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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Characterization of Neutron Emission Rates of Commercial 252 Cf Sources

Accurate knowledge of 252 Cf neutron emission rates is critical for modeling and fast-neutron experiments, yet commercial sources are often supplied without traceable calibration or uncertainty estimates. This report describes a method to characterize 252 Cf sources using a pulse-shape-discrimination (PSD) scintillator. The scintillator was efficiency-calibrated against a time-tagged 252 Cf fission chamber manufactured at ORNL over 40 years ago. Isotopic evolution of the calibration source was modeled using Bateman equations to account for changes in the effective average neutrons per fission at the time of measurement. The calibrated scintillator was then used to measure absolute emission rates of several commercial 252 Cf sources, revealing deviations of up to 12% from nominal manufacturer values. This methodology provides a practical approach for establishing uncertainty-quantified 252 Cf neutron emission rates, improving fidelity in simulations and supporting the accurate interpretation of measurements.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2024

Boron-10 Doped Polysiloxanes as Matrix Materials for Application in the Simultaneous Detection and Discrimination of Gamma Rays and Fast and Thermal Neutrons

Three boron-10 enriched aromatic molecules have been synthesized and incorporated into two different commercial polysiloxane resins, Shin Etsu KER-6000 and Wacker SilRes H62-C. Scintillating fluorophores, 9,9-dimethyl-2-phenylfluorene (PhF) and 2,5-diphenyloxazole (PPO), were tested in combination with each resin and boron-10 molecule for the simultaneous detection of gamma rays, fast neutrons, and thermal neutrons. Here, the H62-C resin was able to solubilize a large amount of PhF, in excess of 20 wt%. Cure kinetics were controlled through the addition of divinylbenzene and phenyl tris(dimethylsiloxy)silane cross-linker solution to the H62-C resin, with rheology experiments demonstrating a large reduction in time to gelation when 20 wt% cross-linker solution was added, from more than 4 hours to less than 1 hour. These polysiloxane resins can be cured in 3 hours under air, while common poly(vinyltoluene) scintillators require at least 4 days of heating and oxygen-free conditions. PhF-doped KER-6000 with tolyl boronate pinacol ester exhibited the best overall performance with a light yield of 62% relative to EJ-200 and thermal neutron capture at energies up to 103 keVee (84 keVee for EJ254B-5). Additionally, four samples exhibited light yields surpassing an industry-standard boron-doped plastic scintillator, Eljen’s EJ254B-5. Over the course of ten months, the KER-6000 samples showed precipitation of dopant molecules, which reduced their light yield by 15% on average, while H62-C proved to be more stable with only a 6% reduction.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

A compact neutron spectrometer system

A portable neutron spectrometer, nSpec, has been developed for passive interrogation of fast neutrons ( 1 MeV) from neutron sources. The design of the system includes four EJ-301 liquid scintillation cells, a custom stand, and tungsten caps for gamma-ray shielding. Data acquisition settings were optimized to provide a dynamic range of up to > 15 MeV neutron energy. Additionally, the detector response matrix was characterized using a time-of-flight method based on gamma-ray tagged neutron emission from 252 Cf. The spectrum unfolding was characterized using measured and simulated spectra of AmBe and 252 Cf with the GRAVEL and MLEM methods. Monte Carlo methods were used to examine the uncertainty propagation and determine sufficient counts for a measurement.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Experiments with neutron induced neutron emission from U-235, Pu-239, and graphite

A neutron induced neutron emission experiment was conducted as the Los Alamos Neutron Science Center (LANSCE) facility at Los Alamos National Laboratory (LANL). In this experiment, a sample was placed in a well collimated neutron beam and was surrounded by an array of 28 fast neutron detectors (EJ-309). The experiment was performed with a neutron flight path of 21.5 m from the source to the sample, and 1 m from the sample to the detectors. The neutron emission from the sample was measured as a function of neutron time of flight covering an incident energy range from 0.7- 20 MeV. The samples included U-235, Pu-239, carbon (graphite), and blanks that matched the encapsulation of the sample. The measured samples were constantly cycled in and out of the neutron beam. This type of experiment measures neutron emission from all reactions occurring in the sample such as fission and elastic and inelastic scattering. Similar to the methodology previously developed at RPI, the measurements were compared with detailed simulations of the experiment using different cross section evaluations for the sample. The observed differences can be attributed to the evaluated neutron cross section and angular distributions. The carbon sample was used as a reference to validate both the experiment and simulation methodology and showed good agreement between experiments and simulations. A review of the experimental setup, analysis methods, and some of the results will be presented.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Compounds and composition for preparation of lithium-loaded plastic scintillators

A scintillator material includes a polymer matrix, a primary dye in the polymer matrix, the primary dye being a fluorescent dye; a secondary dye, and a Li-containing compound in the polymer matrix, where the Li-containing compound is a Li salt of a short-chain aliphatic acid. In addition, the scintillator material exhibits an optical response signature for thermal neutrons that is different than an optical response signature for fast neutrons.

Zaitseva, Natalia P.↗
DOE OSTI2022

Characterizing the Performance of Fission Chambers for Local Neutron Flux and Spectrum Measurement

Three fission chambers were tested at The Ohio State University Research Reactor (OSURR) to evaluate fission chambers for neutron flux and spectrum measurements. Two of the fission chambers are commercial Photonis CFUR43 fission chambers—one utilizing highly enriched uranium-235 fissile deposits for thermal neutron detection and the other with highly depleted uranium-238 for fast neutron detection. This experiment was also performed in collaboration with French Alternative Energies and Atomic Energy Commission (CEA) with CEA suppling a uranium-235 loaded fission chamber built from the Photonis CFPR CE8/9 kit. The fission chambers were irradiated in a movable 6.5 inch dry tube at reactor powers of 50 kW, 200 kW, and 450 kW at ambient temperature and at 350°C. While all sensors demonstrated good linear sensitivity to reactor power, a leakage current of 6 nA was measured for the CFUR43 detectors. Additionally, the measured signal ratios of the two CFUR43 detectors did not match the theoretical values and are subject to further investigation on individual signal contributors.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Safeguards Facility Model for Lead-cooled and Gas-cooled Fast Reactors

Efforts have been underway for many years on the development of safeguards for fast neutron reactors. The efforts have been primarily for sodium-cooled fast reactors (SFRs) in cooperation with Japan, to properly implement these safeguards as the reactors are conceptualized and ultimately commercialized. The primary challenge to development of adequate safeguard approaches for fast reactors has been in understanding how to apply different aspects of safeguards when facilities widely vary from one type of fast reactor to another and is largely dependent on several key aspects of the specific reactor type. For this report, the following key areas were identified for further analysis to determine where Lead-cooled Fast Reactors (LFRs) and Gas-Cooled Fast Reactors (GCFR) fuel material flow and key measurement points would differ from previously studied SFR safeguards.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Genetic algorithm optimization of tin–copper graded shielding for improved plutonium safeguards measurements

Plutonium nuclear material safeguards is essential to ensure that states do not divert plutonium to clandestine nuclear weapons programs. Traditional safeguards methods rely on correlated neutron measurements. Detection systems based on organic scintillators, which are sensitive to fast neutrons and photons, are gaining support for use in safeguards applications due to increased capabilities and shorter measurement times compared to the current state-of-the-art. A drawback of organic scintillators in plutonium measurements is data throughput limitations; 241 Am builds up in plutonium samples and emits 60-keV gamma rays with a high specific activity. In this work, a genetic algorithm is used to design graded shielding made of tin and copper to attenuate the 60-keV gamma rays. The shield is designed to limit the attenuation of the higher-energy, fission photons, because fission photons can be used as a signature. The genetic algorithm optimization takes days to run, while a complete search space exploration would take years. The genetic algorithm was run for different levels of 60-keV gamma-ray attenuation, and the thickness of the required shielding layers was determined at each level. Tin–copper shielding was placed in front of an array of organic scintillation detectors for a measurement of approximately 4.5 kg of alpha-phase, weapons-grade plutonium, an AmLi source, and a 252 Cf source. We show that the shield successfully attenuated low-energy, non-fission gamma rays, which significantly improved the organic scintillator performance.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

Feasibility of an Active Interrogation System to Classify Waste with He-4 Neutron Spectroscopy

This work investigates a 4 He-detector active interrogation system that leverages neutron spectroscopy to classify nuclear waste streams. MCNP models tested the concept through the simulation of a D-D neutron generator, an array of 4 He detectors, and various waste compositions. The fast-neutron Differential Die-Away signature was augmented with a neutron-energy discrimination signature. This signature isolates induced fission neutrons, the energy of which is greater than that of the D-D monoenergetic spectrum. With the incorporation of this spectroscopic technique, the measurement time decreased by 3–9% (depending on the degree of neutron moderation and absorption presented by the sample), demonstrating how neutron spectroscopy can enhance active interrogation methods. The reduced measurement times would have significant financial and logistical benefits for facilities with large footprints of low-level waste production.

4He detectors↗
DOE OSTI2023

Analysis of WIRE-21 SPND and Optical Fiber Sensor Measurements

The Wireless Instrumented RB Experiment 2021 (WIRE-21) was a highly instrumented experiment designed to test prototype wireless temperature and pressure sensors developed by Westinghouse Electric Corporation (WEC) in the High Flux Isotope Reactor (HFIR). In addition to these sensors, a suite of other sensors was also integrated into the experiment capsule, including various types of distributed optical fiber sensors and self-powered neutron detectors (SPNDs). This report presents an analysis of the data generated by the SPNDs and optical fiber sensors that were tested under the highest reported neutron flux. The distributed optical fiber sensor results demonstrate that F-doped optical fibers, particularly those inscribed with fs fiber Bragg gratings (FBGs), are capable of surviving fast neutron fluences on the order of 1021 n fast /cm 2 at temperatures relevant to light-water reactors (between 200 and 400°C). However, a significant blue-shift in the optical spectra of these sensors was observed over the course of irradiation which cannot be explained based on the current understanding of radiation-induced compaction in fused silica glass. A mechanistic understanding of this drift has not yet been developed and is proposed as future scope under the Advanced Sensors and Instrumentation program. Of the four SPNDs, only one appeared to operate normally during three cycles of irradiation. A radiation transport model of the experiment in HFIR was used to determine time-dependent neutron flux in this SPND and to calculate the neutron sensitivity of the device. Results showed a cycle averaged sensitivity of 1.5 ×10 -22 and 1.4 ×10 -22 A/nν for the first and third cycles of irradiation, respectively. Additional details pertaining to the optical fibers and SPNDs are included herein.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Towards Three-Dimensional Neutron Imaging with Light-field Technology

A broad range of applications in nuclear safeguards and security can benefit from compact, fast neutron imagers with large angular acceptance. However, accurate 3D multi-vertex reconstruction in a monolithic detector remains a significant barrier to realization. Current approaches, such as the centroiding approach and the use of shadow masks, have not yet resulted in a successful demonstration in a light-starved environment. This project explored the use of cutting-edge technology—a light field camera—which inherently preserves both the spatial and directional information of incident light. Such technology can, in principle, resolve multi-vertex events and offer accurate position reconstruction in 3D with relatively simple readout electronics. Since the project's inception in May 2023, we have built an experimental setup to calibrate and characterize a commercial light-field camera (Lytro Illum). We assessed its 3D event reconstruction capability in a relatively low-light intensity environment by analyzing the cross-correlation of a series of 2D images of a characterized tunable light source taken at various distances. A sub-cm resolution, lower than typical neutron interaction separations in a compact scintillator volume, was observed with the LED light source, suggesting the promising capability of its nominal optics design to provide the adequate resolution required for a compact monolithic neutron directional detector. However, using a light-field-based readout for a neutron camera requires incorporating an ultra-low-noise light sensor, which is beyond the scope of this project. We also initiated the development of a 3D light-field-based reconstruction algorithm tailored to sparse scenes, as expected from particle interactions in a scintillator medium. Additionally, we demonstrated the capability of the algorithm to replicate the ground truth. Finally, we began the development of a neutron directional detector simulation to determine the performance criteria for light-field-based reconstruction that allows for good neutron directional reconstruction. This Feasibility Study has led to a successful follow-up project under the DNN R&D innovation portfolio starting May 2024.

42 ENGINEERING↗
DOE OSTI2020

Neutron heating assessment in Minerve reactor by using alanine/ESR dosimetry

In the framework of the ICERR (International Center based on Research Reactors) project, the effect of neutron/gamma irradiation on material and dosimeters has been studied. In this work, we explore the alanine dose-response in the core of the MINERVE research reactor at the CEA Cadarache and we investigate the neutron heating effects in order to evaluate alanine neutron sensitivity. The measured dose of the alanine detector in the mixed gamma-neutron field was estimated by electron spin resonance spectrometry (ESR) with reference to an absorbed dose in water from a 60 Co gamma-ray beam. The experimental sources of uncertainties were reduced by using the optimum conditions of charged-particle equilibrium (CPE), obtained by introducing alanine dosimeters into the aluminum pillbox. Cavity correction factors and dose components were evaluated by the Monte Carlo code MCNP. The calculation showed an underestimated neutron dose with a (C/E) value of about 0.993 +/- 10.83%(k = 1). Results indicate that the fast neutron dose, mainly due to elastic neutron scattering on hydrogen nuclei, represents about 70% of the total dose, while the gamma dose was found to be around 45% of the neutron dose. Additionally, the neutron relative efficiency (REn) to 60 Co gamma-radiation was estimated to be 0.544 +/- 8%(k = 1).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Design of Premoderated Pulsed-Neutron Die-Away Experiments for Validating Beryllium Thermal Neutron Scattering Laws

Accurate thermal scattering laws (TSLs) are essential for reliable neutron transport simulations, particularly for systems with low-energy neutrons. Pulsed-neutron die-away (PNDA) experiments provide a highly sensitive platform for TSL validation but face challenges from fast neutron leakage and material constraints in nonhydrogenous targets such as beryllium. This work investigates the use of high-density polyethylene (HDPE) as a premoderator to reduce leakage, improve thermalization, and shorten experimental run times while preserving sensitivity to the Be TSL. Here, three premoderated configurations, encapsulation, slab, and interstitial, were evaluated using MCNP6.3® simulations with ENDF/B-VIII.0 nuclear data. The effects of the Be TSL was quantified by comparing decay constants from simulations with and without 𝑆⁡(𝛼,𝛽) treatments for both beryllium and HDPE. The results show that premoderation enables the use of as little as 5.20% of the beryllium volume required for an unreflected geometry, achieving uncertainties below 0.5% and reducing run time by up to 72.2% in the studied configurations. The interstitial configuration achieved the best balance between low statistical uncertainty and high TSL sensitivity, outperforming the slab and encapsulation designs. These findings demonstrate that carefully optimized premoderation can significantly enhance the efficiency, feasibility, and precision of PNDA experiments for Be TSL validation.

Integral Experiments↗
DOE OSTI2021

Technical Documents for Gateway for Accelerated Innovation in Nuclear (GAIN)

The Gateway for Accelerated Innovation in Nuclear (GAIN) provides the nuclear energy community with access to the technical, regulatory, and financial support necessary to move new or advanced nuclear reactor designs toward commercialization. GAIN provides the nuclear community with a single point of access to the broad range of capabilities (i.e., people, facilities, materials, and data) across the U.S. Department of Energy (DOE) complex and its National Lab capabilities. The Fast Flux Test Facility (FFTF) is the most recent liquid metal reactor (LMR) to be designed, constructed, and operated by DOE. The 400-MWt sodium-cooled, fast-neutron flux reactor plant was designed for irradiation testing of nuclear reactor fuels and materials for liquid metal fast breeder reactors. Following the demise of the breeder reactor program in the United States, FFTF continued to play a key role in providing a test bed for demonstrating performance of advanced fuel designs and demonstrating operation, maintenance, and safety of advanced liquid metal reactors. The FFTF Program provides valuable information for potential follow-on reactor projects in the areas of plant system and component design, component fabrication, fuel design and performance, reactor control, prototype testing, and site construction. This report provides documents related to three important aspects of FFTF design and operation: 1) irradiation behavior of structural alloys and absorber materials, 2) thermohydraulics of rod bundles (i.e., coolant mixing), and 3) natural circulation heat transfer in the areas of modeling and validation. These technical documents are believed to be of interest to the nuclear industry and in particular to designers of new liquid metal reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

FFTF Acceptance and Startup Testing for GAIN

The Fast Flux Test Facility (FFTF) is the most recent liquid metal reactor (LMR) to be designed, constructed, and operated by the U.S. Department of Energy (DOE). The 400-MWt sodium-cooled, fast-neutron flux reactor plant was designed for irradiation testing of nuclear reactor fuels and materials for liquid metal fast breeder reactors. Following the demise of the breeder reactor program in the United States, FFTF continued to play a key role in providing a test bed for demonstrating performance of advanced fuel designs and demonstrating operation, maintenance, and safety of advanced liquid metal reactors. FFTF operations ceased in April 1992 after a determination by DOE that no combination of proposed missions was financially feasible over a ten-year period. The reactor is currently deactivated and in a long-term surveillance and maintenance (S&M) mode. This report provides information on the extensive and rigorous process that was used to conduct turnover from construction followed by acceptance and startup testing of the FFTF. This paper is in support of the Gateway for Accelerated Innovation in Nuclear (GAIN), which provides the nuclear energy community with access to the technical, regulatory, and financial support necessary to move new or advanced nuclear reactor designs toward commercialization while ensuring the continued safe, reliable, and economic operation of the existing nuclear fleet. The information obtained from the design, startup, and operation of the FFTF provides valuable insight for follow-on reactor projects, such as the Versatile Test Reactor (VTR), in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, reactor startup and operations, and reactor deactivation and decommissioning (D&D). The focus of this report is on the process used to startup the FFTF and to ensure that operations could be conducted efficiently and safely. A reference section is provided of documents detailing the successful turnover and testing process implemented for startup of the reactor and its supporting systems. The documents listed can be retrieved upon request and are believed useful for future reactor startup endeavors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Assessing the Impact of Large Removable Beryllium Reflector Experiments on HFIR Performance and Safety Metrics

This study investigated the effect of various removable beryllium (RB) experiment configurations on High Flux Isotope Reactor (HFIR) metrics to address increasing interest in these facilities for materials and fuels irradiation research. The RB reflector contains eight large and four small irradiation experiment facilities that offer excellent neutron flux conditions to perform fission and fusion reactor materials and fuels irradiation research. This work aims to outline acceptable RB configurations based on safety, performance, and programmatic metrics. This study assessed experiment effects on reactivity, cycle length, fission rate density distributions, and neutron flux distributions using the Shift, HFIRCON, and SCALE ORIGEN codes. Initial evaluations focused on generic experiment materials including aluminum plugs, stainless steel plugs, molybdenum plugs, and aluminum plugs with gadolinium shields. Subsequent analyses of MiniFuel experiments were performed to evaluate a heterogenous experiment, consisting of a more complex geometry and bearing several materials, and to test the correlations developed with the generic materials on a real experiment. Furthermore, the effects of neutron poison concentrations in the standard beryllium plugs were evaluated. When assuming a reference RB configuration with eight large fresh beryllium plugs, perturbed configurations with three aluminum plugs, one stainless steel plug, one molybdenum plug, one gadolinium-shielded aluminum plug, and one MiniFuel experiment resulted in a cycle length reduction of less than 1.3 days, the current threshold before requiring additional approvals. Configurations with five aluminum plugs, one stainless steel plug, one molybdenum plug, one gadolinium-shielded aluminum plug, and two MiniFuel experiments meet the 1.3 day limit if irradiated beryllium plugs are considered. Fuel element fission rate density distributions remained within safety limits, with maximum local increases under 9%. Neutron flux calculations revealed large thermal flux depressions inside and around the perturbed RB facilities, while epithermal and fast neutron fluxes increased because of reduced neutron moderation by the perturbed materials. These findings provide valuable guidance for optimizing RB configurations to balance safety and performance at HFIR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

MC-Lite: Development of a new lightweight multiplicity counter

This report details the development of a neutron multiplicity counter based on lithium doped plastic scintillators. This system has the capability to measure and discriminate fast neutrons, thermal neutrons, and gamma-rays allowing for multi-particle correlations in one device. The system was built and tested at Lawrence Livermore National Laboratory with Cf-252 in both bare configurations and surrounded by polyethylene and compared against the MC-15 multiplicity counter. Additionally, the detector was also placed outside of a subcritical assembly and demonstrated the ability to use correlated gamma-rays as a probe on the multiplication of the item.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Distribution of Neutrons in a Cold Lucite Moderator

A series of experiments was performed in 2010 to study the behavior of neutrons in a cold Lucite moderator. The diffusion of neutrons in a cold moderator is described as a two-step process. In the first step, the fast neutrons produced by a neutron source collide with the nuclei of Lucite, losing energy until they reach thermal energies. In the second step, the thermal neutrons continue to diffuse without any significant loss of energy until they are absorbed or leak out of the system. However, because the Lucite moderator is at low temperature, there is less up scattering in the neutron energy compared to a Lucite moderator at room temperature. Experiments were performed to study the behavior of neutrons and their distribution in a cold Lucite moderator and their results compared to simulations. Based on these measurements, an estimation of the absorption cross section at -78.5 °C yielded a value of 0.54 barns ± 0.02 barns.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗