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At least 379 records · Page 21

Rossi-α Measurements on a System Consisting of Weapons Grade Plutonium Moderated by Polyethylene [Poster]

A series of experiments nicknamed "TEX," standing for Thermal-Epithermal eXperiments, were performed at the National Criticality Experiments Research Center (NCERC) on the Planet critical assembly. The TEX experiment campaigns are a collaboration between Los Alamos National Laboratory (LANL) and Lawrence Livermore National Laboratory (LLNL) and funded by the Nuclear Criticality Safety Program (NCSP). The TEX experiment consists of several campaigns; this work specifically discusses measurements made on one series of TEX experiments designed to examine the thermal scattering law (TSL) for polyethylene. This series consisted of two different critical configurations which contain plutonium fuel. The plutonium plates are 2 in. by 3 in. by 0.125 in. Zero Power Physics Reactor (ZPPR) plates and are placed in a tray in an array of 6 plates by 4 plates wide creating a 12 in. by 12 in. square. These trays of plutonium are interleaved between polyethylene plates. These measurements differ from previous TEX experiments in that they examine a more thermal neutron spectrum than previously examined by interleaving much thicker interstitial moderating plates than the previous measurements. The two configurations differ in the thickness of polyethylene between the fissile material. While the experiment was designed to examine the TSL, this work only examines the results of Rossi-α measurements on the two different configurations. For both configurations, Rossi-α measurements were performed to experimentally validate the neutron spectrum.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Sulfur Isotopes in Gas-rich Impact-Melt Glasses in Shergottites

Large impact melt glasses in some shergottites contain huge amounts of Martian atmospheric gases and they are known as gas-rich impact-melt (GRIM) glasses. By studying the neutron-induced isotopic deficits and excesses in Sm-149 and Sm-150 isotopes resulting from Sm-149 (n,gamma) 150Sm reaction and 80Kr excesses produced by Br-79 (n,gamma) Kr-80 reaction in the GRIM glasses using mass-spectrometric techniques, it was shown that these glasses in shergottites EET79001 and Shergotty contain regolith materials irradiated by a thermal neutron fluence of approx.10(exp 15) n/sq cm near Martian surface. Also, it was shown that these glasses contain varying amounts of sulfates and sulfides based on the release patterns of SO2 (sulfate) and H2S (sulfide) using stepwise-heating mass-spectrometric techniques. Furthermore, EMPA and FE-SEM studies in basaltic-shergottite GRIM glasses EET79001, LithB (,507& ,69), Shergotty (DBS I &II), Zagami (,992 & ,994) showed positive correlation between FeO and "SO3" (sulfide + sulfate), whereas those belonging to olivine-phyric shergottites EET79001, LithA (,506, & ,77) showed positive correlation between CaO/Al2O3 and "SO3".

Rao, M. N.↗

Resolving the Discontinuity Between Thermal Scattering Data and Fast Data

The task of resolving discontinuities between thermal cross sections and fast cross sections in AMPX processing is part of a broader project to provide guidance on the potential effect of unknown thermal neutron scattering law (TNSL) data, primarily from the standpoint of criticality safety. The primary goal of this work is to deliver code fixes in AMPX that increase confidence that SCALE libraries provide physically correct cross sections across the full range of neutron energies encountered in transport applications. This report discusses the resolution of two sources of discontinuity. The first issue occurs when the Bragg edges in an evaluated nuclear data file do not extend to 5 eV. The second issue concerns parameters chosen during the processing of the nuclear data library—specifically, the parameter that defines the energy at which the thermal data ends and the fast nuclide data begins. This report also discusses a third issue, which is related to the calculation of probability distributions from a densely gridded double-differential cross section.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Distinguishing fissile uranium isotopes using an active well neutron coincidence counter

Proposed thorium-based nuclear fuel cycles are likely to require quantification and verification of 233 U within nuclear material. Because of their similar fission cross sections, active neutron nondestructive assay (NDA) systems may respond similarly to 233 U and 235 U. Traditional safeguards equipment has been optimized for 235 U and 238 U quantification associated with conventional uranium/plutonium fuel cycles and may not be directly applicable to 233 U quantification when mixed with other actinides. This work used models of the large volume active well coincidence counter (LV-AWCC) at Oak Ridge National Laboratory to evaluate the performance of this neutron NDA system to differentiate fissile uranium isotopes. The models were developed to simulate NDA system performance in response to a number of triangular radiation signature training device sources within the central cavity or well. This work predicted that the LV-AWCC can effectively differentiate 233 U from 235 U in certain modes of operation. In active mode, the LV-AWCC with the cadmium liner results in different doubles count rates between the fissile isotopes for a given fissile uranium mass. Without the cadmium liner, the uranium isotopes provide a statistically indistinguishable doubles count rate response for the fissile masses considered in this work (up to approximately 150 g). The cadmium liner serves to harden the neutron interrogation spectrum, which better exploits the notable difference in the 233 U and 235 U fission cross sections at approximately 1 eV. In passive mode, the two fissile isotopes exhibit different doubles and singles count rates regardless of liner presence because the passive source strength of 233 U is approximately 2 orders of magnitude stronger than that of 235 U due to the shorter half-life and correspondingly higher (α, n) yield. We conclude that using neutron interrogation in the LV-AWCC, two measurements are needed to quantify 233 U content in mixed uranium items. The first measurement is used to determine the total fissile uranium mass using a mode that cannot distinguish fissile isotopes (i.e., where a similar response is observed for both fissile uranium isotopes such as active doubles without cadmium or using a thermal neutron interrogation source). In conclusion, the second measurement is used to determine the 233 U content by using a differentiating technique (e.g., passive doubles, passive doubles to singles ratio, active doubles with cadmium).

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Multiple shape coexistence near {sup 118}Sn: First 0+ 3 lifetime measurement

The intruder bands in Sn isotopes, built on the 2p-2h excitation across the Z = 50 proton shell gap, are well-known examples of shape coexistence near the neutron mid-shell region. Spectroscopic signatures for shape coexistence include enhanced E0 transitions between the 0+ band heads. However, the underlying shape coexistence and mixing has been unclear because lifetime information for the excited 0+ states was incomplete in 118Sn. We thus present here the first measurement of the 0+ 3 lifetime in 118Sn using the fast-timing technique following thermal-neutron capture. The observed enhancement in ρ2(E0;0+ 3 → 0+ 2 ) of 150(30) milliunits provides compelling indications for multiple shape coexistence in 118Sn. Additionally, three distinct shapes in 116,118,120Sn naturally emerged in theoretical calculations based on the quantum-number-projected generator coordinate method employing a relativistic energy density functional.

Wu, F.↗

Observed Correlation Between Local Topography and Passive Neutron Measurements from the Dynamic Albedo of Neutrons (DAN) Instrument on the Mars Science Laboratory (MSL) Rover

The Dynamic Albedo of Neutrons (DAN) instrument on the Mars Science Laboratory Curiosity rover primarily measures neutrons that have undergone interactions with rocks and materials in the rover's local environment. As the rover ascends Aeolis Mons, it may encounter more extreme local topography (e.g., cliffs, gullies, canyons). We present three parts of the rover's traverse in which local topography, expressed as the average local relief relative to the rover, is moderately to strongly correlated with an increase in passive thermal neutron count rates. These increases in count rates are consistent with results from radiation transport models of the instrument's performance near simulated topographic features. Additional DAN measurements in areas of high average local relief (>0.25 m) within 5 m of the instrument could bolster this correlation. DAN's sensitivity to topography in its passive mode could be utilized as a new measurement capability and has implications for the operation of future landed missions carrying neutron spectrometers (e.g., VIPER, MoonRanger, Lunar-VISE).

Steven D Dibb↗

Analysis of Second Target Station Target Removal Dose Rates

The Second Target Station (STS) project at Oak Ridge National Laboratory’s spallation neutron source is a crucial initiative for maintaining U.S. leadership in neutron sciences. The STS aims to create the world’s brightest pulsed cold neutron source, enabling cutting-edge research across various scientific disciplines. To ensure safe and efficient maintenance operations, understanding the effects of shutdown dose rates from activated components within the STS target systems is essential. This study establishes a computational framework for calculating decay gamma sources and subsequent shutdown dose rates utilizing advanced methods to account for all activation channels, including high-energy interactions down to thermal neutron capture. This study describes a novel integration of multiple tools and provides an effective means of analyzing activation and shutdown dose rates at spallation neutron facilities. A custom-developed script automates the decay gamma source generation process, ensuring proper sampling during the variance reduction phase, which is critical for accurate predictions of shutdown dose rates.

Transmutation↗

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↗

Measured Thermal and Fast Neutron Fluence Rates ATR Cycles 171A

This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for ATR Cycle 171A SR which were measured by the Radiation Measurements Laboratory (RML) as requested by the Power Reactor Programs (ATR Experiments) Radiation Measurements Work Order.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Measured Thermal and Fast Neutron Fluence Rates ATR Cycles 171B.SR

This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for ATR Cycle 171B SR which were measured by the Radiation Measurements Laboratory (RML) as requested by the Power Reactor Programs (ATR Experiments) Radiation Measurements Work Order.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Power density in direct nuclear-pumped He-3 lasers

The interaction of neutron beams with He-3 gas is of interest for nuclear pumped lasers. The effects of spectral dependence of the neutron beam, neutron attenuation in the gas-filled laser tube, and transport of the charged-particle He-3(n, p)He-3 reaction products are treated in detail. An expression for the energy density as a function of position within the tube, tube radius, operating pressure, and neutron fluence is given. The maximum energy density within the optical cavity is achieved when the tube radius is 3.26/P where P (atm) is the operating pressure. The variation of radius by 50% above and below optimum will change the energy density at most by 10%, although performance degrades quickly for radii outside this range. If the optimum tube radius is used for each operating pressure, then the power density on the centerline (kW/cu cm) is given as 9.3 x 10 to the -18th power times the operating pressure times flux in a thermal neutron environment of flux (n/sq cm sec).

Wilson, J. W.↗

The neutron number probability distribution in coupled lumped assemblies

Here, the validity of the gamma distribution in describing the neutron number probability distribution function for both isolated and coupled multiplying assemblies when constrained to reproduce the true mean and variance is investigated in lumped geometry by numerical comparison with kinetic Monte Carlo simulations. The mean and variance are obtained from numerical solution of moment equations constructed from the relevant forward Master equation with assembly coupling coefficients obtained from a view factor method. Numerical results for a two-group, two coupled assemblies model, with static and dynamic reactivity insertion, show that except for subcritical assemblies, the gamma distribution well-approximates the number distribution. Differences in the fast and thermal neutron population shapes are explained in terms of effective source strengths due to downscatter and coupling.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

A sensitivity analysis to predict the neutronics behavior of samples irradiated in the VTR rabbit system

We report a low-order neutronics model is developed to carry out hundreds of simulations efficiently and investigate the neutronics behavior of samples being irradiated in a test reactor setting under different geometrical constraints. The low-order model allowed for simulations that yield the expected neutronics behavior of any irradiated sample in any environment and allows for the calculation of highly accurate spatially averaged statistics and idealized spatial distributions in the neutron flux. Several benchmarks are performed to evaluate the performance and limitations of the low-order model revealing many important findings. The low-order model predicted the LHGR in the EBR-II driver fuel to within 2.34% by only simulating the fuel rod by itself, which served as a validation for the model. Sensitivity studies investigated 3% enriched UO 2 and U-10Zr being irradiated in the Versatile Test Reactor rabbit system. The analyses investigated a range of combinations of 15 radii and 5 heights for each sample in the rabbit system. Similar data sets are also provided for irradiations in the Advanced Test Reactor’s B-10 irradiation position, which is a thermal neutron spectrum environment. Generalized fits and fit coefficients are obtained for sample heating, reaction rate densities, and local multiplication rate characteristics, allowing the predictions of the neutronics behavior of the samples based on their geometrical constraints. The analyses and fits laid the groundwork for developing a user-end Multiphysics analysis framework to assist and accelerate irradiation experiment design and optimization.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Boron‐polymer composites engineered for compression molding, foaming, and additive manufacturing

Abstract Boron (specifically 10 B) is the element of choice to shield thermal neutrons due to its large (n, α) cross‐section; however, very few polymer composites containing high boron concentrations are available. This study aimed to determine the maximum possible amount of boron that could be introduced into a polymer matrix. Diverse manufacturing techniques, ranging from additive manufacturing to compression molding, were employed to fabricate inks and filaments for 3D printing, foams, and flexible pads. Composites using siloxanes, poly(lactic acid), and acrylonitrile butadiene styrene containing up to 80 wt% boron were sucessufully fabricated. The addition of known plasticizers (polyethylene glycol) and reinforcing agents (carbon nanofibers and fumed silica) helped to overcome fabrication problems such as clogging of the printing nozzle or crumbling of compression molded parts. In addition, the thermal‐mechanical properties of these novel boron composites were determined and shown to vary according to boron concentration, presence of additives, and fabrication techniques utilized.

36 MATERIALS SCIENCE↗

Fabrication of Black Body Grids by Thick Film Printing for Quantitative Neutron Imaging

Neutron imaging offers deep penetration through many high-Z materials while also having high sensitivity to certain low-Z isotopes such as 1H, 6Li, and 10B. This unique combination of properties has made neutron imaging an attractive tool for a wide range of material science and engineering applications. However, measurements made by neutron imaging or tomography are generally qualitative in nature due to the inability of detectors to discriminate between neutrons which have been transmitted through the sample and neutrons which are scattered by the sample or within the detector. Recent works have demonstrated that deploying a grid of small black bodies (BBs) in front of the sample can allow for the scattered neutrons to be measured at the BB locations and subsequently subtracted from the total measured intensity to yield a quantitative transmission measurement. While this method can be very effective, factors such as the scale and composition of the sample, the beam divergence, and the resolution and construction of the detector may require optimization of the grid design to remove all measurement biases within a given experimental setup. Therefore, it is desirable to have a method by which BB grids may be rapidly and inexpensively produced such that they can easily be tailored to specific applications. In this work, we present a method for fabricating BB patterns by thick film printing of Gd2O3 and evaluate the performance with variation in feature size and number of print layers with cold and thermal neutrons.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

Irradiation Experiments and Thermal Analysis for Reactor System Design and Analysis at INL

Idaho National Laboratory (INL) is the nation's lead nuclear laboratory working to enhance reactor systems' safety, security, economics, and efficiency. Research and development (R&D) programs at INL support the current fleet of nuclear reactors for safer operation and newer reactors technology design, development, demonstration, and deployment. A major focus of INL's mission is the reactor system design and analysis supported by the irradiation experiments and thermal analysis of advanced and current-generation nuclear fuels and materials. The irradiation experiments and thermal analysis provide a deeper understanding of basic radiation damage processes that can determine the basis for performance improvements and verification of modeling assumptions. These experiments and analyses include experiment management: design, fabrication, characterization, irradiation, and post-irradiation examination. This research involves thermal, neutronic, and material investigation using the INL's Advanced Test Reactor (ATR) and Transient Reactor Test (TREAT) facilities. The results from these experiments and analysis are required to develop the regulatory basis for deploying new or modified fuels and materials. This seminar talk will also provide a general overview of the INL's research facilities, ongoing research programs, core capabilities, and opportunities for students and faculties.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Measured Thermal and Fast Neutron Fluence Rates ATR Cycles 167A

This report contains the thermal (2200 m/s) and fast (E>1 MeV) neutron fluence rate data for ATR Cycle 167A which were measure by the Radiation Measurements Laboratory (RML) as requested by the Power Reactor Programs (ATR Experiments) Radiation Measurement Works Order.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Final Design for Thermal/Epithermal eXperiments (TEX) with Chloride Absorbers to Provide Validation Benchmarks for Y-12 Electrorefining Facility

Uranium electrorefining operations at Y-12 require validation for chlorine absorption and reflection. Plutonium chloride solution operations at Los Alamos National Laboratory require chlorine absorption validation. The growing need for chlorine validation is evident to the wider criticality safety community, with multiple attendees at the recent TEX 2.0 meeting at Lawrence Livermore National Laboratory (May 2023) requesting validation for chlorine (Idaho National Laboratory (INL), Institute de radioprotection et de surete nucleaire (IRSN), Savannah River Nuclear Site (SRNS), LANL, and Y-12). Los Alamos National Laboratory has recently performed, and benchmarked, an experiment titled Chlorine Worth Study (CWS) for the internal operations at the lab, but due to the difficulties in precisely characterizing the material compositions of the chlorine absorbers it is advantageous to perform a complimentary study with a different chlorine-based absorber material. Furthermore, having a uranium-based vs plutonium-based experiment provides a separate and important validation basis for criticality safety and nuclear data evaluation. The original final design report for the study of chlorine absorption using the TEX-HEU experimental base was presented in 2022, but used the same chlorine-bearing materials that were found to be difficult to characterize in the LANL benchmark. A complete redesign of the experiment has been performed looking at alternative absorber materials in various forms to produce an experiment that is fully characterizable. This report presents five novel chlorine experiments using the TEX-HEU test bed which provides direct comparison to the Y-12 and INL/Terrapower application needs, utilizing sodium chloride (NaCl) absorber plates. The absorber plates will consist of granulated NaCl (≥99.5% pure), which will be fully encapsulated in aluminum tins, providing a simple but effective chlorine-based absorber material that can be completely characterized. Of the five configurations presented in this report, it is expected that two or three configurations will be down selected for the actual experiment, with the other configurations being alternates. Three of the configurations are in the standard configuration, where the absorber is placed directly on the HEU fuel plates, and two in the sandwich configuration, where the absorber is surrounded by polyethylene moderators to force additional neutron thermalization prior to reaching the absorber. There are two thicknesses of NaCl absorber plates: 3/16” and 1/4” active thicknesses (i.e. not including the encapsulation). Both variations of the absorbers have an active absorber radius of 6” and a total radius of 7.5” to match the diameter of the HEU plates, with the outer 1.5” being aluminum encapsulation. The high-density polyethylene (HDPE) moderators are of the thicknesses: 27/16”, 7/4”, 1/8”, 11/16”, and 3/4”. The final configurations have six (one sandwich and one standard configuration), eight (one sandwich and one standard configuration), and 18 total fuel layers (standard configuration). The standard and sandwich configurations were designed such that the differences in moderator thicknesses are 1” HDPE, which were already procured for the original CED-2. The proposed configurations were precisely tuned to closely match the sensitivity profiles and neutron spectra of the Y-12 upset cases and were also compared to the INL/Terrapower upset cases. The assessment of experimental uncertainties of the non-absorber components was predicted to be 0.00114 Δk eff . The assessment of uncertainties resulting from the absorbers was predicted to be 0.00029 Δk eff . This results in a total uncertainty of 0.00118 Δk eff . Many of the largest uncertainties, namely the moderator densities, may be reduced with precision dimensional inspection of the components. The 1” HDPE moderators as well as the HDPE reflectors from the original CED-2 were incorporated in the final designs presented here. Additional HDPE moderator plates must be fabricated to complete the configurations. NaCl absorber plates will by fabricated at LLNL. The total additional cost is expected to be $\$$54,250 for the remaining components. Precise inspection, including dimensional, mass, density, and impurity, is recommended for all components. LLNL estimates that these costs are around $\$$12,000. It is expected, based on previous TEX-HEU experiments, that three weeks of experimental facility time is needed to complete the experiments.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗