Neutron Flux Energy Research Articles

A Monte Carlo simulation study for designing collimators for a CZT-based spent nuclear fuel characterization system

As an attempt to develop a spent nuclear fuel characterization system that can acquire both high-resolution gamma-ray energy spectrum and neutron flux simultaneously at multiple positions around the spent nuclear fuel assembly and provide more information compared to conventional equipment, simulation study was conducted to design a collimator around a gamma-ray detector to obtain high-resolution gamma-ray spectra in a high-flux gamma-ray environment. The ORIGEN depletion code was used to evaluate the source terms of spent nuclear fuel, and MCNP 6.2, the Monte Carlo simulation code, was used to simulate the detector response. Considering the realistic operating conditions of Korea's pressurized light water reactor, the source term for the ACE7 17×17 nuclear fuel assembly was derived under the conditions of initial enrichment ranging from 3 to 5%, burnup ranging from 30 to 60 GWd/MTU, and maximum cooling time of 30 years. MCNP 6.2, the Monte Carlo simulation code, was used to design a spent nuclear fuel characterization system using a CZT detector as a gamma-ray detector to obtain a high-resolution gamma-ray energy spectrum. In particular, the research focused on designing collimators to be used in high-flux radiation environments, such as spent nuclear fuel inspection, with CZT detectors. A simulation model of the CZT detector to be installed internally was made and the validity of the detector simulation model was confirmed by comparing simulation results with the measurement experiment results. Finally, the response of the CZT detector was calculated by changing various setting parameters of the collimator. By comparing the calculated detector response with the design standard count rate, the design of a collimator that can utilize the CZT detector in a spent nuclear fuel inspection environment was derived.

Characterization of flux distribution in the pool of the RMC SLOWPOKE-2

The neutron flux distribution within the pool of the SLOWPOKE-2 reactor at the RMC has been characterized using neutron activation measurements as well as MCNP simulations. Westcott equivalent thermal neutron flux values were calculated from measured activities of solutions of Co, Au, and Cd-shielded Au at several reactor flux settings and compared to tabulated fluxes from MCNP simulations. Good agreement was found between the simulated and experimental thermal flux values, while larger uncertainties were highlighted in higher energy neutron fluxes.

IFMIF/EVEDA achievements overview

In a fusion reactor, the plasma based on a deuterium-tritium reaction generates high energy neutron flux interacting with the materials of the plasma facing components. In order to study the irradiated behavior of these materials, the International Fusion Materials Irradiation Facility (IFMIF) was conceived to generate these fusion relevant neutrons through d-Li stripping source. IFMIF is presently in its Engineering Validation and Engineering Design Activities (EVEDA) phase under the Broader Approach agreement signed between EURATOM and Japanese Government in 2007. This agreement mandates the IFMIF/EVEDA project to validate the design of the different systems of IFMIF and to produce an integrated engineering design of IFMIF, together with the data necessary for future decisions on its construction and operation. While the Engineering Validation Activity (EVA) of the Lithium Target Facility and the Test Facility was completed by constructing prototypes, the EVA of the Accelerator Facility with the Linear IFMIF Prototype Accelerator (LIPAc) is still on-going at the Rokkasho Fusion Institute, Japan. This paper overviews the achievements in the previous phase ended on 31 March 2020 and the progress in the current phase to complete the commissioning of the LIPAc.

Application of the MELCOR for Fusion Code to the Transient Accident Analysis of the IFMIF-DONES Test Cell

According to the European Fusion Roadmap, to demonstrate the feasibility of fusion energy, it is necessary to build a comprehensive database of materials properties to be used in future fusion power plants. This is the objective of the IFMIF-DONES facility, which will drive a deuteron beam on a liquid lithium target to produce high energy neutron fluxes for irradiating candidate fusion materials. Among the several ongoing activities in the frame of the EUROfusion Early Neutron Source Work Package (WPENS) project, deterministic accident analyses play an important role, since they help identifying a set of reference accident scenarios and related safety class components. Some of these scenarios are being studied with the MELCOR-fusion code, an integrated engineering code which is able to perform thermal-hydraulic transient calculations. In this work, the MELCOR-fusion code has been applied to two potential accident scenarios involving the degradation of the primary lithium loop of IFMIF-DONES. A rupture in the Quench Tank and a break in the inlet nozzle to the Target Vacuum Chamber were postulated as the two initiating events followed by a lithium spill into the Test Cell room. The purpose of this study was to obtain different key metrics, such as the maximum pressure and temperature loads reached in the TC room, the amount of the leaked lithium mass, and the time for lithium solidification. The computed results will help identify the safety requirements to be applied to the final design of the TC room.

On the Employment of a Chloride or Fluoride Salt Fuel System in Advanced Molten Salt Reactors, Part 1: Thermophysical Properties and Core Criticality

Molten salt reactors (MSRs), as one of the six main technologies of Gen IV, can meet the broad area of sustainability, economics, safety and reliability, proliferation resistance and physical protection goals. One of the main and first challenges in designing molten salt fast reactors (MSFRs) is the selection of an appropriate molten salt fuel system based on the envisaged applications and objectives. In this study’s series, a full-scope evaluation has been conducted about employing either chloride or fluoride salt fuels as the main competitors’ candidates for fuel salt in MSFR designs. Two distinguished projects, EVOL (CNRS, Grenoble-France), based on fluoride salt, and iMAGINE (The University of Liverpool, UK), based on chloride salts, were considered in order to achieve this goal as case studies. The first part of this series (part 1—this article) deals with the investigation of the thermophysical properties of the salt fuel system, criticality search and neutron-flux energy spectrum. An identical typical design of the MSFR core has been considered for a neutronic simulation by using MCNPX V2.7 based on the chemical composition of the fuel salt mentioned in both projects. The thermophysical evaluation has been conducted through literature research and theoretical methods based on the experimental values for the salt component properties. The results of the study are presented for thermophysical properties, including the melting point, vapour pressure/boiling point, specific heat capacity, thermal conductivity and density, in addition to neutronic simulation for the core critical dimension and neutron-flux spectrum of both the chloride- and fluoride-based salt fuel systems. In the discussion of the results, it is concluded that both the chloride and fluoride salt fuel systems have benefits that are seen on different comparative parameters. The delivered data will provide future decision makers with evidence for the salt choice for balancing their design objectives with the opportunities and expectations. Thus, a final selection of the most appropriate salt fuel system to be used in MSFRs will be postponed for more investigation in the final part of this article series, combining the data with different potential user profiles.

Long Range Optical Distance Sensors for Liquid Metal Free Surface Detection

For the demonstration of fusion power plant technology, DEMO dedicated materials are necessary to cope with the harsh environment of high energy neutrons. For this purpose, the international neutron irradiation facility for fusion materials IFMIF/DEMO Oriented Neutron Source (DONES) is planned to be built in Granda, Spain. In the DONES facility, a deuteron beam hitting the lithium target produces a high energy neutron flux. Due to the high-power density, the windowless target is a free surface liquid lithium flow in a duct with a concave backplate. In order to keep the heat released by the beam within the liquid lithium and to avoid its intrusion in the backplate, a stable configuration of the free surface flow with a setpoint layer thickness of 25 ± 1 mm is crucial. In particular, stable wave structures, so called wakes, which occur from accumulated impurities at the nozzle edge, can cause a critical local decrease in the layer thickness of more than 1 mm. Therefore, it is necessary to better understand the nature of these wakes and to be able to monitor the surface profile to shut down the beam in case of a critical thickness loss, but to avoid unintended shutdowns. In the context of this work, currently available optical sensors were tested on their capability of detecting a specular liquid metal surface at measurement distances of several meters. After an initial selection, two optical sensors were further considered. Experiments with the liquid metal alloy GaInSn and simulations with the software Blender of the selected optical sensors for their capability of measuring distances of liquid metal were conducted. The results showed a significant dependency of the measurement results on the waviness of the liquid metal surface. Nevertheless, it was possible to resolve the wavy liquid metal surface with a sufficient resolution to detect critical wake structures.

Estimation of the neutron production of KSTAR based on empirical scaling law of the fast ion stored energy and ion density under NBI power and machine size upgrade

Deuterium-tritium reaction is the most promising one in term of the highest nuclear fusion cross-section for the reactor. So it is one of urgent issues to develop materials and components that are simultaneously resistant to high heat flux and high energy neutron flux in realization of the fusion energy. 2.45 MeV neutron production was reported in D-D reaction in KSTAR and regarded as beam-target is the dominant process. The feasibility study of KSTAR to wide area neutron source facility is done in term of D-D and D-T reactions from the empirical scaling law from the mixed fast and thermal stored energy and its projection to cases of heating power upgrade and DT reaction is done.

Regularized heuristic method for activation monitor neutron spectrum unfolding

This paper describes an activation foil measurement based neutron flux unfolding method utilizing a heuristic optimization approach. A regularization parameter based on the differential neutron flux was introduced into the optimization cost function to remove large nonphysical oscillations in the unfolded neutron energy spectrum. The procedure was completed for two example case studies ranging from thermal energies up to 20 MeV demonstrating recovery of neutron flux energy information on radiation environments where a model was partially missing or there was a discrepancy in a reaction channel. The unfolded results provided a distribution of unfolded spectra that conformed to the activation products but indicated degenerate solutions to the general activation problem. Additionally, increasing fidelity of initial estimate on the neutron spectrum reduced the degeneracies and lowered the uncertainty in the unfolded spectrum, even in a low-fidelity, physics-based estimate of the neutron spectrum. Epithermal neutron energies, where few nuclear reactions are available to inform on the neutron flux, were shown to be the most difficult region. This method can be extended to other heuristic optimization routines and offers an alternative option for traditional activation dosimetry neutron flux unfolding techniques while improving the characterization of the neutron flux in applications with poor initial models of the environment.

Neutron pumping of active medium formed by gadolinium isotopes Gd155 and Gd156 pair

The possibility of accumulation of excess energy in the active medium formed by the gadolinium isotopes Gd-155 and Gd-156 possible due to the formation and accumulation of Gd-156m nuclei in the isomeric state during radiation capture of neutrons by the nuclei of a stable isotope with a lower mass is shown. A numerical simulation of the dynamics of the energy levels population of Gd-156 nuclei at repetitively-pulsed pumping of the active medium formed by isotope-modified gadolinium oxide was carried out. The calculation for various combinations of neutron duty cycle of pulses, flux density, and neutron energy spectrum in pulses was performed. It was established that combinations of these parameters, which provide the possibility of inverting the population of the excitation levels of Gd-156 isotope nuclei were largely determined by the energy neutron flux spectrum in pulses.

Neutronic and burn‐up calculations of the ( ThO2‐UO2 ) pin cell benchmark using DRAGON5 and MCNP6 .2 codes with ENDF / B‐VIII .0 nuclear data library

A (ThO2-UO2) fuel pin benchmark, consisting of 75 w/o Th and 25 w/o U, was utilized for analysis of depletion and inventory of the fuel for the pressurized water reactor (PWR) pin-cell model. To perform this study and evaluate cross-section data, the new version VIII.0 of the ENDF/B nuclear data library was utilized to generate a data set in ACE (A Compact ENDF) format at different temperatures (583.1, 600, 621.1, and 900 K), utilizing NJOY2016 and MAKXSF6.2 codes to process the data at the specific temperatures. In this paper, the infinite multiplication factor (kinf), neutron energy flux spectrum (utilizing 172 energy groups), the concentrations and activities of the most important products of fission, the actinide radionuclides accumulated in the pin-cell, and the total radioactivity were calculated and compared, utilizing the MCNP6.2 and DRAGON5 codes. Furthermore, the burn-up (BU) dependent behavior of the PWR thorium pin-cell was validated and compared with the reference results obtained utilizing the CASMO-4, MIT-MOCUP, and INEEL-MOCUP codes. The results of this study show good agreements between all codes analysed.

EVALUATION OF THE DOSE LEVELS IN THE IRRADIATION CHANNELS OF PU-BE NEUTRON SOURCES STORAGE DEVICE BY USING NEUTRON ACTIVATION ANALYSIS AND MCNP6 MODELLING

PuBeneutrons sources of Educational Neutron Laboratory in the Center for Physical Sciences and Technology are used in several activities including neutron activation for educational purposes and laboratory irradiation experiments. In this work we address the dose rate estimation in the irradiation channels of reconstructed PuBe neutron sources storage device in the particular positions dedicated for neutron irradiation experiments using neutron activation analysis (NAA) and MCNP6 modelling.The results on neutron activation of V2O5, MnO2, Al and Na2CO3were analyzed and compared with MCNP6 model prediction results. The conclusions on the actual neutron flux energy distribution and model corrections are drawn. The absolute neutron intensity at dedicated irradiation points as well as neutron and gamma dose rates are obtained.

Verification and Validation of the Geant4 Monte Carlo Code Toolkit for DEMO TBR Evaluations

Verification and Validation analyses of the Monte Carlo code toolkit Geant4 were conducted with nuclear data from the JEFF-3.3 library to show the suitability for fusion applications. We present benchmarking against data of the engineering design oriented HCPB (Helium Cooled Pebble Bed) Tritium Breeder Module (TBM) mock-up experiment taken from the SINBAD compilation and nuclear analyses in comparison with MCNP of the DEMO reactor equipped with a HCPB breeding blanket. Missing functionalities for Geant4 were developed such as a track length detector with multiple energy bins and tally multiplication, reflective boundaries, and statistical checks in a post-processing step. The activation foil measurements in the TBM mock-up revealed Geant4 underestimates the uncollided and fast neutron flux increasingly with penetration depth, while being consistent with MCNP for the low energy neutron flux. Geant4 underestimates the experimental tritium activity in the breeder pellets by up to 2.7% more than MCNP, which is within the experimental uncertainty. The DEMO analyses yielded a 0.97% lower Tritium Breeding Ratio (TBR) of Geant4 than MCNP for a homogenized breeder blanket, and a 1.46% higher TBR of Geant4 than MCNP for a mostly heterogenic blanket. This good agreement with MCNP demonstrates Geant4’s suitability for DEMO TBR evaluations.

Energy resolution and neutron flux of the 4SEASONS spectrometer revisited

The elastic energy resolution, integrated intensity, and peak intensity of the direct-geometry neutron chopper spectrometer 4SEASONS at Japan Proton Accelerator Research Complex (J-PARC) were re-investigated. This was done with respect to the incident energy and the rotation speed of the Fermi chopper using incoherent scattering of vanadium and simple analytical formulas. The model calculations reproduced the observed values satisfactorily. The present work should be useful for estimating in instrument performance in experiments.

A detailed study of the high energy neutron flux (15-20 MeV) at NCSR “Demokritos” using a BC501A liquid scintillator

The current study was carried out for neutron beam energies ranging between 14.8 and 19.2 MeV using the 3H(d,n)4He reaction and included neutron/gamma discrimination and neutron monitoring utilizing a BC501A liquid scintillator, followed by a pulse shape discrimination-capable circuit. Tests were conducted in order to determine the characteristics and limitations of the employed experimental setup. The deconvolution of the acquired spectra was performed using the DIFBAS computer code. The algorithm is based on the Bayesian conditional probability,whilethe covariance filter method was employed to calculate the a posteriori neutron flux spectrum along with its covariance matrix.The produced neutron beam proved to be practically monoenergetic for low energy deuterons, while at higher deuteron beam energies, the strong presence of parasitic and scattered neutrons was revealed.

Simultaneous measurements of heat flow rate and thermal properties of nano boron trioxide under neutron irradiation at the low and high temperature

In the present work, nano boron trioxide (B2O3) samples (purity of 99.5% and true density of 3.10 g/cm3) were irradiated by using neutron irradiation. Neutron irradiation of the samples was carried at the fast neutron flux 4·1012, 8·1012, 1.3·1013, 4·1014 n ⋅cm⁻2 s⁻1 and the neutron flux 1015 n ⋅cm⁻2 s⁻1 and energy of neutron flux E ≥ 1 MeV at room temperature to 350 h at the IBR-2 M pulsed reactor in Joint Institute for Nuclear Research, Dubna, Russia, was used. The unirradiation and after neutron irradiation the thermal kinetics and parameters changes with a differential mechanism. In the DSC curves at the low temperature for unirradiation and irradiation samples does not change. However, at the 100 ≤ T ≤ 900 K temperature range widely endo effects showed separation the water vapor, the absorbed water and hydroxide groups. Under neutron irradiation mechanism of separation of the hydroxide groups, water vapor and the absorbed water dependence of irradiation doses. The heat flow rate kinetics, specific heat capacity, thermodynamic function, degree of [OH] groups in the nano boron trioxide is a more completed.

Ductile to brittle transition in ITER specification tungsten assessed by combined fracture toughness and bending tests analysis

The evaluation of the fracture toughness of tungsten is required for the design of plasma-facing components in order to ensure safe and durable operation in ITER reactor, being under construction in France. During operation, plasma facing materials will be exposed to cyclic thermo-mechanical loads combined with high energy neutron flux, which, in general, reduce the fracture toughness. Characterization of the degradation of the mechanical properties after exposure to the neutron flux involves time consuming and expensive procedures due to nuclear activation and special handling. Subsequently, development of sample miniaturization and protocols to reduce the volume of material under inspection is critical to speed up the progress in R&D. In this work, we propose a combined approach for the reconstruction of the fracture toughness – temperature curve, which is applied in the ductile to brittle transition temperature range. The approach consists of two steps: (i) application of the three point bending tests using miniaturized samples to reveal the transition temperature range on the basis of flexural strain data; (ii) execution of standardized fracture toughness tests at the upper temperature of the transition regime. The results allow the determination of the fracture toughness as a function of temperature with a reasonable accuracy. The validity of the approach has been demonstrated on two commercial tungsten grades produced according to ITER specification and tested in the as-fabricated state. The conclusions are supported by microstructural analysis performed on both standardized and miniaturized samples.

Experimental result of back-streaming white neutron beam characterization at Chinese spallation neutron source

The Chinese spallation neutron source was completed in May 2018 and then subsequently commissioned. The Back-streaming white neutron beam line can be used in neutron nuclear data measurement, neutron physics research, and nuclear technology. In these experiments, it is necessary to know the neutron energy spectrum, the neutron flux, and the neutron beam profile of the neutron beam. In this paper, we present the preliminary measurements of these parameters. The neutron energy spectrum and neutron flux are measured by the time-of-flight method with a fission chamber equipped with <sup>235</sup>U and <sup>238</sup>U samples and a <sup>6</sup>Li-Si detector. The neutron beam profile is measured by a scintillator-CMOS detection system. The preliminary experimental measurements of the beam line are obtained. Among them, the energy spectrum measurement range of white neutrons is from eV to more than 100 MeV, which also gives an uncertainty analysis; the neutron fluence rate gives the full power value of the two experimental halls; the collimated white neutron beam spot is given under a diameter of 60 mm. The future plan is also given. The results of these experimental parameters can serve as the foundation for the future nuclear data measurement and detector calibration experiments of the beam line.

Small modular reactor core neutronic evaluation via Monte Carlo method

The main purpose of this study is a core neutronic parameter evaluation via the Monte Carlo method, for the SMART reactor which is a certified design Small Modular Reactor (SMR). The SMART neutronic parameters such as axial and radial distributions of neutron fluxes, Power Peaking Factors (PPFs), effective delayed neutron fraction, xenon and samarium effects, burnup calculation and neutron flux energy spectrum have been assessed. Daily load follow operation in soluble-boron-free with control regulating banks is one of the best SMART reactor core advantages. Accordingly, the effects of main regulating bank insertion in SMART core have been evaluated. For developed model verification, some of the neutronic parameters have been compared with the SMART Standard Safety Analysis (SSAR) and show proper match. Then other neutronic parameters of the SMART core as a pioneer SMR have been calculated and evaluated.

Small modular reactor core neutronic evaluation via Monte Carlo method

The main purpose of this study is a core neutronic parameter evaluation via the Monte Carlo method, for the SMART reactor which is a certified design Small Modular Reactor (SMR). The SMART neutronic parameters such as axial and radial distributions of neutron fluxes, Power Peaking Factors (PPFs), effective delayed neutron fraction, xenon and samarium effects, burnup calculation and neutron flux energy spectrum have been assessed. Daily load follow operation in soluble-boron-free with control regulating banks is one of the best SMART reactor core advantages. Accordingly, the effects of main regulating bank insertion in SMART core have been evaluated. For developed model verification, some of the neutronic parameters have been compared with the SMART Standard Safety Analysis (SSAR) and show proper match. Then other neutronic parameters of the SMART core as a pioneer SMR have been calculated and evaluated.

Radiation Damage Tests on Diamond and Scintillation Detector Components for the ITER Radial Neutron Camera

During the International Thermonuclear Experimental Reactor (ITER) reactor operation time, the plasma will give rise to high energy neutron and gamma flux, and this intense radiation field will result in serious radiation damage and activation effects on various detectors components. In this paper, neutron detector candidates for the ITER radial neutron camera (RNC), i.e., scintillator components [crystal and plastic scintillators, optical windows, and photomultipliers (PMTs)] and single-crystal diamond detectors, were investigated to establish their radiation hardness and stability under gamma irradiation. Radiation test was carried out at the Italian National Agency for New Technologies, Energy and Sustainable Economic Development Calliope plant (Casaccia R.C., Rome, Italy). The facility is a pool-type irradiation plant equipped with a 60Co source (energy = 1.25 MeV). Gamma radiation tests were performed in the dark and at room temperature for different total absorbed doses, as required for the application in ITER RNC. Scintillators, PMTs, and optical windows were irradiated up to around 100-kGy absorbed dose, while single-crystal diamond detectors up to around 5 MGy. Scintillators and optical windows transmittance measurements were performed in the UV-VIS range (300–700 nm), paying particular attention to the behavior at 390 and 420 nm (scintillating emission wavelengths). Samples were measured in the dark before and after irradiation and their performances were monitored at room temperature for some weeks in order to study the damage recovery in different conditions. For plastic scintillators, photobleaching and optical bleaching followed by thermal annealing processes in air were made to reduce the radiation damage. Quantum efficiency measurements were performed on the PMTs, and the pulse-height spectra and pulse shape capability of the scintillators were investigated by using gamma and neutron sources. The most relevant results of this paper concern the radiation damage observed in scintillators and diamond detectors. The crystalline scintillator sample showed much less radiation resistance than plastic scintillators. A strong damage to the silver deposition of the diamond contacts was observed already at 1.0 MGy, almost leading to contact destruction at 4.7 MGy.