Neutron Spectrum Research Articles

Development of a graphite pile to produce a neutron thermal field at the CIEMAT - Neutron standards laboratory

The CIEMAT Neutron Standards Laboratory (LPN) is the Spanish national reference laboratory in neutron metrology. Within the framework of the LPN expansion project, it has been planned to incorporate a thermal neutron field based on a graphite pile. A study has been carried out with MCNP 6.2 for a graphite pile with dimensions 180 × 150 × 150 cm3, with a 185 GBq Am–Be source (B = 1.1·107 s−1) and the internal cavity with a section of 40 × 40 cm2 and a depth of 50 cm. The main parameters of the graphite pile and the position and emission rate of the Am–Be neutron source have been varied to determine the fluence rates at several points in the pile, their thermal component and homogeneity. The optimal configuration produces a neutron spectrum with more than 98% of thermal component, fluence rates around 400 cm−2 s−1 inside the cavity and 55 cm−2 s−1 outside, and acceptable homogeneity in this thermal neutron field. From a Radiation Protection point of view it is necessary to incorporate borated high-density polyethylene (HDPE-B) around or at least as a shielding wall, and a Cd window to filter the thermal component.

Investigation of Cs27LiYCl6:Ce scintillator energy response for D-D fusion neutron spectrometer

During plasma experiments on EAST (Experimental Advanced Superconducting Tokamak), the discharges heated with neutral beam injection (NBI) for plasma results in the production of 2.45 MeV deuterium-deuterium (D-D) fusion neutrons. In order to measure the energy spectrum of D-D fusion neutrons, a new Cs27LiYCl6:Ce (CLYC-7) crystal with 7Li-enrichment has been developed and fabricated into a CLYC-7 scintillation detector. Before utilizing this detector for studying neutron energy spectra, it is crucial to explore the response function of the detector within a relevant neutron energy range. Therefore, the detector has been experimented using the 4.5 MeV electrostatic accelerator at Peking University to investigate its response to fast neutrons, energy resolution, and pulse discrimination capability in the neutron energy range 2.07 MeV–5.51 MeV. The neutron reactions occurring within the CLYC-7 scintillator were simulated using MCNPX, focusing on the reactions 35Cl(n,p) and 35Cl(n,α), as well as the energy response curve and the contributions of different particles (protons, alphas) to the neutron spectrum. The experimental results were successfully validated on EAST, where the 2.45 MeV fusion neutron energy spectrum was measured. That demonstrates the ability of the CLYC-7 scintillation detector to diagnose the fusion neutron energy spectrum for deuterium plasma discharge.

Development and validation of a Bayesian unfolding program using OpenBUGS open-source software for Bonner spheres neutron spectrometry

This study presents the development of a novel unfolding program based on a Bayesian method for Bonner spheres spectrometry. The program incorporates a priori information on the neutron spectrum, employing a linear superposition of three non-negative parametric functions: thermal-Maxwellian distribution, an intermediate region represented by a straight line in lethargy space (1/E behavior), and the Watt fission distribution for the fast region. The effectiveness of the developed program was validated through its application to two reference cases from the EURADOS 2018 exercise (LINAC and Skyshine), followed by its implementation in characterizing the neutron field around an OB26 neutron irradiator at the Secondary Standards Dosimetry Laboratory (SSDL) of the Nuclear Research Center of Algiers (CRNA), Algeria. While validation results for the LINAC case were promising, achieving comparable neutron spectrum shape and integral quantities to the reference, challenges were encountered with the Skyshine case. However, these challenges were partially addressed through sensitivity analysis, resulting in slightly improved outcomes compared to the previous participation and to the reference. Application of the program to characterize the neutron field around the OB26 irradiator yielded satisfactory results when compared to reference data. A two-step sensitivity analysis further enhanced the outcomes, initially substituting the Watt fission distribution with the evaporation distribution, thereby improving spectrum shape and integral quantities. Subsequently, utilizing the obtained neutron spectra as prior information for unfolding codes MAXED and GRAVEL ensured stability and agreement with reference data and preliminary results.

Neutronic design analysis of CEA Cadarache URANIE subcritical assembly

This paper aims to study neutronic modeling using decommissioned fuel pins of the URANIE subcritical assembly at the Cadarache center of CEA for the future Tunisian Subcritical Assembly. Indeed, the CNSTN intends to exploit the fuel to implementing its subcritical assembly. Therefore, a comprehensive computational study of the URANIE fuel was carried out to analyze the neutronic parameters for future facilities including criticality safety, the neutron flux, and neutron spectrum, for future facilities. Indeed, MCNP Monte Carlo code was used to develop a three-dimensional sketch for URANIE core. Two versions were used to estimate an effective multiplication factor (keff): MCNP5 and MCNP6. The keff was varied from 0.85842 to 0.86463 with a maximal difference of 623 pcm and depended on ENDF/B data library releases. The nominal value of keff is acquired at the optimum lattice pitch of 4.6 cm, 30 cm in reflector thickness, and 185 cm in moderator level. Moreover, the reactivity of the facility is always negative. In addition, the energy spectrum was depicted at the three planes using 283 energy groups. Furthermore, the radial neutron flux was evaluated at eight positions. The axial neutron flux was estimated at various spatial planes of the assembly core. Inside the reactor core, the findings revealed indicate that the fast flux is higher and can be reach 7 × 10−4 n/cm2.Q and decreased steadily moving away from the center to the reflector part about 8 × 10−4 n/cm2.Q.

Characterisation of the neutron sources in use at STUK SSDL through a new transfer instrument with spectrometric capabilities

Characterising neutron sources for calibration or generic testing purposes is a complex task, as the neutron spectrum may depend, to some extent, on the construction characteristics of the source. Bonner spheres (BS) are the traditional spectrometric transfer instruments, but they are very sensitive to room scatter and required measurements are highly time consuming. The recently developed NCT-WES device, belonging to the family of the Single Moderator Neutron Spectrometers, has been proposed as a convenient alternative to BS. This work presents the results of a demonstration campaign organised at the STUK neutron metrology laboratory (Finland) in July 2023, where the spectra from two 252Cf and two 241Am-Be sources were determined. For the 241Am-Be sources, the results were compared to the spectra of two categories recently introduced by ISO 8529-1. Based on the results it was concluded that the “large” source category is an appropriate selection for STUK sources. The campaign also proved the operational advantages of NCT-WES as spectrometric transfer instrument.

Optimization of photoneutron source for use in subcritical reactors

This work is an effort to maximize the number of output neutrons of a photoneutron target for 100 MeV incident electrons. The work steps were; to select the appropriate material, select the appropriate shape, and find the optimized dimensions. The simulations of this work were done with MCNPX2.6 simulation code. At first, U-235 was selected as the best material for the target, among some heavy atoms from the number of outlet neutron point of view. Then the best shape for the target was selected from geometric shapes that have been considered for the target and the dimensions of the target were optimized. After these parts, the energy spectrum of outlet neutrons was estimated and after that, the deposited energy of neutrons, electrons, and photons in the target was estimated and drowned with Tec plot.

Thermal and Fast Neutron Measurement in the STU Mini Labyrinth Experiment

The Mini Labyrinth experiment is a neutron and gamma shielding experiment constructed at the Slovak University of Technology in Bratislava (STU). The STU Mini Labyrinth consists of NEUTRONSTOP shielding, blocks of moderators, various neutron sources, a graphite prism, and the detector handling robot. It was designed for research and education purposes, while several experiments are also available online or a hybrid form. There have been several versions of the Mini Labyrinth developed, while currently the V3 version is available. This paper presents the methodology to effectively perform thermal and fast neutron measurements using a PuBe neutron source. In the experiment presented in this paper also active and passive detectors were used, and moderator materials were investigated to slightly shift the neutron spectrum. As for active detectors, the SNM-11 boron coated corona detector was used. Among passive detectors the CR-39 track detectors were investigated. The measurements were carried out in two configurations and the results were evaluated by simulations using the SCALE6 system

Preliminary design and analysis activities of the deuteron accelerator target for tritium breeding unit test

A breeding blanket (BB) is a key component that multiplies neutrons and generates tritium in the fusion demonstration reactor. Evaluation of the performance and reliability of the BB is a very important engineering task, and it is going through the ITER TBM program. The Korean DEMO program needs to complement the ITER TBM program to validate the BB, and the test facility is necessary to validate the BB in the continuous wave operation. Based on this need, the Korea Institute of Fusion Energy has initiated a pre-conceptual study to verify the long-term performance and reliability of the BB.This study deals with neutronics and engineering analyses of a solid beryllium (Be) target that generates neutrons through a deuteron accelerator for the fusion neutron source. In the neutronics analysis, the yield and energy spectrum of neutrons generated according to the incident particles and energy were analyzed. As a result, in the case of 40 MeV deuteron, the yield of neutrons with 10∼20 MeV similar to the fusion neutron was the highest in the forward direction of 1.08E+15 n/s. In addition, vanadium (V) was selected as the blistering mitigation layer (BML), and the thickness of the Be target and the BML were optimized through neutronics analysis to minimize the blistering. In the engineering analysis, a preliminary conceptual design of the target to cool the nuclear heating caused by the nuclear reaction was performed. To improve the cooling capacity of the target composed of Be-V-CuCrZr with the pressurized water coolant, the design of the target was optimized and its thermal integrity was evaluated. Next, thermal stress was evaluated by coupling the temperature distribution of the target obtained through thermal analysis with structural analysis. Through this, the preliminary design concept of the target that satisfies thermal and mechanical integrity requirements was derived.

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.

Characterization of a diamond detector during experiments under wide energy spectrum neutrons

The detection of wide energy spectrum neutrons is important in nuclear data measurement and nuclear physics. Diamond detectors play a significant role in experiments based on wide energy spectrum neutron beams due to the excellent radiation resistance and electrical properties. In usage, the polarization effect and pumping effect of diamond detectors can significantly influence the stability of the diamond detector. For the first time, we conducted a comprehensive characterization of the performance of a diamond detector under wide energy spectrum neutrons irradiation by extracting and analyzing key parameters including Pulsewidth, Peakmax, Qcal, and count rate. This work provides a technical procedure for assessing the stability of diamond detectors in complex radiation fields. Signals of different time-of-flight (ToF) can be selected based on the experimental objective, allowing for the observation of the stability of parameters of interest, such as Qcal and count.

Method of the known cross sections for calibration of the fast neutron spectrometer with a single-crystal stilbene based detector

The present work is devoted to implementation of the stilbene-based neutron spectrometer energy calibration method. The results of experiments with portable neutron generators and 238PuBe source and scattering materials with known cross sections are used for this method. It is shown that the submitted method makes it possible to carry out fast neutron spectrometry in the energy range from 1 to 15 MeV with the uncertainty of the unfolded neutron energy no more than 200 keV. Neutron spectra unfolding was carried out based on the measured spectra and a Geant4 simulated response matrix. Unfolded spectra were compared with the literature data and reference spectra.

Ion Kinetics and Neutron Generation Associated with Electromagnetic Turbulence in Laboratory-Scale Counterstreaming Plasmas.

Electromagnetic turbulence and ion kinetics in counterstreaming plasmas hold great significance in laboratory astrophysics, such as turbulence field amplification and particle energization. Here, we quantitatively demonstrate for the first time how electromagnetic turbulence affects ion kinetics under achievable laboratory conditions (millimeter-scale interpenetrating plasmas with initial velocity of 2000 km/s, density of 4×10^{19} cm^{-3}, and temperature of 100eV) utilizing a recently developed high-order implicit particle-in-cell code without scaling transformation. It is found that the electromagnetic turbulence is driven by ion two-stream and filamentation instabilities. For the magnetized scenarios where an applied magnetic field of tens of Tesla is perpendicular to plasma flows, the growth rates of instabilities increase with the strengthening of applied magnetic field, which therefore leads to a significant enhancement of turbulence fields. Under the competition between the stochastic acceleration due to electromagnetic turbulence and collisional thermalization, ion distribution function shows a distinct super-Gaussian shape, and the ion kinetics are manifested in neutron yields and spectra. Our results have well explained the recent unmagnetized experimental observations, and the findings of magnetized scenario can be verified by current astrophysical experiments.

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Dual particle imaging: Applications in security and environmental imaging

Radiation imaging detectors have been applied in a variety of fields for detection, localization and characterization of radioactive sources. In this work, we describe a handheld detection system (H2DPI) that can simultaneously image neutrons and gamma rays, and reconstruct the energy spectra of the incident particles. This system was applied to the detection and characterization of a plutonium metal source in experiments performed at the National Criticality Experiments Research Center, where the 4.5 kg plutonium source was placed approximately 60 cm from the H2DPI. Neutron and gamma-ray images were measured that show good agreement with the known location of the source. Furthermore, the H2DPI was also used to measure the neutron spectrum, which shows the expected shape of a Watt fission spectrum, and the gamma-ray energy spectrum, which shows the energies common to plutonium metal. Lastly, we describe the application of the H2DPI in augmented reality, whereby real-time neutron and gamma-ray imaging data can be streamed to the Microsoft HoloLens to allow the user to visualize the source locations.

Unfolding neutron spectra using splines

Organic scintillators detect neutrons by scattering reactions with hydrogen nuclei (protons) in organic compounds. The recoil protons then produce the light output from the scintillator. The energy spectrum of neutrons impinging on the scintillator has to be determined from this light output using a so-called unfolding algorithm. However, this problem is ill-posed and therefore difficult to solve. In this study, a new method for unfolding neutron spectra is presented. This method uses B-splines approximation of neutron energy spectrum. Results of this method are shown for test data and experimental data and compared to the results of the previously developed unfolding method.

A Micromegas-based gaseous detector for neutron-induced charged-particle reaction studies in nuclear astrophysics

The quasistellar neutron spectrum produced via 7Li (p, n)7Be reaction at a proton energy of 1.912 MeV has been extensively studied and employed reaction for neutron-induced reaction studies. We are working towards using this reaction at various proton energies from 1.9 MeV to 3.6 MeV to produce a neutron field at a temperature range of ∼ 1.5–3.5 GK to conduct measurements of neutron-induced charge particle reaction cross sections on various unstable nuclei at explosive stellar temperatures. In this paper, we are reporting our design and simulation study with regards to experimental set-up and a gaseous detector with a segmented Micromegas detector for conducting neutron-induced charge particle reactions studies for nuclei of astrophysics importance, for example, 26Al(p, n)26Mg, 26Al(n, α)23Na and 40K(p, n)40Ar, 40K(n, α)37Cl reactions. We plan to perform our experiments with a 10-μA proton beam at the Physikalisch Technische Bundesanstalt facility (PTB, Germany), with a Micromegas-based gaseous detector under construction as discussed in the paper.

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.

Neutron Yield Predictions with Artificial Neural Networks: A Predictive Modeling Approach

The development of compact neutron sources for applications is extensive and features many approaches. For ion-based approaches, several projects with different parameters exist. This article focuses on ion-based neutron production below the spallation barrier for proton and deuteron beams with arbitrary energy distributions with kinetic energies from 3 MeV to 97 MeV. This model makes it possible to compare different ion-based neutron source concepts against each other quickly. This contribution derives a predictive model using Monte Carlo simulations (an order of 50,000 simulations) and deep neural networks. It is the first time a model of this kind has been developed. With this model, lengthy Monte Carlo simulations, which individually take a long time to complete, can be circumvented. A prediction of neutron spectra then takes some milliseconds, which enables fast optimization and comparison. The models’ shortcomings for low-energy neutrons (<0.1 MeV) and the cut-off prediction uncertainty (±3 MeV) are addressed, and mitigation strategies are proposed.

Improving the reactor safety aspects by the implementation of (Th-U233-Pu) fuel in a PWR assembly

The aim of this work is not only to demonstrate how to introduce thorium as an alternative nuclear fuel into pressurized water reactors, but also to examine the several advantages that (Th, U233, Pu) has over (Th, U233) and (Th, Pu) fuels when (U233O2) is mixed with (Th-Pu)O2 fuel. To investigate these advantages, three thorium fuels, (Th, U233), (Th-U233-Pu) and (Th-Pu) are simulated and analyzed by the MCNPX 2.7 code. The reactor safety aspects of these fuels are compared with those of the reference UO2 fuel. The concentration of Th-233 and fissile isotopes is adjusted to obtain criticality (Keff = 1.03) at the same burnup value (35 GWd/tonne) for all fuel types. Based on the predicted results, the author revealed that: 1) 56 % of plutonium (238-242) is depleted in the case of (Th-U233-Pu), while 41 % of plutonium is depleted in the (Th-Pu) at 60 GWd/tonne. This result indicates that (Th-U233-Pu) is more proliferation-resistant than (Th-Pu). 2) In terms of improving the quantity of radioactive waste, (Th, U233, Pu) model produces 0.6 and 0.9 times the amount of Am (241-243) and Cm (242-244) in UO2, respectively. On the other hand, (Th, Pu) fuel produces about 7 times the amount of Am and 8 times the amount of Cm in UO2 fuel. 3) The fuel temperature coefficient (FTC) and moderator temperature coefficient (MTC) of (Th-U233-Pu) are more negative than that of UO2 and (Th-U233) at the early stages of fuel burnup (0–5) GWd/tonne.4) the total mass of fissile isotopes in (Th-U233-Pu) is greater than that of UO2 and (Th-U233) and as a result, the fuel life time is longer in the case of (Th-U233-Pu). 4) The conversion ratio provided by (Th-U233-Pu) is greater than that provided by UO2 and (Th-U233).This behavior is expected owing to the hardening of neutron spectrum in (Th-U233-Pu). 5) The peaking factors are lower in the case of (Th-U233-Pu) compared to (Th-Pu). 6) The concentration of xenon and samarium is noticeably lower for (Th-U233-Pu) in comparison with (Th-Pu) and consequently, the time required for restarting the reactor is accelerated.

Maxwell Spectrum as a Parameter to Verify the Dose in Brain Cancer (Glioblastoma) by Boron Neutron Capture Therapy (BNCT) using Monte Carlo Method

To evaluate the efficiency of neutron capture therapy (BNCT) treatment in glioblastoma multiforme, it is necessary to evaluate the impact of the neutron beam on the tumor cell and find better results so that BNCT treatment is viable. Glioblastoma multiforme is one of the most lethal cancers and conventional radiotherapy is almost ineffective for this type of tumor. Among several approaches to describe the procedure and the neutron spectrum, the Maxwell spectrum in the epithermal neutron range was used. For this, T=0.0025 MeV was used to describe this spectrum. MCNP software was used to simulate a BNCT treatment using the Maxwell spectrum to describe the neutron source. The user provided the quantities of interest, such as fluence and dose. These are extremely important quantities to describe a BNCT planning protocol. A concentration of 30 ppm of Boron-10 was simulated in the tumor. Output data provides normalized values. It was necessary to carry out some mathematical operations to obtain values closer to reality. Thus, a dose of 32 Gy was obtained for the Maxwell spectrum described with T=0.0025 MeV and a neutron fluence of 1.5 x 1012 n/cm². The values calculated based on the simulation in MCNP5 described by an epithermal neutron source obeying a Maxwellian function, were in agreement with the reference values in the literature.