a high-fidelity multi-physics coupling analysis incorporating a neutronic-thermal–mechanical-heat pipe coupling model is conducted. The work is based on the multi-physics coupling framework MOOSE, which integrates the Monte Carlo code OpenMC for neutronic analysis and built-in modules for 3-dimensional thermal–mechanical analysis. Furthermore, a heat pipe simulation method called the Thermal Resistance Grid Method is implemented within the MOOSE framework and employed in the multi-physics coupling analysis. To facilitate transient analysis, the Point Kinetics method is also integrated into the MOOSE framework to account for the time-varying power factor during reactor transients. The various modules and physical fields are coupled together using either loose or tight coupling strategies. These coupling strategies are applied to assess a test heat pipe reactor known as KRUSTY. Steady-state simulations are conducted for both normal operation and single heat pipe failure conditions, enabling the calculation of fission power distribution, temperature distribution, stress and strain profiles, as well as power and temperature redistribution following a heat pipe failure. Moreover, a load-following transient of KRUSTY is simulated, demonstrating its capability for load following without reactivity control. The aforementioned calculations validate the feasibility of the coupling framework utilized in this study and highlight its ability to accurately capture local phenomena within the reactor core.

Neutron spectrum optimization for Cf-252 production based on key nuclides analysis

Curium targets are irradiated in high-flux reactors to produce californium-252 (Cf-252). Thermalized neutron spectrum is favorable for Cf-252 production, so traditional methods deploy moderators around the curium targets to optimize the neutron spectrum. However, using moderators alone cannot achieve fine spectrum optimization, and does not necessarily result in better Cf-252 production. We conducted refined neutron spectrum analysis and optimization research to improve production efficiency further. First, based on analysis of the transmutation chain during Cf-252 production, intermediate nuclides with a significant impact on production efficiency, known as “key nuclides,” are identified. Then, the “fission-absorption ratio” and “direct loss rate” of these intermediate nuclides are quantitatively analyzed to determine the energy regions that have a greater negative impact on the production process. Finally, filter materials are selected to reduce the neutron flux in these energy regions, optimizing the neutron spectrum around the curium targets and increasing yield and conversion rate during Cf-252 production. The neutron spectrum optimization method for Cf-252 production based on key nuclides analysis helps improve the efficiency and economy of Cf-252 production.

A comparison of different Fourier transform procedures for analysis of diffraction data from noble gas fluids

A comparison is made between the three principal methods for the analysis of neutron and x-ray diffraction data from noble gas fluids by direct Fourier transform. All three methods (standard Fourier transform, Lorch modification, and Soper–Barney modification) are used to analyze four different sets of diffraction data from noble gas fluids. The results are compared to the findings of a full-scale real-space structure determination, namely, Empirical Potential Structure Refinement. Conclusions are drawn on the relative merits of the three Fourier transform methods, what information can be reliably obtained using each method, and which method is most suitable for the analysis of different kinds of diffraction data. The mathematical validity of the Lorch method is critically analyzed.

Optimized core design and shield analysis of a medium temperature heat pipe cooled reactor

Medium Temperature heat pipe cooled Reactor (MTR) is a new reactor concept proposed by Li et al. in (2022). Low enriched uranium zirconium hydride fuel and mercury heat pipe are introduced in MTR. The unique advantages of MTR have been verified (Li et al., 2022) and it is considered as a potential candidate for energy supply in decentralized markets. In this work, an optimized MTR core design is proposed with detailed neutronic analysis and evaluation on mercury heat pipe, reflector and control rod arrangement, further increasing the economic efficiency and mobility of MTR. Heat transfer limit of the improved heat pipe model is further increased, contributing to a great reduction on core volume. Reflector material and thickness are analyzed in detail for mass minimization. Core physics characteristics are analyzed for four control rod arrangement schemes. The control mechanism is further simplified with only 9 control rods involved in the normal regulation of the core. Core lifetime is far more than 10 years. High inherent safety of the core is ensured. Besides, preliminary design of the shielding system is also carried out. A shieling layer of 60 cm LiH and 20 cm Pb can reduce the total dose to less than 40 % of the design limits. The total mass of the optimized core is greatly reduced and its specific power is increased by 60 % compared with that of the previous design.

Irradiation experiments of titanium beryllide samples in the WWR-K reactor

This paper presents the results primary characterization of unirradiated titanium beryllide samples and neutron analysis of irradiation experiments of titanium beryllide Be12Ti samples in the WWR-K reactor. The XRD analysis of the samples was carried out. The surface morphology of the samples was studied using optical and electron microscopy. Two options of the irradiation capsule design are considered. The first design is designed for low-temperature irradiation, and the second design is for high-temperature irradiation. The effect of the capsule design on the neutron irradiation conditions of titanium beryllide samples is shown. The dynamics of tritium, helium-4 and helium-3 generation in samples is given. The values of the specific energy release of samples and structural materials of the capsule are calculated. The rate of displacements per atom in the irradiated material is estimated.

SUMMER：A small modular lead-bismuth-cooled fast reactor for mobile energy supply

The pursuit of small modular reactors (SMRs) has become a trend in this century due to their potential to expand the use of clean and reliable nuclear energy to a broader range of customers and applications. One promising candidate is the lead-based fast reactor. A mobile energy supply unit based on a lead-bismuth-cooled, pool-type, small modular fast reactor named SUstainable Modular Mobile Enhanced Reactor (SUMMER) is being developed by the University of Science and Technology of China (USTC) and China Nuclear Power Technology Research Institute. SUMMER is intended to provide stable energy supply in the field or remote regions, which are potential application scenarios for SMRs. Key features of SUMMER include a long lifetime of over a decade, mobility (transportable by truck or train), and inherent safety. In this paper, the reactor design is completed under the constraints of fulfilling these features, such as the limitation of size and lifetime. Neutronic and thermal-hydraulic analyses are implemented, as well as safety analyses of the reactor under several typical accidents. The results show that the design meets all these criteria and that the reactor has an adequate safety margin under both steady-state and transient conditions.

Methanol diffusion and dynamics in zeolite H-ZSM-5 probed by quasi-elastic neutron scattering and classical molecular dynamics simulations.

Zeolite ZSM-5 is a key catalyst in commercially relevant processes including the widely studied methanol to hydrocarbon reaction, and molecular diffusion in zeolite pores is known to be a crucial factor in controlling catalytic reactions. Here, we present critical analyses of recent quasi-elastic neutron scattering (QENS) data and complementary molecular dynamics (MD) simulations. The QENS experiments show that the nature of methanol diffusion dynamics in ZSM-5 pores is dependent both on the Si/Al ratio (11, 25, 36, 40 and 140), which determines the Brønsted acid site density of the zeolite, and that the nature of the type of motion observed may vary qualitatively over a relatively small temperature range. At 373 K, on increasing the ratio from 11 to 140, the observed mobile methanol fraction increases and the nature of methanol dynamics changes from rotational (in ZSM-5 with Si/Al of 11) to translational diffusion. The latter is either confined localized diffusion within a pore in zeolites with ratios up to 40 or non-localized, longer-range diffusion in zeolite samples with the ratio of 140. The complementary MD simulations conducted over long time scales (1 ns), which are longer than those measured in the present study by QENS (≈1-440 ps), at 373 K predict the occurrence of long-range translational diffusion of methanol in ZSM-5, independent of the Si/Al ratios (15, 47, 95, 191 and siliceous MFI). The rate of diffusion increases slightly by increasing the ratio from 15 to 95 and thereafter does not depend on zeolite composition. Discrepancies in the observed mobile methanol fraction between the MD simulations (100% methanol mobility in ZSM-5 pores across all Si/Al ratios) and QENS experiments (for example, ≈80% immobile methanol in ZSM-5 with Si/Al of 11) are attributed to the differences in time resolutions of the techniques. This perspective provides comprehensive information on the effect of acid site density on methanol dynamics in ZSM-5 pores and highlights the complementarity of QENS and MD, and their advantages and limitations. This article is part of the theme issue 'Exploring the length scales, timescales and chemistry of challenging materials (Part 2)'.

Core design and neutronic analysis of a long-life LBE-cooled fast reactor NCLFR-Oil

The exploitation of heavy oil reservoirs calls for the implementation of thermal recovery techniques. To provide the heat source for oil recovery energy system, NCLFR-Oil, a Natural Circulation Lead-bismuth Fast Reactor was designed for heavy oil extraction, with a designed power of 40 MW and a lifetime of at least 15 years. In order to achieve the designed object for a reactor core that is durable, compact, transportable, and exhibits strong passive and active safety features, the selection of core materials such as fuel and absorber, evaluation of reactivity control system and corresponding neutronic performance have been optimized by iterative design in this paper with the reference of many small modular reactors. The depletion calculation confirms that the designed reactor can operate for 19 years without refueling. Neutronic performance were analyzed for multiplication factor within lifetime, isotopes conversion, neutron energy spectrum, radial and axial power distributions, nuclear fission reaction rates, reactivity feedback coefficients, and control rod worth with shutdown margins. The core analysis shows that the reactor has good power and safety characteristics, which can fully meet the energy demand of heavy oil exploitation.

Searching for the viability of using thorium-based accident-tolerant fuel for VVER-1200

This study explores the feasibility of employing (U, Th)-based accident tolerant fuels (ATFs), specifically (0.8UO2, 0.2ThO2), (0.8UN, 0.2ThN), and (0.8UC, 0.2ThC). The investigation assesses the overall performance of these proposed fuel materials in comparison to the conventional UO2, focusing on deep neutronic and thermal-hydraulic (Th) analyses. Neutronic analysis utilized the MCNPX code, while COMSOL Multiphysics was employed for thermal-hydraulic analysis. The primary objective of this research is to overcome the limitations associated with traditional UO2 fuel by exploring alternative fuel materials that offer advantages in terms of abundance and potential improvements in performance and safety. Given the limited abundance of UO2, long-term sustainable nuclear energy production faces challenges. From a neutronic standpoint, the U–Th based fuels demonstrated remarkable fuel cycle lengths, except (0.8UN, 0.2ThN), which exhibited the minimum fuel cycle length and, consequently, the lowest fuel burn-up. Regarding thermal-hydraulic performance, (0.8UN, 0.2ThN) exhibited outstanding performance with significant margins against fuel melting compared to the other materials. Overall, when considering the integrated performance, the most favourable results were obtained with the use of the (0.8UC, 0.2ThC) fuel configurations. This study contributes valuable insights into the potential benefits of (U, Th)-based ATFs as a promising avenue for enhanced nuclear fuel performance.

Configurational and Dynamical Heterogeneity in Superionic Li5.3PS4.3Cl1.7−xBrx

AbstractThe correlation between lattice chemistry and cation migration in high‐entropy Li+ conductors is not fully understood due to challenges in characterizing anion disorder. To address this issue, argyrodite family of Li+ conductors, which enables structural engineering of the anion lattice, is investigated. Specifically, new argyrodites, Li5.3PS4.3Cl1.7−xBrx (0 ≤ x ≤ 1.7), with varying anion entropy are synthesized and X‐ray diffraction, neutron scattering, and multinuclear high‐resolution solid‐state nuclear magnetic resonance (NMR) are used to determine the resulting structures. Ion and lattice dynamics are determined using variable‐temperature multinuclear NMR relaxometry and maximum entropy method analysis of neutron scattering, aided by constrained ab initio molecular dynamics calculations. 15 atomic configurations of anion arrangements are identified, producing a wide range of local lattice dynamics. High entropy in the lattice structure, composition, and dynamics stabilize otherwise metastable Li‐deficient structures and flatten the energy landscape for cation migration. This resulted in the highest room‐temperature ionic conductivity of 26 mS cm−1 and a low activation energy of 0.155 eV realized in Li5.3PS4.3Cl0.7Br, where anion disorder is maximized. This study sheds light on the complex structure–property relationships of high‐entropy superionic conductors, highlighting the significance of heterogeneity in lattice dynamics.

The applicability study and validation of TULIP code for full energy range spectrum

NECP-SARAX is a neutronics analysis code system for advanced reactor developed by Nuclear Engineering Computational Physics Laboratory of Xi'an Jiaotong University. In past few years, improvements have been implemented in TULIP code which is the cross-section generation module of NECP-SARAX, including the treatment of resonance interface, considering the self-shielding effect in non-resonance energy range, hyperfine group method and nuclear library with thermal scattering law. Previous studies show that NECP-SARAX has high performance in both fast and thermal spectrum system analysis. The accuracy of TULIP code in fast and thermal spectrum system analysis is demonstrated preliminarily. However, a systematic verification and validation is still necessary. In order to validate the applicability of TULIP code for full energy range, 147 fast spectrum critical experiment benchmarks and 170 thermal spectrum critical experiment benchmarks were selected from ICSBEP and used for analysis. The keff bias between TULIP code and reference value is less than 300 pcm for all fast spectrum benchmarks. And that bias keeps within 200 pcm for thermal spectrum benchmarks with neutron-moderating materials such as polyethylene, beryllium oxide, etc. The numerical results indicate that TULIP code has good performance for the analysis of fast and thermal spectrum system.

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.

Effects of Neutron Source Models on Neutron Wall Loading in the EAST Tokamak

As a significant input for neutronics analysis of tokamak devices, the plasma neutron source is a bridge connecting fusion plasma physics and engineering design. A plasma neutron source model is established based on the EAST plasma configuration. The maximum neutron wall loading (NWL) is near the outboard midplane and is more than ~20% that of the inboard midplane. The investigations demonstrate that special care should be taken for radiation shielding and protection of the key components located on the outboard midplane and on the inboard midplane. The effect of the tritium accumulation on neutronics analysis should be carefully considered for EAST tokamak D-D plasma operations. The effect of density peaking (DP) on the neutronics analysis is also investigated. It is evident that the peak NWLs are all near the outboard midplane and that the poloidal distributions of the NWL are slightly different for these cases. With increasing DP, both the outboard and inboard peak NWLs decrease. However, the decrease in NWL is very small; NWL decreases only 11% when the neutron source peak increases about 1.5 times.

Parametric neutronic analysis of different cladding options for ThO2 pellet of advanced dual-cooled annular PWR assembly

This study presents a parametric neutronic evaluation of different candidate claddings for a purposed candidate fuel (Th-233U-235U)O2 in an advanced dual-cooled annular PWR assembly. The investigated claddings include Zircaloy-2/4, zirconium carbide (ZrC), silicon carbide (SiC), ferritic Fe-20Cr-5Al (FeCrAl) and APMT alloys, and austenitic 310 (SS-310), compared to the by design cladding for the AP1000 assembly, Zirlo™. Neutronic calculations were performed using the deterministic code DRAGON, and the parameters included infinite multiplication factor, criticality periods, neutron flux spectra, radial power distributions, rim effects, and reactivity feedback coefficients.The results show that the candidate fuel with Zirlo™ achieves a higher criticality period than UO2 fuel in both solid and annular configurations. The candidate fuel with SS-310 cladding exhibits a higher criticality period penalty of about 12.60% compared to Zirlo™, APMT, and FeCrAl claddings show a criticality period penalty of about 10.94% and 8.79%, respectively. The candidate fuel with ceramic claddings, SiC and ZrC, can achieve slightly higher criticality periods compared to Zirlo™ of about 1.11% and 0.24%, respectively, due to their low thermal neutron absorption cross-sections. Regarding the power profile, the annular (Th-233U-235U)O2 fuel exhibits a flatter behavior compared to UO2 fuel. However, the claddings with lower capture cross-sections exhibit slightly increased fission power through the fuel pellet. The fuel temperature coefficient for candidate fuel exhibits more negative values as the capture cross-section of the cladding increases, while the molybdenum and zirconium-containing claddings exhibit more favorable moderator temperature coefficients.

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.

Neutronic Analysis of the RSG-GAS Fuel Using Burnable Poison

Control and safety of nuclear reactors are significantly influenced by the determination of safety parameters. The three most crucial safety factors for assessing reactor status are the infinite multiplication factor, reactivity coefficients, and power peaking factor. The objective of the present study is to examine how the RSG-GAS fuel safety parameters behave in a typical reactor operation state. A lattice cell fuel model of the fuel lattice of the RSG-GAS reactor core was modeled using WIMSD-5Bwith cross-section library data based on ENDF/B-VIII.0. The value of the infinite multiplication factor with various burnable poison concentrations, as well as the moderator and fuel temperatures, were the variables that were examined. The reactivity coefficient parameters were similarly analyzed. By comparing the WIMSD-5B code results with information from the SAR document, the WIMS model for RSG-GAS fuel was verified, and it was inferred that the parameters are in good agreement. Safe behavior uses the predicted reactivity coefficient values as an example.

Neutronic and thermal analysis of heat pipe cooled graphite moderated micro reactor

Heat pipe micro reactors are drawing a lot of interest lately due to their distinct features, including the elimination of moving parts required for transferring the heat, as they do so passively through heat pipes, combined with the ability of such reactors to provide energy to remote locations. This present study introduces a proposed design for heat pipe cooled graphite moderated micro reactor suitable for remote locations and space applications. The main requirements of the design include long core lifetime, two shutdown mechanisms, uniform power distribution, and strong negative temperature coefficient. Neutronic calculations were performed using MCNP 6.2 code. Criticality analysis demonstrated the ability of both control devices incorporated in the design to independently control and shutdown the reactor. Calculations of the flux and power distribution showed that, for a thermal power of 0.25 MW, the maximum flux at mid-plane of the reactor is in the order of 1012n/(cm2·s), while radial and axial power peaking factors were found to be 1.2 and 1.13, respectively. The results also showed that the reactor has an overall negative temperature coefficient. Additionally, burnup calculations were carried out to calculate the core lifetime which was estimated to be more than 70 years. A preliminary steady-state thermal analysis was conducted for the reactor using ANSYS FLUENT, assuming 1/12th symmetry. The obtained temperature distribution of the reactor showed that the operating temperatures are within the accepted limits where the highest temperature reached in the fuel was 1074 K. Furthermore, soft coupling of MCNP and ANSYS FLUENT was conducted through an iterative process of performing neutronic and thermal simulation using updated results for the temperatures and power distributions in each iteration.

Neutronic analysis of ABR-1000 sodium cooled fast reactor with radially and axially heterogeneous core configurations

This study presents two proposed models for an ABR-1000 SFR metallic fuel core utilizing different combinations of Thorium, Uranium, and Plutonium fuels. The axial and radial models were evaluated using the OpenMC, open-source Monte Carlo simulation tool. The results of the depletion analysis revealed that ThUPu and UPu models demonstrated a higher cycle length than the ThU model, with the ThUPu(A) model being the most favorable option regarding safety and environmental sustainability. Furthermore, only ThUPu and UPu exhibit fissile inventory increment as time progresses. The simulation of the axial and radial models also demonstrated a higher neutron flux for all axial models and negative reactivity coefficients for all models. UPu(A) exhibits the highest beta effective value, followed by ThUPu(A) then by ThU(A). It is concluded that the axial model is the most favorable option, with the ThUPu(A) model being the optimal choice among the six proposed models.