Thermal Neutron Spectrum Research Articles

Analysis of zirconium hydride moderator effect on the micro lead-based reactor

As a crucial parameter for the design of micro reactors, the neutron spectrum directly impacts the keff, the critical size, burnup characteristics, reactivity temperature coefficients and so on. When considering the optimization of the reactor size, designs utilizing thermal neutron spectrum and fast neutron spectrum each present their unique advantages and challenges. To investigate the impact of energy spectrum on the reactor performance, a micro lead-based reactor with annular channel fuel elements is proposed and the study of the influence of varying the volume ratio of moderator and fuel (the M/F ratio) on the keff, the critical size, burnup characteristic and reactivity temperature coefficients by using the Reactor Monte Carlo code (RMC code) is conducted. The results show that for the micro lead-based reactor proposed, when the reactor energy output demand is greater than 350 MWt·years, the design with fast neutron spectrum (without moderator) exhibits the minimum size. While the reactor energy output demand is less than 350 MWt·years, the design with a thermal neutron spectrum (with moderator) achieves the minimum size. Moreover, to ensure the negative reactivity temperature effect, the M/F ratio should be maintained below 3.4. The insights presented in this paper will serve as valuable references for the design of the micro reactor.

Uncertainty Quantification of 237Np, 241Am, and 243Am Reaction Rates in Highly Enriched Uranium Fuel Cores at Kyoto University Critical Assembly

Experimental analyses of 237Np, 241Am, and 243Am fission, as well as 237Np capture reaction rates, are conducted with the Serpent 2 code together with ENDF/B-VIII.0 and JENDL-5 using experimental data for the neutron spectra of thermal and intermediate regions obtained in the solid-moderated and solid-reflected cores with highly enriched uranium fuel at the Kyoto University Critical Assembly. Also, uncertainty quantification of the fission and capture reaction rate ratios of the test samples of 237Np, 241Am, and 243Am with reference samples of 235U and 197Au are evaluated by the MARBLE code system. In terms of the fission reaction rate ratios of 237Np/235U, 241Am/235U, and 243Am/235U, a comparison between experiments and Serpent 2 calculations shows an accuracy of about 5%, 15%, and 10%, respectively, together with ENDF/B-VIII.0 and JENDL-5. For the capture reaction rate ratios of 237Np/197Au, Serpent 2 calculations reveal a fairly good accuracy at the thermal neutron spectrum. The total uncertainties of the 237Np/235U, 241Am/235U, and 243Am/235U fission reaction rate ratios by MARBLE with the covariance data of ENDF/B-VIII.0 and JENDL-5 are found to be about 4% at most in all cores, except for about 8% for 243Am/235U with ENDF/B-VIII.0 at the intermediate neutron spectrum.

Optimizing irradiation conditions for natural molybdenum in WWR-K reactor

The production of the radioisotope molybdenum-99 in the WWR-K research reactor is achieved through the activation method 98Mo(n,γ)99Mo, utilizing a target of natural molybdenum trioxide irradiated under standard conditions (thermal neutron spectra and water environment). Under such conditions, the maximum specific activity of molybdenum-99 reaches (2.3 ± 0.3) Ci/g Mo after 7 d of irradiation. However, the escalating demand for molybdenum-99 and the need to reduce its production cost, necessitates urgent and increased productivity. This study aims to optimize the irradiation conditions for molybdenum powder in the WWR-K reactor to increase the specific activity of molybdenum-99. For this purpose, we evaluated various irradiation capsule designs comprised various neutron moderator materials and thicknesses. Through extensive modeling calculations, we obtained an optimal capsule design that increases the specific activity of molybdenum-99 to 3.31 Ci per 1 g of Mo.

Analyze the Impact of Nuclear Energy on Many Aspects of Life Cycles—Review

AbstractEnergy has long been seen as an important factor in meeting needs of people, in order to sustain and support the economic growth and livelihood. To understand the various technological processes in the nuclear power life cycle, it is first necessary to know the types of fuels used today. There are two types of gasoline in use today, “open” and “partially closed,” respectively, referred to as “single cycle” (OTC) and “double cycle” (TTC). A distinction must be made between “partial” closed fuel cycles, as they apply to current critical (thermal neutron spectrum) reactor technology, which is limited to two passes, and “full” closed cycles, which allow for greater circulation of spent oil but must be used. Fast reactors (FR) are included in the reactor fleet. Many recoveries of FR—already proven but not yet widely used in industry—will lead to reduced demand for uranium mining. In this section, references to “closed” fuel cycles include “partially” closed cycles, i.e., TTCs. What type, if any, is the correct content or information in question? Analysis of the world's energy consumption in recent years shows that the world's energy consumption has increased by 50% in less than a decade.

Effective Multiplication Factor Analysis of Gas-cooled Reactor using OpenMC code with ENDF/B-VII.1, ENDF/B-VIII.0 and JENDL-5.0 Nuclear Data

As the demand for nuclear energy keeps rising, Generation IV reactor designs continuously improve. One such potential reactor design is the Gas-cooled Reactor. Neutronic analysis is the foundation for ongoing design modifications and alternatives for the Helium Gas-cooled Reactor. An important aspect of neutronic analysis is the nuclear data library. This research compares how different nuclear data libraries impact a Helium Gas-cooled Reactor design by comparing the neutron effective multiplication factor (k-eff) value. This research was conducted by the Monte Carlo method using OpenMC code. ENDF/B-VII.1, ENDF/B-VIII.0 and JENDL-5.0 were used as nuclear data libraries to simulate the reactor design for ten years of depletion time with 1-year timesteps. The reactor is designed in two different spectrums: the fast neutron spectrum design and the thermal neutron spectrum design. The differences between the two designs are in the cladding and reflector materials where the thermal neutron spectrum uses graphite moderator. The results for thermal spectrum design show that while each nuclear data library’s average k-eff values vary, they all produce the same results at timestep zero, with k-eff values around 1.02500. The average k-eff value from simulation using the ENDF/B-VII.1 library is 1.01477, the average k-eff value from simulation using the ENDF/B-VIII.0 library is 1.01122, and the average k-eff value from simulation using the JENDL-5.0 library is 1.01047. Each simulation result has an average uncertainty of 0.00040. The difference is due to the different amounts of data in each library. JENDL-5.0 has the highest amount of data inside its nuclear libraries, followed by ENDF/B-VIII.0 and ENDF/B-VII.1. Meanwhile, the average difference between the different libraries is insignificant for the thermal gas-cooled reactor simulation results. ENDF/B-VIII.0 produces an effective k-average of 1.04437, and JENDL-5.0 produces an effective k-average of 1.04273. Mass change over time results also shows that the transmutation of Uranium-238 into Plutonium-239 occurred. The fast spectrum reactor design shows a greater increase in plutonium-239 than the thermal spectrum design. The three libraries used did not show large differences in results in calculating changes in the mass of Uranium-238 over time due to Uranium-238‘s large mass, but they still showed differences in Plutonium-239‘s mass change over time. The average value difference between libraries in the fast neutron spectrum reactor design is 50 kg, and in the thermal spectrum it is 30 kg.

Safeguards and Security for High-Burnup TRISO Pebble Bed Spent Fuel and Reactors

Several high-temperature thermal neutron–spectrum pebble bed reactors are being commercialized. China has started up two helium-cooled pebble bed high-temperature reactors. In the United States, the X-Energy helium-cooled and the Kairos Power salt-cooled pebble bed high-temperature reactors will produce spent nuclear fuel (SNF) with burnups exceeding 150 000 MWd per tonne. The reactor fuel in each case consists of small spherical graphite pebbles (4 to 6 cm in diameter) containing thousands of small TRISO (microspheric tri-structural isotropic) fuel particles embedded in the fuel of zone these pebbles. The unique isotopic, chemical, and physical characteristics of this high-burnup SNF create a technical case to eliminate safeguards based on the low risk for use in nuclear weapons, while maintaining safeguards in terms of risk for use in radiological weapons. These safeguards could be reduced to the simple counting and monitoring of pebbles in storage. Alternatively, there is the option to create a special category with reduced requirements for this SNF in storage, transport, and disposal. No safeguards would be required for a repository with only this type of SNF. Reactor safeguards are required for fresh fuel, partly burnt fuel, and to identify unconventional pebbles with depleted uranium or other materials that might be used to create weapons-useable materials.

The impact of fast spectrum graphite moderated systems on thermal spectrum graphite moderated reactors through the resonance physics of n+12C

Understanding the nuclear data used in predictive modeling and simulation of advanced reactors is vital for their development, licensing, and deployment. In this paper, we use the Kairos gFHR and the X-Energy Xe-100 as demonstration examples to test the scientific hypothesis that the resonance physics of n+12C allows for integral experiments with fast-neutron-spectra to be useful in helping to reduce nuclear data uncertainty on k-effective of new reactor systems with thermal-neutron spectra. This work hypothesizes that legacy criticality experiments hold value for these new reactors. HMF071 (highly enriched, metal-fuel, fast-spectrum critical experiment) (Parkey et al., 2014; Nuclear Science Committee of The Nuclear Energy Agency, 2016) is a series of legacy criticality benchmarks with graphite reflectors which were selected to test this scientific hypothesis by leveraging the large correlation visually identified in the covariance matrix of n+12C. This hypothesis will be tested through calculating the similarity and uncertainty-reduction provided from evaluating the effect of a fast-neutron-spectrum system, HFM071, on a thermal-neutron spectrum system. The sensitivity with respect to elastic scattering of carbon had greater than 70% similarity for one of the cases of HMF071 compared against each of the advanced reactor designs considered. This result quantifies the relevance that the fast spectrum experiments can have for thermal spectrum applications due to the resonance physics of 12C. With this connection quantified, the uncertainty reduction from the fast-spectrum experiments on these thermal-spectrum reactors was calculated using the SCALE module TSURFER (Rearden and Jessee, 2018; Wieselquist and Lefebvre, 2023). The HMF071 series of critical experiments were able to lower the nuclear data uncertainty associated with carbon on the Xe-100 and gFHR only by about 6 pcm using SCALE 6.2.4 and the ENDF/B-VII.1 nuclear data library. In addition, the HMF071 series of critical experiments were able to lower the nuclear data uncertainty associated with carbon on the Xe-100 and gFHR by about 11 pcm and 16 pcm respectively using SCALE 6.3.1 and the ENDF/B-VIII.0 nuclear data library. When all-materials were used in the TSURFER evaluation, the uncertainty reduction on the gFHR was about 39 pcm using SCALE 6.3.1 and the ENDF/B-VIII.0 nuclear data library. This is very small; however, due to the fact that carbon is such a well-studied element, any further reduction of uncertainty for thermal-spectrum systems is difficult (but useful) to obtain. These null results disproved our hypothesis and do not allow us to demonstrate this specific use case for evaluating legacy fast-neutron-spectrum experiments, which are otherwise often neglected, for modeling modern thermal-neutron spectrum graphite moderated/reflected reactors.

Turbulent induced vibration of a guide tube in experimental reactor

The design of advanced experimental nuclear reactors consists in integrating safety and operational requirements as well as reaching targets in terms of thermal power and neutron spectrum. In order to meet theses constraints, slender structures with little supports and crossing the entire reactor vessel are implemented in the reactor and are subjected to an axial flow that generates flow-induced vibration (FIV). From the industrial point of view, the mastering of the occurrence of FIV and its associated wear in case of contacts is requested. The stationary fluid forces applying on slender tubular structure in response to its motion are of primary interest since they can significantly affect the vibration amplitudes and even the stability of the system, especially in case of confined axial flow with high velocities (typically higher than 10 m/s). In this study it proposed to compare two approaches to estimate the vibration induced by turbulent excitation of an industrial device encountered in a research nuclear reactor. The control-rod guide-tube mock-up of the Jules Horowitz Reactor, previously tested in an hydraulic channel at the Technical Center from Le Creusot in France, is retained for this benchmark. Two models are proposed, one based on derivation of leakage flow theory and the other one was based on potential flow theory with adjusted coefficients given by CFD simulations. The flow induced vibration amplitude is consistent with the experimental data. Also, the calculation and experiment provide similar trends when the boundary conditions are changed.

Measurement of leakage neutron flux from reactor tank surface of AHWR-critical facility

AHWR - Critical Facility (AHWR - CF) is a “zero power” reactor designed to carry out various reactor physics experiments for validation of AHWR physics design. AHWR-CF is a vertical tank type reactor. This paper describes the experiments carried out for neutron flux measurements on the reactor tank surface of AHWR-CF. A number of bare Au and cadmium covered Ni foils were used to measure thermal and fast neutron flux at the outer surface of the reactor tank. One cadmium covered Au foil was also irradiated at one of the locations on the tank surface, to estimate the epithermal component of neutron flux. Westcott thermal flux on the tank surface (at an elevation of 120 cm from tank bottom) was found to be (5.48 ± 0.57)E+6 n cm−2 s−1 on one of the sides of the reactor tank. Cadmium ratio for Au for this location was 40.28, indicating a highly thermalized neutron spectrum. No statistically significant activity was found in the irradiated Ni foils, indicating the absence of fast neutron flux above the measurable threshold. Low value of flux on the reactor tank surface implies that any irradiation damage in the AHWR-CF reactor tank material will be insignificant over its period of operation.

Peculiarities of Changes in the Isotopic Composition of Pilot Fuel Elements of a VVER-SKD Reactor under Successive Irradiation in a Fast and Thermal Neutron Spectrum

Peculiarities of Changes in the Isotopic Composition of Pilot Fuel Elements of a VVER-SKD Reactor under Successive Irradiation in a Fast and Thermal Neutron Spectrum

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.

OpenMC interpretation and analysis of two series highly-enriched-uranium thermal-spectrum benchmarks for nuclear data validation

This study aims to provide additional validation of nuclear data performance through a comparative analysis of the evaluated nuclear libraries ENDF/B-VII.1, ENDF/B-VIII.0, JEFF-3.3, and JEFF-4 T1. The study utilizes the OpenMC Monte Carlo neutron transport code and integral benchmarks provided by the International Criticality Safety Benchmark Evaluation Project (ICSBEP) which have not been interpreted since their first modeling. The criticality benchmarks assemblies used in this study cover 2 series which are fueled with Highly Enriched Uranium (HEU) in Metal form (MET) and have Thermal neutron spectra (THERM), referred to as HMT-026 and HMT-027, respectively. The results show that for the HMT-026 benchmarks, ENDF/B-VII.1 and ENDF/B-VIII.0 have results close to experiments. The best global agreement with the HMT-027 experiments was found to be ENDF/B-VII.1. Through relevant critical and nuclear data sensitivity analysis, it is found that Be and BeO are respectively of crucial importance for HMT-026 and HMT-027 and the nuclear data of 9Be (especially for ENDF/B-VIII.0) should be reviewed. Additionally, it is worth noting that the Thermal Scattering Law (TSL) of BeO has a great influence on the keff calculation for the HMT-027 benchmarks, while the influence of TSL of metal Be is negligible for the HMT-026 experimental benchmarks.

The acceleration of improved Tone's method in advanced reactor lattice neutron spectrum calculation

This paper focuses on the acceleration of solving fixed-source equations in improved Tone's method. Since each equation is independent of the other with respect to the energy variable, the parallel calculation based on energy group domain decomposition was used in the advanced reactor lattice neutron spectrum calculation code TULIP. In order to balance the load of each CPU, the equally-divided strategy and a weighting-based strategy, which were used for the division of energy groups, were proposed and discussed. In the weighting-based strategy, the cross sections were used to calculate the weighting so that the iteration behavior could be caught. Therefore, the CPU load was balanced, and the parallel efficiency was improved. Moreover, the selection of the initial value was also evaluated. To verify the improvement, several tests based on fast and thermal neutron spectrum were calculated. According to the analysis of calculation time and parallel efficiency, using the initial value that was calculated with the homogeneous problem could save more than half of the calculation time. Meanwhile, the weighting-based strategy for parallel calculation could balance the CPU load so that the parallel efficiency was improved about 10%–80%, which is based on different tests.

A study on the feasibility of using Uranium Nitride (UN) fuel in Pressurized water reactors (PWR)

MCNPX computer code is used to design a three-dimensional model for a typical Pressurized Water reactor (PWR) core. Uranium Nitride fuel (UN) was tested for the reactor and the results were compared with a typical UO2 fuel. Neutron spectrum, power distributions, plutonium isotopes productions, and kinetics and safety parameters were evaluated. It was found that thermal neutron spectrum is harder in UN than in UO2. UN can achieve a longer fuel residence time in the reactor than UO2. Kinetic and safety parameters are within the acceptable range similar to UO2. Control rod worth is lower in UN due to thermal flux hardening. This research was carried out with the addition of three types of variations of burnable poison.

A sensitivity analysis to predict the neutronics behavior of samples irradiated in the VTR rabbit system

A low-order neutronics model is developed to carry out hundreds of simulations efficiently and investigate the neutronics behavior of samples being irradiated in a test reactor setting under different geometrical constraints. The low-order model allowed for simulations that yield the expected neutronics behavior of any irradiated sample in any environment and allows for the calculation of highly accurate spatially averaged statistics and idealized spatial distributions in the neutron flux. Several benchmarks are performed to evaluate the performance and limitations of the low-order model revealing many important findings. The low-order model predicted the LHGR in the EBR-II driver fuel to within 2.34% by only simulating the fuel rod by itself, which served as a validation for the model. Sensitivity studies investigated 3% enriched UO2 and U-10Zr being irradiated in the Versatile Test Reactor rabbit system. The analyses investigated a range of combinations of 15 radii and 5 heights for each sample in the rabbit system. Similar data sets are also provided for irradiations in the Advanced Test Reactor’s B-10 irradiation position, which is a thermal neutron spectrum environment. Generalized fits and fit coefficients are obtained for sample heating, reaction rate densities, and local multiplication rate characteristics, allowing the predictions of the neutronics behavior of the samples based on their geometrical constraints. The analyses and fits laid the groundwork for developing a user-end Multiphysics analysis framework to assist and accelerate irradiation experiment design and optimization.

Experimental validation of cold neutron source performance with mesitylene moderator installed at RANS

The RANS (RIKEN Accelerator driven Neutron Source), one of compact accelerator neutron sources (CANS), tries to expand its performance by installing a cold neutron which may provide new opportunities in many applications. RANS is a low power CANS with a proton beam of 7 MeV and 100 µA at maximum. A moderator system was constructed based on results of optimization design study with mesitylene. Recently, we have done performance tests aiming at showing characteristics as cold neutron source. Cryogenic mesitylene moderator was installed on a plug with a new target moderator reflector configuration of RANS. Experiment using a gas electron multiplier (GEM) detector was carried out to measure neutron spectra of the cold moderator. This paper describes performance of the cold moderator in terms of 1) Cold neutron gain of optimization design with respect to a polyethylene moderator, 2) Temperature dependency of cold neutron spectrum flux regarding scattering kernel (SK), and 3) comparison between experiment and calculation. A note is given for comparison between calculations with different SKs available. Also, two-dimensional imaging of cold and thermal neutron spectrum flux on the viewed surface is shown with a pinhole slit configuration.

Nondestructive and destructive assay for forensics characterization of weapons-grade plutonium produced in LEU irradiated in a thermal neutron spectrum

A post-irradiation examination (PIE) of an enriched uranium sample was performed to advance a nuclear forensics methodology, based on intra-elemental nuclide ratios. The PIE was carried out on a few milligrams of low enriched uranium dioxide (LEUO2) sample, irradiated to a burnup of 1 GWd/MTU. The PIE of the LEUO2 sample provided the concentration of selected plutonium and fission product nuclides (239Pu, 240Pu, 241Pu, 91Y, 95Zr, 95Nb, 103Ru, 133Cs, 134Cs, 135Cs, 137Cs, 136Ba, 138Ba, 140Ba, 140La, 141Ce, 144Ce, 149Sm, 150Sm, 152Sm, 153Eu, and 154Eu). The PIE supported validation of the MCNP (Monte Carlo N-Particle neutronics simulation code) predicted concentration of these nuclides in the irradiated LEUO2 sample. This study showed that most of the MCNP predicted nuclide concentrations were within 15% and several within 10% compared to the PIE results. Gamma and mass spectrometry of nuclides in irradiated LEUO2 enabled the determination of uranium burnup and time since irradiation.

The effect of errors in the neutron flux density on the uncertainties of nuclear concentrations of nuclides arising during the calculation of fuel burnup in cells with different neutron spectra

Computational studies have been carried out showing the complex time dependence of uncertainties in nuclear concentrations of various nuclides arising from the propagation of the neutron flux density errors in the burnup calculation process in cells with different neutron spectra on the above errors. It is found that these uncertainties not only depend on the burnup time in a complex way, but also depend on the spectrum of the cell. The variants of the cell with thermal and fast neutron spectra were considered. The calculations were performed using the VisualBurnOut program (Kolesov et al. 2009), which makes it possible to estimate these uncertainties arising due to errors in the input parameters of the burnup problem (reaction rates, neutron flux density, etc.). The influence of the number of calculated burnup points on the results of burnup calculations by the Monte Carlo method was investigated. Uncertainties arising in nuclear concentrations at intermediate calculation steps due to errors in nuclear concentrations appearing at the previous step were taken into account in the calculations.

Innovative Compact Molten Salt Reactor (ICMSR) Analysis for Mo-99 Production

Mo-99 isotope production calculation in the ICMSR (Innovative Compact Molten Salt Reactor) with the computer code MCNP6 has been carried out. ICMSR is a conceptual design of the MSR type reactor that uses NaF-ThF4-UF4 fuel with an enrichment of 235U of 19.75%. This reactor operates in thermal neutron spectrum with a graphite moderator. ICMSR is a power reactor that produces Mo-99 as a by-product. Calculations carried out for 12 days of operation show that the reactor condition is still critical so that there will be no intervention from refueling. The total Mo-99 produced until the 12th day is 9.118 x 106 Ci. Mo-99 can be extracted from the reactor as long as the power reactor is operating so it will be economically advantageous.Keywords: 99Mo, isotope production, ICMSR, MCNP, criticality

Energetics of oxidation and formation of uranium mononitride

Uranium mononitride (UN) is an advanced nuclear fuel currently being considered for use in several generation IV fast and thermal neutron spectrum core designs, with additional applications to thermal and electric nuclear propulsion reactors. To better understand the thermal behavior and thermodynamic stability of UN, we investigated the bulk thermal oxidation process and thermochemical reactions, including the enthalpy of oxidation and standard enthalpy of formation, by conducting thermalgravimetric analysis – differential scanning calorimetry coupled with mass spectrometry (TGA-DSC-MS), and high temperature transposed temperature drop and oxide melt drop solution calorimetry. The bulk oxidation of UN (containing a small amount of α-UN1.5+x) in air was found to follow a step-wise process characterized by consecutive oxidative reactions UN-UN1.5+x-UO2 → UO2-UO3Nk → UO3Nk → UO3 → U3O8. TGA results support that the UO2 – U2N3+x passivating layer delays the onset of rapid bulk oxidation of UN in air up to 662 K. Synchrotron X-ray diffraction (XRD) and extended X-ray absorption fine structure (EXAFS) analyses were performed to characterize UN and its final oxidized product. The standard enthalpy of formation (ΔH°f) of UN was determined to be –144.4 ± 5.9 kJ/mol·atom, in good agreement with previously determined values from Pt encapsulation and bomb calorimetric experiments. Lastly, a negative linear correlation between ΔH°f and the N/U molar ratio was established based on the thermochemical data obtained in this work and previously reported enthalpies of formations of β-UN1.5-x and α-UN1.5+x.