Neutron Flux Spectrum Research Articles

Evaluation of thermal and epithermal neutron flux using the ko­NAA method with Gold and Zirconium monitors

Neutron flux analysis plays a crucial role in various nuclear research studies and sample analysis in research reactors. Accurate knowledge of the neutron flux spectrum and the ratio of thermal to epithermal neutrons is essential in these applications. In recent years, the newly k₀ method has been utilized to determine thermal and epithermal neutron flux. This method involves the irradiation of two monitors, typically gold and zirconium, followed by the counting of emitted gamma rays using suitable detectors like HpGe. By analyzing the photonic peak area of the collected gamma-ray spectrum from the monitors, the thermal and epithermal neutron flux at the irradiation site can be calculated. In this study, the k₀ method was employed to determine the thermal and epithermal neutron flux of Miniature Neutron Source Reactors (MNSRs).The method involved the irradiation of gold and zirconium monitors, followed by the measurement of emitted gamma rays. The photonic peak area of the gamma-ray spectrum obtained from the monitors was utilized to calculate the thermal and epithermal neutron flux. The research achieved a calculated error of 4 percent in determining the neutron flux. The k₀ method, using gold and zirconium monitors, proved to be an effective approach for determining the thermal and epithermal neutron flux in the Isfahan's MNSR. The use of zirconium, with its suitable neutron absorption cross-sections, along with cadmium calculations, contributed to the accuracy of the flux determination. It is recommended to apply the k₀ method in neutron and boron therapies and perform simulations using popular computational codes alongside the dual metal (gold and zirconium) measurement. This would enhance the accuracy of dosimetry calculations for absorbed doses resulting from neutron therapy.

Comparison of Neutronic Analysis for 250 MWth Molten Salt Reactor using Monte Carlo OpenMC Code with Different Nuclear Data Libraries of JENDL 5.0, ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF 3.3

Abstract Nuclear data libraries playian important role in the accuracy of neutronic aspect calculations, which determine various factors in a nuclear reactor design. Likewise, deterministic or Monte Carlo methods significantly affect the calculation results. Thus, the combination of nuclear data library and calculation methods needs to be studied carefully. This research will conduct neutronic analysis for the Molten Salt Reactor (MSR) design, one of the Generation IV reactor types with advanced features for future energy resources, using the Monte Carlo-based program OpenMC and different nuclear data libraries. MSR will be designed for a 250 MWth power capacity, operable for five years. The fuel used will be LiF-BeF2 as a coolant and ThF4-UF4 as the fuel. The compositions for each fuel will be optimized to achieve critical conditions during five years of operation. This research aims to evaluate the use of differentinuclear data libraries on reactor criticality conditions, using the nuclear data libraries ENDF/VIII-B, ENDF/VII.0, JEFF 3.3, and JENDL 5.0. To perform the calculation and analysis, MSR chain depletion modification has to be done for different nuclear data libraries. Then, evaluation will be conducted on effective multiplicationifactor (k-eff), neutron flux spectrum, and cross-section. The results of this research show that the use of JENDL 5.0 provides more optimal critical conditions. This is due to the completeness of JENDL 5.0 nuclear data covering 795 nuclide data and higher fission cross-section values. While ENDF/VIII-B has 557 nuclide data evaluations compared to ENDF/VII.0’s 423 nuclide data, more stable k-eff values are shown for the use of ENDF nuclear data libraries. This research is expected to significantly contribute to meeting Indonesia’s energy needs by providing efficient solutions in MSR design in neutronic aspects using the most accurate nuclear data library.

Ion optical design of the magnetic proton recoil neutron spectrometer for the SPARC tokamak.

A magnetic proton recoil (MPR) neutron spectrometer is being designed for SPARC, a high magnetic field (BT = 12T), compact (R0 = 1.85m, a = 0.57m) tokamak currently under construction in Devens, MA, USA. MPR neutron spectrometers are versatile tools for making high fidelity abinitio calibrated measurements of fusion neutron flux spectra and have been used to infer fusion power, ion temperature, fuel ion ratio, and suprathermal fuel populations at several high performance fusion experiments. The performance of an MPR neutron spectrometer is in large part determined by the design of the magnetic field, which disperses and focuses recoil protons. This article details the ion optical design of a high-resolution MPR neutron spectrometer, including the amelioration of image aberrations due to nonlinear effects. An optimized design is presented that achieves ion optical energy resolution δE/E < 1% and focal plane properties that enable straightforward integration with the hodoscope detector array.

On the Use of a Chloride or Fluoride Salt Fuel System in Advanced Molten Salt Reactors, Part 3; Radiation Damage

Structural materials in fast reactors with harsh radiation environments due to high energy neutrons—compared to thermal reactors—potentially suffer from a higher degree of radiation damage. This radiation damage can change the thermophysical and mechanical properties of materials and, as a result, alter their performance and effective lifetime, in some cases leading to their disintegration. These phenomena can jeopardize the safety of fast reactors and thus need to be investigated. In this study, the effect of radiation damage on the vessels of molten salt fast reactors (MSFR) was evaluated based on two fundamental radiation damage parameters: displacement per atom (dpa) and primary knock-on atom (pka). Following the previous part of this article (Parts 1 and 2), an iMAGINE reactor core design (University of Liverpool, UK—chloride-based salt fuel system) and an EVOL reactor core design (CNRS, Grenoble, France, fluoride-based salt fuel system) with stainless steel and nickel-based alloy material vessels, respectively, were considered as case studies. The SPECTER and SPECTRA-PKA codes and a PTRAC card of MCNPX, integrated with a module which has been developed in MATLAB, named PTRIM and SRIM-2013 (using binary collision approximation), were employed individually to calculate and compare dpa and PKA (this master module containing all three tools has been appended to the iMAGINE-3BIC package for future use during reactor operations). Additionally, SRIM-2013 was applied in a 3D simulation of a radiation damage map on a small sample of vessels based on the calculated PKA. Our results showed a higher degree of radiation damage in the iMAGINE vessel compared to the EVOL one, which could be expected due to the harder neutron flux spectrum of the iMAGINE core compared to EVOL. In addition, the nickel alloy vessel showed better radiation damage resistance against high energy neutrons compared to the stainless steel one, although more investigations are required on thermal neutrons and alloy corrosion mechanisms to determine the best material for use in MSFR vessels.

Development of diamond detector as fast neutron spectroscopy in the EAST tokamak

We report in this paper the development of the single crystal diamond detector as a fast neutron spectroscopy in the EAST tokamak. The diamond detector is used to detect the fast neutron directly without any neutron converter during the deuterium-deuterium fusion experiment, then the neutron energy spectrum is reconstructed from the recorded continuous scattered spectrum by using a deconvolution algorithm. The results indicate the capability of the diamond spectroscopy which can be used directly to monitor the fast neutron flux and energy spectrum in the EAST tokamak.

Neutronic assessment and optimization of ACP-100 reactor core models to achieve unit multiplication and radial power peaking factor

This study focuses on the neutronic and burnup assessment of the PWR-based Chinese ACP-100 SMR core. Four different core models with varying fuel compositions and assembly configurations were analyzed using the Monte Carlo code SERPENT 2. Three of these models shared uniformly enriched fuel assemblies of 3 wt.%, 4 wt.%, and 4.45 wt.% 235U respectively, while the fourth model (the proposed design for the ACP-100 in this study) featured a mix of differently enriched fuel assemblies along with the inclusion of burnable absorbers. The burnup analysis of the cores spanned 1570 EFPDs, evaluating several neutronic parameters (beta effective, reactivity coefficients) as well as fuel cycle characteristics (cycle burnup, discharge burnup and cycle length). Changes in the isotopic concentration of fissile, fertile, and poison materials were observed and the assembly-wise linear power distribution, neutron flux spectra, and radial power peaking factors of the models were determined. In the mixed core, a combination of borated water, IFBAs and control rods were employed to achieve exact criticality (unit keff) and uniform power distribution (unit radial power peaking factor). Insertion of control rods into 41 of the 57 assemblies, in conjunction with IFBAs placed strategically, attained a near unit radial power peaking factor (PPF) of 1.0586. Finally, the combination of IFBAs, sol-bor of 936 ppm concentration, and CRs with rod worth of 414.33 pcm per assembly brought down the keff to 1.000, but increased the PPF to 1.117. While the achievement of uniform power distribution came at the cost of fuel burnup, the proposed mixed core successfully remained critical for over 1.5 years and reached a discharge burnup of 23.37 GWd/t for a two-batch refueling scheme.

Verification of shielding calculation capability of cosRMC with SINBAD fusion benchmarks

In the design of fusion reactors, extensive shielding calculations are necessary to verify the feasibility and safety of the design. The issue of deep penetration is a common and significant problem in fusion reactor shielding calculations. When the thickness of the shielding layer exceeds a certain value, the statistical error in Monte Carlo method calculations significantly increases. Therefore, it is imperative to adopt variance reduction techniques to enhance the computational accuracy for deep penetration problems. cosRMC is a reactor Monte Carlo code, and its accuracy in calculating neutron-deep-penetration shielding problems should be verified. For this purpose, three deep penetration benchmarks from Shielding Integral Benchmark Archive Database (SINBAD) were simulated by cosRMC and neutron cross section data library ENDF/B-VII.0. These benchmarks describe the penetration behaviors of neutrons in models with different geometric shapes and materials. This paper shows detailed cosRMC models based on benchmark experimental models, used to calculate neutron flux spectra at different penetration depths, which are recorded by detectors in the related benchmark experiments. In this way, a direct comparison between cosRMC results and experimental measurements is provided. In order to improve the accuracy of cosRMC in calculating deep penetration problems, the variance reduction technique of cell importance was used. The neutron flux spectrum calculated by cosRMC was compared with the neutron flux spectrum results calculated by MCNP5 and the experimental values of the benchmarks. The comparison results show that the calculations of cosRMC are in compliance with the MCNP5 reference values and experimental values, and the relative deviations between calculations of cosRMC and MCNP5 is within three time of standard errors. Therefore, this paper has verified that cosRMC can applies to neutron deep penetration problems of fusion reactors.

OpenMC-based high-fidelity model for the Moroccan TRIGA Mark II research reactor

The Moroccan TRIGA Mark II serves as a pivotal cornerstone in the advancement of nuclear science and technology within the region. Precise modeling is crucial for ensuring safe and efficient reactor operation. OpenMC, known for its versatility as a stochastic code with a powerful Python API, offers unique capabilities for reactor analysis, optimization, and automation. This makes OpenMC the ideal choice for neutronic analysis in the development of multi-physics models aimed at enhancing the accuracy and reliability of reactor analysis. Leveraging its Python interface, we have developed a high-fidelity model for the Moroccan TRIGA Mark II, enhancing modeling precision. A central element of this study involved making crucial modifications to OpenMC’s source code, enabling the prompt method and calculating the βeff. These modifications are essential, representing a significant advancement in the capacity to conduct a comprehensive analysis of a wide range of critical neutronic parameters. These parameters include the neutron flux spectrum, keff, control rods worth and core excess reactivity, and power peaking factors. The results exhibit strong agreement with the performed experimental measurements and MCNP6.2 calculations.The findings underscore the model’s precision and reliability, making it a valuable asset for neutronic analysis and, particularly, for developing a comprehensive multi-physics model of the Moroccan TRIGA Mark II.

Neutronic analysis of a heat-pipe cooled reactor

Heat-pipe cooled reactors belong to the group of nuclear reactors using heat pipes filled with liquid metals (such as sodium or potassium) to cool the core. Due to the passive heat removal system, there is no need to use closed loops with pumps, and the reactor can be operated with reduced operational requirements. Consequently, this system can be used in remote locations without access to an electrical grid, or it can be used for space applications.This paper deals with a neutronic study of the Special Purpose Reactor Design-B concept from the Idaho National Laboratory. Using Monte Carlo code Serpent 2 and ENDF/B VIII.0 library, a fixed-temperature model was created to calculate the safety characteristics of the system. This included reactivity coefficients, power distribution, neutron flux spectrum, and criticality safety. In a simplified depletion calculation, an effect of fuel depletion on safety systems was determined as well as a decay heat.

Estimation of the buildup of He-3 and Li-6 poisoning, He-4 and tritium activity in beryllium oxide reflector block in APSARA-U reactor

The Apsara-U research reactor, situated at the Bhabha Atomic Research Center, began its operation in 2018. This facility employs a fuel core configuration employing low-enriched uranium and reflector assemblies made of beryllium oxide. In neutron-rich environment of a nuclear reactor, these beryllium oxide reflector blocks participate in nuclear reactions identified as 9Be(n,α) and 9Be(n,2n). The 9Be(n,α) reaction initiates the formation of isotopes such as 4He, 6Li, 3He, and 3H in the process. The accumulation of gasses like 3He, 4He, and 3H causes the beryllium oxide reflector to swell, ultimately restricting how long it can effectively serve as a reflector or moderator material within the reactor. Extended periods of reactor shutdown result in a significant increase in the concentrations of 3He, mainly due to the radioactive decay of 3H. 6Li and 3He nuclei have significant neutron capture cross-sections, leading to a reduction in the reactor's reactivity, a phenomenon termed "poisoning". This effect varies depending on factors like neutron flux, neutron flux spectrum, and the reactor's operational history. Thorough investigations have been carried out to gain insights into the accumulation of 4He, 3He, 6Li, and 3H within beryllium oxide and their effects on reactivity considering the diverse levels of neutron exposure it encounters.

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.

Revolutionizing LWR SMR reactors: exploring the potential of (Th-233U-235U)O2 fuel through a parametric study

ABSTRACT Small modular reactors (SMRs) have garnered significant attention for their operational adaptability and ease of deployment. Thorium, with its well-documented advantages, shows promise as a viable fuel option for SMRs. However, the absence of intrinsic fissile components in thorium necessitates the exploration of different thorium-based fuel combinations. One such combination, (Th-233U)O2 fuel, has limitations due to the presence of pure U-233. To overcome this challenge, a new fuel mixture, (Th-233U-235U)O2, was investigated for SMRs. This study examined the reactor physics characteristics of the (Th-233U-235U)O2 fuel, including fuel burnup, neutron flux spectra, power distribution, the evolution of actinides, and reactivity coefficients. Results indicate that the (Th-233U-235U)O2 fuel allows for a longer criticality period compared to UO2 fuel, with up to a 14% improvement, while accumulating fewer plutonium and transuranic elements. Notably, it demonstrates significantly improved negative reactivity coefficients, particularly for moderator temperature, with an average improvement of 45% over (Th-233U)O2 fuel. The conceptual (Th-233U-235U)O2 fuel, therefore, exhibits promising neutronic properties, presenting possibilities for future studies. These findings contribute to the understanding and advancement of advanced fuel designs for SMRs.

Unfolding the Neutron Flux Spectrum on the Surface of Mars Using the MSL‐RAD and Odyssey‐HEND Data

AbstractUnderstanding the long‐term radiation environment at the surface of Mars allows us to estimate the exposure for future robotic and crewed missions. Typically, the radiation environment includes charged particles (i.e., protons and heavier ions) and neutral particles (i.e., gamma rays and secondary neutrons). Previous studies used in‐situ measurements, models, or both to determine the characteristics of the radiation at Mars. For example, the Mars Science Laboratory instrument, the Radiation Assessment Detector (RAD), has provided invaluable in‐situ data since landing in 2012. However, the RAD instrument is only sensitive to neutrons with energies &gt; ∼6 MeV and therefore misses what is expected to be a substantial flux of lower‐energy neutrons. To address this gap, we have developed an approach to derive the surface neutron spectrum using the MSL RAD data augmented by orbital data from the High Energy Neutron Detector (HEND) onboard Mars Odyssey (neutron energy &lt; ∼10 MeV). Using a power law fit, we determine neutron flux spectra that reproduce the measurements recorded by both RAD and HEND. Our approach involves a series of Monte Carlo simulations to develop a set of atmospheric transmission functions that enables us to convert the on‐orbit HEND data to their corresponding surface neutron flux spectra. The combined RAD—HEND data present a unique opportunity to obtain a complete picture of the surface neutron environment.

Characterization of the Fast-Neutron Irradiator and the Fast-Flux Tube Irradiation Fixtures at the Pennsylvania State Breazeale Reactor

Accurate knowledge of the neutron spectrum at a nuclear research reactor is a prerequisite for planning irradiation experiments, as well as for evaluating irradiation exposure results. The neutron-flux spectrum in the fast-neutron irradiator (FNI) and the fast-flux tube (FFT) irradiation fixtures at the Pennsylvania State Breazeale Reactor (PSBR) were characterized using the multi-foil neutron activation method. These irradiation fixtures make use of graded shielding to produce unique neutron fields. Multiple foil sets were irradiated in the fixtures with different exposure times and reactor powers to understand the stability over a wide range of operating conditions. Measured results were evaluated against a MCNP6 simulation to produce a measurement-informed neutron flux-energy spectrum for each fixture using STAYSL_PNNL. Simulated estimates of the FNI fixture, a newer fixture (∼25 years old), demonstrated excellent agreement with measured results; the FFT did not. Thermal neutron measurements from the FFT suggest there is additional thermal leakage not captured in the simulation model. Possible explanations for the discrepancy include burn-out or degradation (i.e., micro-cracking, gaps, etc.) in the boral and cadmium liners over the lifetime of the fixture (∼40 years old).

Depletion chain simplification using pseudo-nuclides

This paper introduces a method for simplifying the depletion chains and reducing the number of nuclides tracked in a given nuclear system while adding pseudo-nuclides in order to preserve decay heat. Pseudo-nuclides are demonstrated to preserve the decay heat effects of fission product irradiation with fewer nuclides tracked overall than decay heat precursors. The functionality is implemented and tested in the reactor physics application Griffin. Extensive testing is performed using the comprehensive ENDF-B/VIII.0 decay data and microscopic cross-section libraries from ORIGEN and MC2-3. The impact on the number of nuclides that must be tracked based on effects such as changing cross-section data, changing energy group structure, changing neutron flux spectrum, and tracking decay energy release is explored in this work. Overall, the addition of pseudo-nuclides allows the number of nuclides tracked to be reduced from 1,700 to 400 or fewer nuclides, based on the criteria used in this work.

Preliminary Study on Alternative Reflectors for LOTUS Reactor

LOTUS reactor - a compact lead-cooled fast reactor is currently being studied to generate 200 MWth of capacity and operate for 20 years without refueling for the floating nuclear power plant (FNPP) application. Therefore, it would be a big advantage to achieve a long operation without refueling. However, a small reactor usually has a higher neutron leakage, and a good neutron reflector is essential to maintain the neutron economy. The main objective is selecting potential reflectors materials for the LOTUS reactor using Monte Carlo code, Serpent. The various candidate reflector materials, including Al2O3, BeO, MgO, PbO, SiO2, and ZrO2 are calculated from the neutronics characteristics to determine a good neutron reflector. In this work, we have investigated the parameters of neutronics characteristics, such as core neutron flux spectrum, evolution of keff due to burn-up, power distribution, and lead coolant void reactivity with each reflector. From the comparison of those parameters, MgO material was found to be a good candidate for the relector of LOTUS reactor.

Verification and Analysis of the Problem-Dependent Multigroup Macroscopic Cross-Sections for Shielding Calculations

Multigroup constants are the foundation of neutron and photon transport problems, and the accuracy of multigroup cross-sections has a significant impact on shielding calculation. Challenges have arisen in generating accurate multigroup macroscopic cross-sections for some problems using the widely used cross-section processing code TRANSX 2.15. We developed the multigroup neutron-photon macroscopic cross-section processing module in the shielding transport code ARES. The module is capable of handling the neutron-photon coupled master libraries in MATXS format and providing the problem-dependent multigroup macroscopic cross-sections tailored to each specific shielding physics problem. The self-designed problems with a single nuclide, as well as the iron and OKTAVIAN experiments, are used to verify and analyze the accuracy of neutron and photon macroscopic cross-sections. Results indicate that the macroscopic cross-sections, neutron flux and neutron-photon leakage spectrum obtained by the MGMXS module are consistent with corresponding reference values. As for the JANUS Phase I fixed source shielding benchmark, the relative differences in the reaction rate between the calculation results and experimental data are within 20%. The module provides better problem-dependent multigroup macroscopic cross-sections for neutron and photon shielding calculations that satisfy the accuracy requirements of engineering applications.

Computationally Optimized Irradiation Chamber Design for Production of 135Xe in the Washington State University TRIGA Reactor

This work summarizes the radiation transport–based design for a new D2O-moderated ex-core irradiation facility in the Washington State University (WSU) TRIGA reactor for optimization of 135Xe sources used for calibration and quality control testing of Xe gas detection equipment in support of the Comprehensive Test Ban Treaty (CTBT). Three-dimensional (3-D) particle transport analysis characterizing the WSU reactor core using MCNP6.2 (3-D Monte Carlo) and PENTRAN (3-D deterministic parallel SN) form the basis for the computational optimization. Excellent agreement between MCNP6.2 and PENTRAN predictions is observed. A fundamental fuel bundle depletion analysis is applied to enable a more accurate prediction of neutron flux and neutron spectrum distribution, which drives production rates of 135Xe and 133Xe. The results of various model simulations were used to inform recommendations for the final irradiation chamber design, which has been optimized for safe placement in the reactor tank prior to startup and will allow for insertion and rotation of xenon “bean” samples using existing WSU irradiation equipment, while remaining within operational parameters. The irradiation chamber is expected to produce samples that will remain viable for use in CTBT standards applications for durations 70% to 80% longer than samples produced using current procedures. Thus, this design is expected to improve CTBT-related calibrations and performance testing and to support the continued stability of the CTBT monitoring network.

Flux and Nuclear Heating Rate Calculations for CALORRE Irradiation in MITR

This work summarizes neutronic calculations performed for the CALORRE differential calorimeter specifically designed by Aix Marseille University to inform the irradiation campaign planning for testing in MITR within the framework of the CALOR-I research program. MCNP software was used to predict the neutron and gamma flux spectrum, and total nuclear heating rate to the components of the calorimeter in a variety of positions. Results were compared to evaluate spatial bias, core loading effects, optimization for axial position, and provide input data for thermal multi-physics modelling.

Characterisation of neutron flux spectra with PGNAA shutter: a computational approach

Characterisation of neutron flux spectra with PGNAA shutter: a computational approach