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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Development of a Fast Spectrum Test Loop at the Advanced Test Reactor

The United States currently has no infrastructure to provide a fast neutron flux for research purposes. With the increased interest in Sodium Fast Reactor (SFR) technology, this type of infrastructure is necessary for further development of key components such as Versatile Test Reactor. The purpose of this research is to characterize the thermal hydraulics of a test loop capable of providing SFR like operating conditions. This includes showcasing feasibility in mechanical design, reaching a high fast to thermal neutron flux ratio ideally near 15, an inlet temperature of 600 K, a temperature increase of 50 K from inlet to outlet, and test chamber radial temperature profiles similar to SFR fuel rods. Existing infrastructure must be used to provide the neutron flux for the test loop, and it was decided that the Advanced Test Reactor (ATR) was the optimal choice due to its availability and previous work conducted similar to this project. To achieve the goals of this project, several mechanical designs were put forward with the most feasible design being chosen. The neutronics and thermal hydraulics analysis were conducted in tandem, with an iterative process used to maximize efficiency of both systems. Analysis of the neutronics system was done using Attila and MCNP while the thermal hydraulic analysis was done using RELAP5. This document outlines the thermal hydraulic analysis, though it includes the results from the mechanical design and neutronics analysis. The loop design includes a primary system with molten sodium, a set of “booster fuel plates” to increase the fast flux within the test section, a water system to cool the booster fuel plates, a helium system to separate the sodium and water systems, and a thermal neutron shield to decrease the thermal flux. Each of these systems is modeled in the neutronics and thermal hydraulic analyses. The RELAP5 model showcased the feasibility of the design proposed by this report. For all designs tested, the desired 50 K temperature increase from inlet to outlet was reached. The fuel rod radial temperature profile was lower than was expected, likely due to a lower rod thermal power than expected. A parametric study was also conducted to characterize the system with three fuel pins. This study showed the effect of thermal power, fluid pressure, fluid velocity, and inlet temperature on the system. ©Copyright by Anton Higgins December 11, 2019 All Rights Reserved

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2024

Godiva IV Simulated Radiation Field Characterization and Variance Reduction

Godiva IV is a system comprised of highly enriched uranium alloyed with molybdenum in the form of fuel plate rings. The reactor, along with its predecessors, was designed with the unique ability to satisfy interests in the super-prompt-critical reactor operation space. Originally, the reactor was part of the Los Alamos Critical Experiments Facility (LACEF) at Technical Area-18 (TA-18). The radiation field around Godiva at this facility was well characterized and understood. As a fast neutron system, the neutron spectrum in and around Godiva was close to a Watt Fission spectrum. The Kiva where Godiva IV was located at LACEF was made of thin, sheet metal walls which did not contribute significantly to the neutron spectrum. Following the transition of LACEF to the National Critical Experiments and Research Center (NCERC) in Nevada, Godiva-IV was moved from TA-18 to the Device Assembly Facility (DAF) at the Nevada National Security Site (NNSS). Part of this move brought renewed interest in radiation field characterization. The new facility introduced significant changes to the environment surrounding Godiva, and preliminary foil irradiation results suggested that the room contribution to the neutron spectrum was significant. Unlike at TA-18, a large thermal neutron signature was added to the fast spectrum from Godiva due to significant room return. A primary goal due to the additional complexity that the room return adds to the Godiva IV radiation emission spectrum was the development of an efficient Monte Carlo N-Particle (MCNP) calculation capable of characterizing the neutron spectrum anywhere in the room around Godiva. A campaign of activation foil irradiations and analysis were completed to support the validation of the MCNP model. The modeling of these foils in MCNP can be easily done with a standard volumetric neutron flux tally. However, given the multitude of locations and reaction rates to be modeled, further steps must be taken to increase the efficiency of these calculations in MCNP. During this study, a benchmark model currently under development for Godiva IV was used. A qualitative assessment of the thermal neutron contributors was performed using spatial neutron distribution plots. Additional detail was added to the model based on the qualitative results showing the thermal spectrum’s large sensitivity to hydrogenous material. Neutron energy spectra was evaluated at discrete locations in the room around Godiva to quantify the relative contribution of various components. It was discovered that the concrete walls are the largest contributor to the thermal signature, with minor contributions from plastic components surrounding Godiva. Following these results, two different variance reduction techniques were implemented to improve the problem efficiency in these calculations. In the first approach, an F5 point detector tally was implemented in the standard Godiva IV criticality problem. The second approach involved a weight-window generator implementation with an F5 point detector tally in a fixed source problem. The weight window implementation reduced the runtime from 42739.55 minutes to 1803.34 minutes (computer time), compared to the F5 KCODE implementation.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Innovative Nuclear Materials Outboard-A Project Specimen Preparation Guide

The Innovative Nuclear Materials (INM) Program was recently established by the U.S. Department of Energy (DOE) to develop advanced material technologies for use in nuclear reactors. The INM program is presently focused on researching in-core non-fueled materials for application in fast spectrum nuclear reactors. The widespread deployment of fast reactors continues to be a prominent aspiration for advanced nuclear technology developers. However, companies working to license these reactors have no choice but to rely on historic material technologies since further optimization and advancement of these materials is impeded by the lack of fast neutron irradiation test facilities. INM-OA is a non-fueled drop-in experiment which will irradiate material specimens of interest to fast reactor applications. This experiment will be irradiated at Idaho National Laboratory (INL) in the Advanced Test Reactor (ATR) outboard-A (OA) position during normal and high temperature steady state (HTSS) cycles. It will include material specimens supplied by members of the INM program and will utilize a cadmium-lined basket to filter out incident thermal neutrons, thus simulating a faster neutron energy spectrum. Material specimens will undergo post-irradiation examination including microscopy and mechanical testing. In addition to absorption reactions, fast neutrons cause microstructural damage in materials by atom displacement, which can cause exacerbated changes in physical properties and behavior. Thus, the data obtained from the INM-OA experiment will be crucial for understanding the engineering-scale behavior of reactor materials. This document is intended for the Principal Investigators providing samples for this project. Topics included are a general description of the experiment, the irradiation experiment/capsule design, sample geometries, number of samples to be provided, documentation to be provided, a brief list potentially useful characterization methods that can be leveraged at INL, and other miscellaneous requirements specific to this project. This document is intended for informational use only.

innovative nuclear materials↗
DOE OSTI2022

Determination of energy correction coefficient for neutron albedo dosimeter. Three approaches

To accurately determine dose, albedo TLD dosimeters require application of a neutron energy correction factor (NECF). Here, in this paper, the results from three different methods were used to determine NECFs. In the first method, NECF values were calculated as a ratio of the known “conventional true” dose and the neutron dose reported by albedo dosimeter(s) irradiated on a phantom. The results from this direct method were utilized for verification of two other methods. The second method used a specially designed device called the Neutron Area Monitor (NAM) model 5. The third method of NECF determination was based on the response of a BF 3 tube under different levels of moderation. The results of this work showed that the Navy-developed Neutron Area Monitor (NAM) model 5 measures NECF values correctly when irradiated with 252 Cf (bare and D 2 O-moderated). This was an important result, because it validated the use of the NAM-5 in nuclear power and marine propulsion reactors, as their neutron energy spectrum is approximated by the field of a Cadmium-coated 30 cm diameter steel sphere of D 2 O moderated 252 Cf. In the case of a broad neutron energy spectrum the 6 Li response to the small amount of slow neutrons outweighs its response to the fast neutrons overwhelmingly responsible for the neutron dose rate. Based on the obtained results we can conclude that NAM-5 provides correct NECF determination if the fraction of slow neutrons in the neutron dose rate is more than 20%. A satisfactory correlation between the NECF determined by NAM-5 and the multi-moderated BF3 tube method was observed, but the multi-moderated BF 3 tube values were ~30% lower than those obtained by the NAM-5 at the same locations for broader neutron energy spectrum locations.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Comparative scintillation performance of EJ-309, EJ-276, and a novel organic glass

An organic glass scintillator developed by Sandia National Laboratories was characterized in terms of its light output and pulse shape discrimination (PSD) properties and compared to commercial liquid (EJ-309) and plastic (EJ-276) organic scintillators. The electron light output was determined through relative comparison of the 137 Cs Compton edge location. The proton light yield was measured using a double time-of-flight technique at the 88-Inch Cyclotron at Lawrence Berkeley National Laboratory. Using a tunable broad-spectrum neutron source and an array of pulse-shape-discriminating observation scintillators, a continuous measurement of the proton light yield was performed for EJ-309 (200 keV–3.2 MeV), EJ-276 (170 keV–4.9 MeV), and the organic glass (50 keV–20 MeV) . Finally, the PSD properties of the organic glass, EJ-309, and EJ-276 were evaluated using an AmBe source and compared via a figure-of-merit metric. The organic glass exhibited a higher electron light output than both EJ-309 and EJ-276. Its proton light yield and PSD performance were comparable to EJ-309 and superior to that of EJ-276. Finally, with these performance characteristics, the organic glass scintillator is well poised to replace current state-of-the-art PSD-capable scintillators in a range of fast neutron detection applications.

47 OTHER INSTRUMENTATION↗
DOE OSTI2025

Comparative analysis of new crystal and plastic scintillators for fast and thermal neutron detection

This paper considers scintillation and physical properties of efficient organic scintillators tested as single crystals and as components of plastics with pulse shape discrimination (PSD). For the first time, single crystals of 9,9-dimethyl-2-phenyl-9H-fluorene (PhF) were grown for studies of the basic scintillation properties of this new compound. Comparison to classical organic crystals, like anthracene, trans-stilbene, and p-terphenyl, and to more recently introduced organic glass showed that the new crystal belongs to a group of the most efficient scintillators, with light output exceeding that of trans-stilbene and PSD comparable to that of p-terphenyl. Furthermore, additional studies were conducted to evaluate scintillation performance and physical properties, like hardness and dye leaching, of plastic scintillators prepared with PhF, organic glass, and liquid diisopropylnaphthalene (DIPN) that were considered as examples for potential replacement of PPO (2,5-diphenyloxazole) in current commercial PSD plastic scintillators. Comparison of the highest performing plastic scintillators prepared with these dyes shows that PhF formulations produced a record light output (LO) increase of 69 % relative to EJ-200. Similar improvements obtained with 6 Li-loaded formulations showed that future development of PSD plastics should not be limited to use of PPO but must involve the search and exploration of new efficient dyes that may lead to discovery of much brighter organic scintillators with improved physical properties required for fast and thermal neutron detection, fast neutron spectroscopy, and antineutrino detection applications.

9,9-dimethyl-2-phenyl-9H-fluorene↗
DOE OSTI2020

Compound pulse characteristics of a heterogeneous composite scintillator in a gamma-ray field

An important aspect of radiation detection in scintillators is the time dependence of emitted fluorescence, which results in characteristic measured pulse shape. In some materials, a relatively strong dependence of the scintillation pulse shape on stopping power exists, which can provide reliable identification of the incident particle. An alternative method to realize particle identification is by combining materials with different fluorescence lifetimes into heterogeneous structures. Such composite scintillators derive their properties from both the constituent material characteristics and the geometry of the composite structure. In composite scintillators with a high degree of heterogeneous loading, a superposition of fluorescence contributions originating in multiple materials, which is observed as single waveform with an intermediate pulse shape, may affect its ability to reject rays. We measure the time dependence of light output from a scintillator composed of 6 Li-containing glass shards and scintillating polyvinyl toluene and identify events that exhibit such compound behavior when exposed to rays and fast neutrons. We develop a modeling and simulation framework that reproduces the pulse shapes in heterogeneous scintillators and use it study the effect that the weight percentage of 6 Li glass has on rejection. The modeling framework is applied to the experimentally studied scintillator, finding a good agreement. Here, the developed modeling and simulation approach will help optimize the design of heterogeneous scintillators to meet the desired trade-o between neutron capture efficiency and -ray rejection.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Novel imaging technique for thermal neutrons using a fast optical camera

A novel imaging technique for thermal neutrons using a fast optical camera is presented. Thermal neutrons are reacted with 6 Lithium to produce a pair of 2.73 MeV tritium and 2.05 MeV alpha particles, which in turn interact in a thin layer of LYSO crystal scintillator to produce a localized flash of light. These photons are directed by a pair of lenses to a micro-channel plate intensifier, and its output is connected to the optical camera, TPX3CAM. The results from the camera are reconstructed through a custom algorithm. Various cutting parameters were found through data analysis to eliminate the background, and they were shown effective in matching the simulated rate of the neutron source. The system is fast with 40 ns decay time and allows free-space light collection, both vastly enhances flexibility of neutron detection.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

An Artificial Neural Network System for Photon-Based Active Interrogation Applications

Active interrogation (AI) is a promising technique to detect shielded special nuclear materials (SNMs). At the University of Michigan, we are developing a photon-based AI system that uses bremsstrahlung radiation from an electron linear accelerator (linac) as an ionizing source and stilbene organic scintillating detectors for neutron detection. Stilbene scintillators are sensitive to fast neutrons and photons and have excellent pulse shape discrimination (PSD) capabilities. The traditional charge integration (CI) method commonly used for PSD analysis eliminates piled-up pulses and relies on a particle discrimination line to separate neutrons and photons. The presence of the intense photon flux during AI creates a significant number of piled-up events in the stilbene scintillator, thereby posing a great challenge to the traditional CI method. Identifying true single neutron pulses becomes challenging due to the presence of a pile-up cloud and overlapping neutron, photon and pile-up clouds in the PSD analysis. To mitigate the effect of pulse pile up and identify true single neutron pulses from stilbene scintillators, an artificial neural network (ANN) system is developed. The developed ANN system identifies single neutron pulses and neutron-photon combinations from piled-up events. The results obtained from a 252Cf measurement in the presence of the intense photon flux show that the developed ANN system outperforms the traditional CI method. Since many piled-up events lie above the particle discrimination line, they get misclassified as neutrons by the traditional CI method resulting in 25% overestimation of the net neutron count rate during the linac pulse. The overall net neutron count rate (single and restored neutrons) during the linac pulse, estimated by the ANN system is 60% of the ground truth. Energy spectroscopy of the ANN attributed single neutron pulses further provides evidence on the detection of prompt fission neutrons from the 252Cf fission source.

42 ENGINEERING↗
DOE OSTI2025

Neutron Yield of Thermo Scientific P385 D-T Neutron Generator vs . Current and Voltage

The Thermo Scientific P385 Neutron Generator is a compact neutron source, producing 14 MeV neutrons through the deuterium-tritium (DT) fusion reaction. It is important to measure and understand the dependence of the neutron production rate on the accelerator current and voltage. In this study we evaluated neutron production with an absolutely calibrated liquid scintillator neutron spectrometer (BTI N-Probe), an absolutely calibrated He-3 detector surrounded by HDPE shells (Detec Nested Neutron Spectrometer, NNS), and two uncalibrated ZnS fast neutron scintillators (EJ-410), for both A3082 and A3083 sealed tubes. Here we also modeled the neutron yield using the TRIM code, which calculates the trajectory and the energy loss of deuterons and tritons within the target. Experimental results showed an essentially linear dependence on beam current, as expected. A 3.59 ±0.08 power law dependence on the operating voltage was measured, in effective agreement with the modeled value of 3.5. A series of absolute NNS and N-Probe measurements, matched against MCNP calculations, showed that the A3083 and A3082 tubes provide a maximum neutron yield of 8.2 × 10 8 n/s and 4.7 × 10 8 n/s respectively, with estimated uncertainty of ±10%.We showed, through modeling, that tritium decay is not a significant consideration for tubes, such as these, with lifetimes of less than 10 years.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Electrochemically modulated molten salt reactor

An electrochemically modulated molten salt reactor (EMMSR) that contains a vessel and a power source. The vessel houses a fuel salt, at least a portion of a neutron moderator, and at least a portion of an insulator. The fuel salt includes enough dissolved fissile isotopes to cause continued self-sustaining fission reactions during the operation of the EMMSR. The neutron moderator is configured to slow down fast neutrons produced by the dissolved fissile isotopes. The insulator is configured to electrically isolate the neutron moderator from the vessel. The power source has a positive potential and a negative potential. The positive potential is received by the neutron moderator and the negative potential is received by the vessel.

Jiang, Junhua↗
DOE OSTI2021

Direct current response of a thin scCVD diamond detector under increased applied field to 14.1 MeV neutrons

An avalanche effect yielding inherent gain can be exploited in thin, single-crystal chemical vapor deposition (scCVD) diamond. It occurs when a high enough bias is applied across the diamond thickness while avoiding breakdown. This charge multiplication effect was studied previously with alpha particles and heavy ions either by using the transient current technique or by measuring the energy spectrum. The measurements we obtained to evaluate the charge multiplication performance of a 10 μm thick scCVD diamond detector used a novel approach—we employed an electrometer to characterize the response of the detector by performing directly coupled current measurements (time-averaged charge, at 1 Hz sampling) when exposed to 14.1 MeV neutrons from deuterium-tritium fusion. We measured both the dark and irradiated currents from the detector over a range of applied displacement field values from 2 to 75 V/μm. A histogram method with central mean and standard deviation width was used to determine the current over each measurement duration typically from 100 to 300 seconds. The dark-subtracted irradiated current (i.e., contrast) was used to evaluate the gain of the detector at each applied displacement field. The contrast at an applied displacement field between 15 and 20 V/μm was higher than the expected linear increase in contrast proportional to the increased applied bias, indicating the possible presence of avalanche events in the diamond. The detector response also indicated possible polarization and charge depletion effects. These results provide an opportunity to further explore the use of thin scCVD diamond as a fast neutron current mode detector with inherent gain.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Response of a high-pressure 4 He scintillation detector to nuclear recoils up to 9 MeV

Helium-4-based scintillation detector technology is emerging as a strong alternative to pulse-shape discrimination-capable organic scintillators for fast neutron detection and spectroscopy, particularly in extreme gamma-ray environments. The 4 He detector is intrinsically insensitive to gamma radiation, as it has a relatively low cross-section for gamma-ray interactions, and the stopping power of electrons in the 4 He medium is low compared to that of 4 He recoil nuclei. Consequently, gamma rays can be discriminated by simple energy deposition thresholding instead of the more complex pulse shape analysis. The energy resolution of 4 He scintillation detectors has not yet been well-characterized over a broad range of energy depositions, which limits the ability to deconvolve the source spectra. In this work, an experiment was performed to characterize the response of an Arktis S670 4 He detector to nuclear recoils up to 9 MeV. The 4 He detector was positioned in the center of a semicircular array of organic scintillation detectors operated in coincidence. Deuterium–deuterium and deuterium–tritium neutron generators provided monoenergetic neutrons, yielding geometrically constrained nuclear recoils ranging from 0.0925 to 8.87 MeV. The detector response provides evidence for scintillation linearity beyond the previously reported energy range. Finally, the measured response was used to develop an energy resolution function applicable to this energy range for use in high-fidelity detector simulations needed by future applications.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Generalized method for the optimization of pulse shape discrimination parameters

Organic scintillators exhibit fast timing, high detection efficiency for fast neutrons and pulse shape discrimination (PSD) capability. PSD is essential in mixed radiation fields, where different types of radiation need to be detected and discriminated. In neutron measurements for nuclear security and non proliferation effective PSD is crucial, because a weak neutron signature needs to be detected in the presence of a strong gamma-ray background. Here, the most commonly used deterministic PSD technique is charge integration (CI). This method requires the optimization of specific parameters to obtain the best gamma-neutron separation. These parameters depend on the scintillating material and light readout device and typically require a lengthy optimization process and a calibration reference measurement with a mixed source. In this paper, we propose a new method based on the scintillation fluorescence physics that enables to find the optimum PSD integration gates using only a gamma-ray emitter. We demonstrate our method using three organic scintillation detectors: deuterated trans-stilbene, small-molecule organic glass, and EJ-309. In all the investigated cases, our method allowed finding the optimum PSD CI parameters without the need of iterative optimization.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Deployment of a double scatter system for directional detection of background neutron radiation

The detection of special nuclear materials (SNM) requires the understanding of nuclear signatures that allow the discrimination against background. In particular, understanding neutron background characteristics such as count rates and energies and their correlations with environmental conditions and surroundings of measurement locations is important in enhancing SNM detection capabilities. The Mobile Imager of Neutrons for Emergency Responders (MINER) was deployed for 8 weeks in downtown San Francisco (CA) to study such neutron background characteristics in an urban environment. Of specific interest was the investigation of the impact of surrounding buildings on the neutron background count rates and to answer the question whether buildings act as absorber of neutrons or as sources via the so-called ship effect. MINER consists of 16 liquid scintillator detector elements and can be operated as a neutron spectrometer, as a neutron imager, or simply as a counter of fast neutrons. As expected, the neutron background rate was found to be inversely proportional to the atmospheric pressure. In the energy range where MINER is most sensitive, approximately 1–10 MeV, it was found that the shape of the detected background spectrum is similar to that of a detected fission spectrum, indicating the limited discrimination power of the neutron energy. The similarities between the detected background neutron spectrum and fission sources makes it difficult to discriminate SNM from background based solely on the energies observed. The images produced using maximum likelihood expectation maximization revealed that neutrons preferentially are coming from areas in the environment that have open sky, indicating that the surrounding buildings act as absorbers of neutrons rather than sources as expected by the ship effect. Here, the inherent properties of a neutron scatter camera limit the achievable image quality and the effective deployment to systematically map neutron background signatures due to the low count rate.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Front-End Design for SiPM-Based Monolithic Neutron Double Scatter Imagers

Neutron double scatter imaging exploits the kinematics of neutron elastic scattering to enable emission imaging of neutron sources. Due to the relatively low coincidence detection efficiency of fast neutrons in organic scintillator arrays, imaging efficiency for double scatter cameras can also be low. One method to realize significant gains in neutron coincidence detection efficiency is to develop neutron double scatter detectors which employ monolithic blocks of organic scintillator, instrumented with photosensor arrays on multiple faces to enable 3D position and multi-interaction time pickoff. Silicon photomultipliers (SiPMs) have several advantageous characteristics for this approach, including high photon detection efficiency (PDE), good single photon time resolution (SPTR), high gain that translates to single photon counting capabilities, and ability to be tiled into large arrays with high packing fraction and photosensitive area fill factor. However, they also have a tradeoff in high uncorrelated and correlated noise rates (dark counts from thermionic emissions and optical photon crosstalk generated during avalanche) which may complicate event positioning algorithms. We have evaluated the noise characteristics and SPTR of Hamamatsu S13360-6075 SiPMs with low noise, fast electronic readout for integration into a monolithic neutron scatter camera prototype. The sensors and electronic readout were implemented in a small-scale prototype detector in order to estimate expected noise performance for a monolithic neutron scatter camera and perform proof-of-concept measurements for scintillation photon counting and three-dimensional event positioning.

47 OTHER INSTRUMENTATION↗
DOE OSTI2026

Comparative neutron-irradiation effects on thermal conductivity degradation and dimensional stability of TiC, TiB 2 , and ZrB 2 at 200–1000 °C

Ultra-high-temperature ceramics (UHTCs), including TiC, Ti 11 B₂, and Zr 11 B₂, show great potential for plasma-facing components due to their excellent high-temperature properties prior to irradiation. However, their response to neutron irradiation remains insufficiently understood, limiting robust assessment of their viability for fusion energy applications. Here, this study examines the thermal conductivity, dimensional stability and microstructure of TiC, TiB₂, and ZrB₂ following neutron irradiation at temperatures of 200–1000 °C and fast neutron fluences of 2.0 × 10 25 to 1.1 × 10 26 n/m 2 (E > 0.1 MeV). Lattice swelling measured by synchrotron X-ray diffraction in all three UHTCs was maximized at 200 °C and decreased with increasing irradiation temperature, with no evidence of amorphization observed at 200 °C. Above 600 °C, significant macroscopic volume swelling was observed in irradiated Ti 11 B₂ and Zr 11 B₂, but not in TiC, likely due to cavity formation in the diborides. The post-irradiation thermal conductivity, measured at the irradiation temperature, ranged from 28 to 45 W/m·K, representing a 34–45% reduction relative to the unirradiated material. Notably, neutron-irradiated UHTCs exhibit recoverable thermal conductivity at elevated temperatures, comparable to ferritic–martensitic steels and potentially superior to W when transmutation effects are considered, highlighting promise for shielding or armor plasma-facing components. At 600 °C, both thermal conductivity degradation and lattice swelling saturated at doses exceeding 2–4 dpa.

fusion materials↗
DOE OSTI2024

TEX-Cl: Integral Experiment Execution of Thermal/Epithermal eXperiments using Highly Enriched Uranium with Polyethylene and Chloride Absorbers

This report documents the experimental configurations and measurements performed for IER-499, TEX with chlorine (TEX-Cl). TEX-Cl is a variant of the TEX-HEU campaign and utilizes highly enriched uranium fuel, high-density polyethylene (HDPE) moderators and reflectors, and interstitial NaCl salt absorbers. These experiments probe the chlorine (from the sodium chloride) absorption cross section in the thermal and intermediate neutron energy regimes, with a small portion in the fast neutron energy regime. Three configurations were selected by maximizing the sensitivity in k eff to the needs of Y-12 for their electrorefining operations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗