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At least 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Review of Direct Neutron Conversion and Detection Processes

Commercial and laboratory neutron detection systems use indirect neutron response of materials like pressurized helium-3 (via nuclear reaction 3He(n, p)3H) to measure and count neutrons emanating from a source. Recently a host of semiconductors, especially a ternary semiconductor of lithium indium diselenide (6LiInSe2) and a quaternary alloy of enriched lithium-6, indium, phosphorous, and selenium (6LiInP2Se6), have shown promising neutron counting possibilities by directly converting neutrons into charge-carrying elements within the body of the semiconductor. These semiconductors have high thermal neutron capture cross sections, suitable energy bandgaps (~2.0 electron volts) for room-temperature operations, and a favorable electronic band structure for efficient electron charge transport. The article examines the semiconductor properties of these compounds in terms of their neutron counting capabilities and possible ways to extract neutron energy information from them. Lithium-6 and boron-10 (with thermal neutron absorption cross sections of 938 ± 6 and 3855 ± 26 barns, respectively) produce charged particles to be measured via indirect neutron interactions. The efficiency of indirect conversion neutron detectors is limited because of the inefficiencies in conversion mechanism. In case of direct conversion, the neutrons create charged particles in a single material for neutron capture and charge collection, increasing detection efficiency. Unlike 3He proportional counters, which provide no neutron energy information, the semiconductors can be used as neutron energy spectrometer. Fully resolved neutron energy by 6LiInP2Se6 from a plutonium-beryllium source has been reported in the literature. We will discuss the influence of these multilayered semiconductors’ crystallographic structures and growth techniques on neutron energy determination.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2025

Understanding neutron capture processes in uranium deposits using combined U-Sm-Nd isotopic compositions

Valuable insights into the history and evolution of a geologic deposit can be found by investigating neutron capture reactions. Thermal neutron capture reactions occur within both the samarium (Sm) and the uranium (U) systems, where 149 Sm and 235 U can capture neutrons to become 150 Sm and 236 U, respectively. Although largely unexplored, paired measurements of 150 Sm and 236 U could be important for understanding neutron capture effects within uranium ore bodies, and such measurements are potentially useful in nuclear forensics for assessing a material's provenance or mineral exploration. In this work, we refined measurement procedures of Sm isotope compositions utilizing MC-ICPMS. While geologic reference materials were found to have indistinguishable Sm isotope compositions, we found significant isotope variations consistent with nuclear field shift among synthetic Sm standards. Here, this observation highlights that future high-precision Sm isotope investigations need to carefully evaluate synthetic standard(s) against geologic reference materials until an unfractionated and agreed-upon standard is identified. Here, we applied this method to a set of nine uranium ores from the South Australian Beverley North uranium deposits. Although 236 U excesses had been previously reported for these U ores, we found no measurable isotopic shifts in 149 Sm- 150 Sm at the current level of precision (±5 parts per million). One possible explanation for this disparity in the observed neutron capture signatures between U and Sm is that the source(s) of the U and Sm in these ores may be decoupled. This is consistent with the finding that these ores have variable 143 Nd/ 144 Nd, thus demonstrating that diverse sources were involved in the formation of the Beverley North deposits. Alternatively, this deposit may be too young (<50Ma) to have accumulated measurable neutron capture effects in Sm to be detected with the methods employed here.

Mineral exploration↗
DOE OSTI2021

Assessment of MOOSE-Based Tools for Calculating Radial Core Expansion

Radial core expansion in liquid-metal cooled fast reactor systems is a well-known phenomenon that produces strong reactivity feedback effects. An inherently safe reactor design takes advantage of negative reactivity feedback in accident scenarios by utilizing a core restraint system which produces a bowed shape that allows for radial expansion of the fuel regions. Detailed modeling and simulation of core radial expansion itself as well as subsequent reactivity feedback is a challenging task involving contact of many fuel assembly elements and physics feedback from neutronics, thermal hydraulics, and thermal mechanical response. A variety of physics codes have been developed to model aspects of radial core expansion but in general invoke geometrical or physics approximations. No code system currently exists which tightly and robustly couples these physics with enough detail to fully resolve the complex core radial expansion reactivity feedback effects. The future availability of such a code system is of vital importance to fully understanding the reactivity feedback effects that occur due to radial expansion, and consequently to optimizing the design of the core restraint system. A high-fidelity code will also be used to benchmark existing lower fidelity, faster running models to understand their benefits, limitations, and range of applications. A code development path using MOOSE-based tools is proposed in order to leverage the detailed geometry capabilities and natural tight coupling and robustness of MOOSE-based applications for modeling this complex phenomena. While simulation of the full phenomenon involves several physics, an assessment has been initiated on the capabilities and readiness of the currently available Tensor Mechanics module within MOOSE for calculation of the structural mechanical responses which occur within the reactor core. This report focuses on modeling the force-deformation response which mimics the physics of a duct contact deformation, as well as differential thermal expansion which produces a thermal bowed shaped for the fuel assemblies in a core. Simple examples were initially performed such as simple supported beam bending under load. The complexity of examples was progressively increased to better mimic the duct behavior by including a differential thermal example and inclusion of hexagonal cross-sections in the geometry. Further assessment of the structural mechanical response simulation capability is still required for modeling duct contact interactions and irradiation creep and swelling. Companion thermal hydraulic and neutronics assessments will also be required; these activities are planned for future years. Finally, integration of the multiple physics components through MOOSE is required to predict the core radial expansion and subsequent feedback effects.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Fission Induced Radiolysis of Uranyl Sulfate Solutions

At Argonne National Laboratory, an experimental pilot facility is being tested for the production of the medical isotope molybdenum-99 ( 99 Mo). AMORE (Argonne Molybdenum Research Experiment) is an experimental pilot facility in which a 40 MeV electron beam from the LINAC strikes a depleted uranium (DU) target. The target generates high energy neutrons, which are thermalized in an 18-liter uranyl sulfate solution of low enriched uranium (LEU). Thermalized neutrons fission the uranium-235 ( 235 U) to produce 99 Mo as well as many other fission products. The 99 Mo is recovered from the uranium solution and purified. The goal of our work is to reduce the use of highly enriched uranium (HEU) for the production of this isotope.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Phase evolution of U-Zr system in a thermal cycling neutron diffraction experiment: as-cast U-35Zr and U-50Zr

Here, phase evolution of as-cast U-35 wt% Zr and U-50 wt% Zr alloys during thermal cycling (303–1073 K) was investigated using in-situ neutron diffraction. Analysis was performed using Rietveld crystal and microstructure refinements from time-of-flight neutron diffraction data, with a focus on evolution of lattice parameter, atom ordering, unit cell volume, and weight fractions during the thermal cycling. Disordered δ-UZr 2 and residual γ-(U,Zr) phase are retained in the field of δ of the U-Zr equilibrium phase diagram for the U-50Zr sample, and in the field between δ and α-U for the U-35Zr sample. The α phase that is present in the reported phase diagrams of U-Zr was not observed in the diffraction patterns. The evolutions of lattice parameter and unit cell volume of δ and γ are affected by both thermal expansion and chemistry.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Boron-loaded organic glass scintillators

Herein we report the progress towards an organic glass scintillator with fast and thermal neutron sensitivity providing “triple” pulse shape discrimination (PSD) through the inclusion of a boron-incorporated aromatic molecule. Here, the commercially available molecule 2-(p-tolyl)-1,3,2-dioxaborinane (TDB) can be readily synthesized in one step using inexpensive materials and incorporated into the organic glass scintillator at 20% by weight or 0.25% 10 B by mass. In addition, we demonstrate that TDB can be easily scaled up and formulated into organic glass scintillator samples to produce a thermal neutron capture signal with a light yield equivalent to 120.4 ± 3.7 keVee, which is the highest value reported in the literature to date.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Coupled Neutronics - Thermal-Hydraulics - Thermomechanics modeling of Molten Chloride Reactor

Molten Salt Reactors (MSRs) are innovative Generation-IV nuclear reactors known for their enhanced safety features, fuel economy, and fuel cycle advancements. However, designing and studying these reactors present unique safety and modeling challenges. The Department of Energy (DOE)'s Nuclear Energy Advanced Modeling and Simulation program addresses these challenges through the development of flexible multi-fidelity, multiphysics simulation tools. This study exemplifies NEAMS tools' capabilities by modeling a chloride-based micro-MSR: Griffin for neutronics simulations, Pronghorn for thermal-hydraulics simulations, and BISON for thermomechanics simulations. These tools are integrated into a comprehensive 3D multiphysics model, successfully capturing steady-state and transient reactor operations. A loss of flow accident is examined as an illustrative example of transient behavior. Additionally, this paper contributes to demonstrating the theoretical feasibility of micro-MSR concepts, which have received less attention compared to other designs. The subsequent sections provide a brief description of the MSR concept studied, an overview of the computational tools used, and present results for both steady-state operation and a loss of flow accident.

42 - ENGINEERING↗
DOE OSTI2026

Preliminary neutronic and thermal analysis of Inertial Fusion Energy systems with a focus on thick liquid wall concepts

This paper provides an assessment of the neutronic behavior of Inertial Fusion Energy systems using the Serpent 2 Monte Carlo code. The study explores different types of reactor concepts, including dry-wall and thick liquid wall systems. In particular, it characterizes the tritium breeding and energy multiplication performance, as well as the effect of long-term irradiation on structural materials. For this purpose, the displacement per atom (DPA) and the gas production rate in structural materials are calculated. A simplified geometry with varying geometrical parameters is used as a representation for the different kinds of devices. Additionally, the case of the HYLIFE-II concept is used to assess modeling inaccuracies arising from certain approximations. A comparison between the simplified geometry and a realistic one is performed, thanks to Serpent’s capability to run transport calculations on unstructured meshes. The study then addresses the impact of these discrepancies on the temperature distribution in the first wall using the GeN-Foam code for conjugate heat transfer analysis. The results highlight the necessity to consider the heterogeneous nature of the liquid wall, as simplified models tend to underestimate radiation damage metrics and heat loads.

Conjugate heat transfer↗
DOE OSTI2023

A portable and monoenergetic 24 keV neutron source based on 124Sb-9Be photoneutrons and an iron filter

A portable monoenergetic 24 keV neutron source based on the 124Sb-9Be photoneutron reaction and an iron filter has been constructed and characterized. The coincidence of the neutron energy from SbBe and the low interaction cross-section with iron (mean free path up to 29 cm) makes pure iron specially suited to shield against gamma rays from 124Sb decays while letting through the neutrons. To increase the 124Sb activity and thus the neutron flux, a >1 GBq 124Sb source was produced by irradiating a natural Sb metal pellet with a high flux of thermal neutrons in a nuclear reactor. The design of the source shielding structure makes for easy transportation and deployment. A hydrogen gas proportional counter is used to characterize the neutrons emitted by the source and a NaI detector is used for gamma background characterization. At the exit opening of the neutron beam, the characterization determined the neutron flux in the energy range 20–25 keV to be 6.00±0.30 neutrons per cm2 per second and the total gamma flux to be 245±8 gammas per cm2 per second (numbers scaled to 1 GBq activity of the 124Sb source). A liquid scintillator detector is demonstrated to be sensitive to neutrons with incident kinetic energies from 8 to 17 keV, so it can be paired with the source as a backing detector for neutron scattering calibration experiments. This photoneutron source provides a good tool for in-situ low energy nuclear recoil calibration for dark matter experiments and coherent elastic neutrino-nucleus scattering experiments.

Dark Matter detectors (WIMPs, axions, etc.)↗
DOE OSTI2021

Aqueous harvesting of 88 Zr at a radioactive-ion-beam facility for cross-section measurements

Isotope harvesting is a method of collecting the long-lived radioisotopes that build up during the operation of ion-beam facilities in a way that is useful for subsequent research. As a demonstration of this method for the collection of a group IV metal at a fragmentation facility, the high-energy Zr 88 secondary beam produced from a 140-MeV/u Mo 92 primary beam at the National Superconducting Cyclotron Laboratory (NSCL) was stopped in a water target. The setup aimed to mimic the aqueous beam dump that will be implemented at the Facility for Rare Isotope Beams (FRIB). The collected Zr 88 and accompanying Y 88 decay daughter were radiochemically extracted from the solution and made into target samples suitable for neutron-capture cross-section measurements. These samples were then irradiated at two reactor facilities, and the Zr 88 average thermal-neutron-capture cross section ( σ T ) and resonance integral ( I ) were determined to be σ T = ( 8.04 ± 0.63 ) × 10 5 b and I = ( 2.53 ± 0.28 ) × 10 6 b. The σ T value agrees well with previous results and I , determined for the first time here, was found to be the largest measured resonance integral by two orders of magnitude. The Y 88 thermal-neutron-capture cross section was determined to be less than 1.8 × 10 4 b. This work demonstrates the steps needed to make cross-section measurements with samples produced via aqueous isotope harvesting.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Sensitivity of the simulation of passive neutron emission from UF 6 cylinders to the uncertainties in both 19 F(α,n) energy spectrum and thick target yield of 234 U in UF 6

Interest in safeguards verification measurements using passive thermal neutron counting to assay 235 U content in large 30B UF 6 canisters has grown in recent years. Here, the prohibitively high cost and impracticality of using reference 30B calibration cylinders extensively will likely make accurate simulations of increasing interest. Accuracy of the simulated response will define the confidence in the predicted response and the extent to which simulations can reasonably be relied upon. With 234 U driven 19 F(α, n) reactions being the main neutron source in low enriched UF 6 the uncertainties of the 19 F(α, n) energy spectrum and the thick target yield of 234 U in UF 6 propagate into the uncertainty in the predicted response and represent a major influence of basic nuclear data. Here sensitivity of the simulated total (Singles) and coincidence (Doubles) count rates are assessed for the Passive Neutron Enrichment Meter using six potential 19 F(α, n) neutron energy spectra over a range of enrichments and material distributions. The results indicate that variations in the Singles and Doubles due to simulated (α, n) neutron spectrum are less than 1.5% for this set of simulated neutron spectra, with dependence varying inversely with enrichment. Singles uncertainty is only slightly less than that of the thick target 19 F(α, n) yield corresponding to the primary neutron source, whereas the 19 F(α, n) yield dependence of the Doubles is reduced by the non-negligible 238 U(SF) coincident neutron emissions. Based on available thick target 19 F(α, n) yield estimates the uncertainty is on the order of 5%, establishing this as the main nuclear data limitation when simulating thermal neutron detectors response for 30B UF 6 storage cylinders. Based on these findings, it appears that the measurement and evaluation of the thick target 19 F(α, n) yield for uranium hexafluoride is due.

19F(α,n) neutron spectrum↗
DOE OSTI2023

Manufacturing and characterization of a boron-loaded fast-cured plastic organic scintillator

A novel fast-cured, boron-loaded, plastic organic scintillator was developed and characterized. A diisopropylnaphthalene-containing scintillator was manufactured in minutes with the addition of two commercially available fast-curing monomers, isobornyl acrylate (IBOA) and 1,6-Hexanediol dimethacrylate (HDDMA). Further, the light yield of the device was measured to be approximately 70% relative to a similarly sized, but several-year-aged, EJ-276 detector. Using a coincidence capture technique, the neutron capture light output was found to be 87.7 ± 4.2 keVee, which was 1.47 ± 0.07 greater than the light output from a 59.5 keV electron. Producing a viable fast-cured and thermal-neutron-sensitive detector enables the capability to pursue cost-effective 3D printed thermal neutron detectors in a timely manner.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Boron Nitride Solid-State Neutron Detectors

The goal of this project is to build upon the work performed under the Texas Tech IDEAS project (DE-AR0000964 monitored by Dr. Isik Kizilyalli) to bring the developed B-10 enriched hexagonal boron nitride (h- 10 BN or 10 BN or BN) semiconductor neutron detector technology to the next level of maturity. In comparison to He-3 ( 3 He) gas detectors, BN neutron detectors possess all the intrinsic advantages of semiconductor devices: light weight, compact size, fast response speed, high spatial resolution, high temperature and low voltage operation capability, ease of mass production via existing semiconductor infrastructures, low cost of operation and maintenance, flexible form factors, and durability. The project will develop a prototype h- 10 BN neutron detector as a potential replacement for the 3 He gas neutron detectors (LND 25185 or equivalent) used in various neutron survey meters currently produced and commercialized by Ludlum Measurements, Inc. (LMI). To increase the overall detection efficiency, it is essential to maximize the collection efficiency of charge carriers (electrons and holes) generated by the nuclear reaction, which demands a further push to achieve 10 BN epilayers with improved material quality. The material growth processes will be guided by characterization of structural, optical, and electrical properties of h-BN. Selective devices will be calibrated at LMI. During ARPA-E supporting period, the key material parameters of h-BN produced by metal organic chemical vapor deposition (MOCVD), including the layer thickness, mobility-lifetime product, surface recombination field and detector size for maintaining a high charge collection efficiency all have been improved by several orders of magnitude. These improvements have enabled the realization of high-performance h 10BN semiconductor thermal neutron detectors. Presently, h- 10 BN thermal neutron detectors developed under ARPA-E’s support hold the record high detection efficiency among all solid-state detectors at 59%. With further development in manufacturing, there is no question that h-BN detectors will replace the traditional He-3 gas detectors in certain application areas by offering obvious advantages of semiconductor technologies over gas detectors and opportunities for users to dedicate the scarce and expensive supply of He-3 gas to other application areas where substitutes of He-3 gas are not possible.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

First observation of isolated nuclear recoils following neutron capture for dark matter calibration

Low-energy nuclear recoils (NRs) are hard to measure, because they produce few e - /h + pairs in solids—i.e., they have low “ionization yield.” A silicon detector was exposed to thermal neutrons over 2.5 live days, probing NRs down to 450 eV. The observation of a neutron capture-induced component of NRs at low energies is supported by the much-improved fit upon inclusion of a capture NR model. This result shows that thermal neutron calibration of very low recoil energy NRs is promising for dark matter searches, coherent neutrino experiments, and improving understanding of ionization dynamics in solids.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2024

Challenges and Solutions for Fast Neutron Irradiation of Bulk Material Specimens

Reactor developers continue to recognize opportunities for further enhancing fast spectrum reactor designs with advanced core materials, but all the material test reactors currently available to the United States are thermal spectrum designs. Fortunately, the Advanced Test Reactor and High Flux Isotope Reactor are versatile high flux facilities where spectral modification strategies can be used to reduce undesirable thermal neutron capture transmutation damage and augment fast flux delivered to specimens. New opportunities to leverage high flux regions and specially designed fast flux boosting experiment configurations can be used to achieve meaningful fast fluences on large specimens in ATR. New optimization potentials can be employed to achieve even higher fluences, albeit for smaller specimens, using thermal neutron filters in HFIR test positions. These capabilities, while not true fast reactors, can provide highly relevant environments for researchers needing to study the effects of fast neutron damage in bulk material specimens.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

General guide concepts for compact, high-brilliance neutron moderators

The trend in neutron sciences is toward integrating compact, high-brightness moderators into new or upgraded facilities. Transporting neutrons from the source to the sample position with a phase-space distribution tailored to specific requirements is crucial to leverage high source brilliance. Here we have investigated four guide concepts using Monte Carlo ray tracing simulations: Montel beamline with nested Kirkpatrick–Baez mirrors, curved-tapered beamline with a bender and straight sections, straight-elliptical beamline, and curved-elliptical beamline. The straight-elliptical (curved-elliptical) beamline features two half-ellipse guides connected by a straight (non-straight) guide section. The neutron transport efficiency and phase space homogeneity have been quantitatively compared. Our results show that the straight-elliptical beamline performs best because of few neutron bounces on the guide surface with small reflection angles, minimizing flux loss. The Montel beamline provides the best spatial confinement of neutrons within the desired region; however, there is a high thermal-neutron loss due to large reflection angles. The curved-tapered beamline suffers from significant flux loss due to high bounces, and it shows a non-uniform angular distribution related to broad ranges of bounces and reflection angles. The non-straight guide section of the curved-elliptical beamline increases the phase space inhomogeneity, leading to a spatially non-uniform beam profile. The results apply to general neutron instruments that require transporting thermal and cold neutrons from a compact, high-brilliance moderator to the sample location with a moderate phase-space volume.

47 OTHER INSTRUMENTATION↗
DOE OSTI2026

IER-620 CED-3b: Experiment Execution Summary for the Pulsed-Neutron Die-Away Experiments (PNDA) with Propylene Glycol and Mobilmet 423

There is a strong need for new benchmarks to validate neutron thermal scattering laws (TSLs). Lawrence Livermore National Laboratory (LLNL) has designed a Pulsed-Neutron Die Away (PNDA) testbed for this purpose. The experiment has a deuterium-tritium (D-T) neutron generator that impinges a 10 -4 s, mono-energetic pulse of 14.1 MeV neutrons on a target sample. After the pulse, the neutron population moderates and establishes a thermal equilibrium within the sample, with a fundamental spatial mode and characteristic decay-time eigenvalue, ⍺. The ⍺ eigenvalue can be extracted from the experimental measurements of the time-dependent neutron flux coming off the surface of the sample and can then be used as an integral parameter (similar to k eff ) to validate nuclear data involved with neutron migration, thermalization, and absorption. For moderating materials and geometric configurations, the ⍺ eigenvalue is heavily dependent on thermal neutron scattering of the target material. For that reason, a PNDA experiment can have a higher sensitivity to TSLs than is commonly available with the k eff parameter in critical experiments.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Characterization of EJ-270 and Ce-doped LiCAF scintillators for the development of high-rate neutron reflectometer detectors

The Second Target Station of the Spallation Neutron Source at Oak Ridge National Laboratory is anticipated to provide a neutron source with ∼20 times increase in peak brightness than the First Target Station. The neutron reflectometers currently in operation at the First Target Station need to be upgraded due to the increased neutron flux. A prototype neutron detector module based upon a pixelated scintillator array readout by silicon photomultipliers is being developed to address the high-rate challenge faced with future neutron reflectometer instruments at the Second Target Station. Two types of scintillator materials were considered for this detector development, i.e., 6 Li-loaded EJ-270 plastic scintillator and Ce-doped LiCAF single crystal. This paper reports the scintillator characterization results, including light yield, pulse shape discrimination performance, capability to detect thermal neutrons in a high γ-ray field, and γ-ray sensitivity. The number of photons produced per neutron capture by EJ-270 and LiCAF:Ce was measured to be 2176 ± 91 and 2651 ± 108, respectively. EJ-270 demonstrated a good capability to discriminate between neutrons and γ-rays by employing the commonly used charge comparison method (figure-of-merit: 1.13 ± 0.01 for an energy cut of 292–426 keVee) and a reasonable performance when using the time-over-threshold techniques; however, no discrimination was observed from LiCAF:Ce regardless of the pulse shape discrimination approaches utilized, making pulse height discrimination necessary for LiCAF:Ce to differentiate between neutrons and γ-rays. Both EJ-270 and LiCAF:Ce exhibited an acceptable capacity to detect thermal neutrons at high exposure rates up to approximately 584 mR/h. Furthermore, the γ-ray sensitivities measured with a 60 Co source at an exposure rate of around 1145 mR/h were determined to be (6.11 ± 0.87) × 10 −6 and (7.64 ± 1.08) × 10 −7 for EJ-270 and LiCAF:Ce, respectively.

EJ-270↗