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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 OSTI2023

BRR Cask Use for Pu-238 Isotope Production

The Department of Energy (DOE), in partnership with its national laboratories and the National Aeronautics and Space Administration (NASA), is responsible to produce Pu-238 isotope in the United States for use in space exploration. Major activities in the DOE complex are focused at Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL). INL is responsible for storing Np-237 feedstock, irradiation qualification in the Advanced Test Reactor (ATR), and irradiating targets containing Np-237 to produce Pu-238. ORNL is responsible for target design, target fabrication, irradiation qualification in HFIR, and processing of targets to extract Pu-238 heat source material. A key part of the program lifecycle is moving irradiated targets from INL to ORNL. The BEA Research Reactor (BRR) cask was identified as a potential shipping cask for the transport of both unirradiated and irradiated targets between the project sites. This paper will discuss the production at INL and shipment of Pu-238 to ORNL using the BRR cask.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Validating ENDF/B-VIII.1beta2 with LLNL Pulsed-Sphere Neutron-Leakage Spectra [Slides]

This presentation covers noteworthy changes from VIII.1beta0/1 to VIII.1beta2. 6 Li has been fixed (also seen in 7 Li sphere). For 9 Be: we would recommend studying what nuclear data led to an increase in C/E right after peak. For Pb isotopes: we are concerned about new structures introduced after peak. Finally discussed for 239 Pu: Is it worth improving right now?

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Accumulation of trace elements in soil and fauna within a site historically contaminated with coal combustion residues

Legacy contaminants tied to energy production are a worldwide concern. Coal combustion residues (CCRs) contain high concentrations of potentially toxic trace elements such as arsenic (As), mercury (Hg), and selenium (Se), which can persist for decades after initial contamination. Coal combustion residue disposal methods, including aquatic settling basins and landfills, can facilitate environmental exposure through intentional and accidental releases. Wildlife exposed to CCRs can experience numerous deleterious effects, such as on development, reproduction, and survival. In the current study, we quantified and compared concentrations of As, Hg, Se, and strontium (Sr) within soils and target fauna (three vertebrate and three invertebrate taxa) from a CCR-contaminated site and a reference site within the U.S. Department of Energy’s Savannah River Site, South Carolina, United States. Our objectives were to (1) compare current concentrations of tested elements in soil and resident fauna to levels from our reference site, (2) assess natural attenuation of elements in soils by comparing current concentrations to historic levels, and (3) evaluate the biomagnification potential of the elements measured via body burden and trophic position correlations among fauna. Here, element concentrations were higher in contaminated soils than reference soils; however, concentrations in 2022 were unchanged from concentrations measured in 2003, suggesting no natural attenuation of tested elements. Additionally, target fauna had elevated As, Se, and Sr levels in comparison to reference samples. A positive correlation was observed in southern toads between Sr concentrations and trophic position, as assessed by nitrogen stable isotope ratios, suggesting potential for biomagnification of Sr within our study system. Collectively, our results demonstrate that legacy contaminants are still present and bioaccumulate in a diversity of taxa in a CCR-contaminated site that has not received effluents in more than 50 years, suggesting monitoring programs in CCR-contaminated sites should be maintained long term in the absence of remediation.

63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗
DOE OSTI2021

National User Resource for Biological Accelerator Mass Spectrometry (Annual Report)

The National User Resource for Biological Accelerator Mass Spectrometry (User Resource) will provide isotopic analysis (primarily radiocarbon or 14C) by accelerator mass spectrometry (AMS) for NIH- funded researchers across the United States and will be the only User Resource of its type in the United States. The User Resource will provide measurement capability and expertise to a research community that requires highly sensitive, quantitative isotope analyses. Since commissioning a new accelerator mass spectrometer in June 2014, we have measured over 4000 samples a year for collaborators and service users. The User Resource will enable us to continue to meet these research needs, as well as provide for new users whose research programs would benefit from AMS as a measurement tool.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

National User Resource for Biological Accelerator Mass Spectrometry Annual Report

The National User Resource for Biological Accelerator Mass Spectrometry (User Resource) will provide isotopic analysis (primarily radiocarbon or 14C) by accelerator mass spectrometry (AMS) for NIH-funded researchers across the United States and will be the only User Resource of its type in the United States. The User Resource will provide measurement capability and expertise to a research community that requires highly sensitive, quantitative isotope analyses. Since commissioning a new accelerator mass spectrometer in June 2014, we have measured over 4000 samples a year for collaborators and service users. The User Resource will enable us to continue to meet these research needs, as well as provide for new users whose research programs would benefit from AMS as a measurement tool.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2026

Accelerator Science & Engineering (AS&E) Workforce Roundtable and AS&E Traineeship Principal Investigator Meeting

A one-and-half-day meeting on the Accelerator Science & Engineering (AS&E) Workforce Roundtable and a one-day meeting of Traineeship Principal Investigators took place at Facility for Rare Isotope Beams at Michigan State University on January 21-23, 2025. The meeting identified critical issues in the AS&E workforce and potential solutions, with a focus on DOE’s major accelerator facilities. The Traineeship PIs meeting focused on sharing experience, efficiency, and improvements of the Traineeship Programs conducted in six US universities. Several managers from the DOE High Energy Physics Office participated in the meetings. The discussions resulted in a decision to conduct similar meetings every year.

43 PARTICLE ACCELERATORS↗
DOE OSTI2019

Destructive and Non-Destructive Characterization of various LLW/ILW at NRG - 19078

The Nuclear Research and consultancy Group is located at Petten, the Netherlands, and operates the High Flux Reactor (HFR). The HFR is used for isotope production for medical purposes and material research. During its exploitation several types of radioactive waste are produced, which partly have accumulated at the reactor site. In addition, NRG operated the Low Flux Reactor (LFR), a small 30 kWmax Argonaut type reactor which was used for research and education purposes. Over the past decade several projects on waste management and decommissioning were carried out requiring various approaches, including the radiochemical characterisation of the different emerging waste streams as required by the Dutch regulator. This paper focuses on the practices and technical aspects of the destructive and non-destructive characterization of various types of low- and intermediate-level waste arising from a number of nuclear activities performed at the premises of NRG, i.e. the Dutch legacy waste management program, the treatment of ion exchange resins, and the decommission of the Low-Flux Reactor. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Metabolic flux and resource balance in the oleaginous yeast Rhodotorula toruloides

The yeast Rhodotorula toruloides is a promising bioproduction organism due to its high lipid yields and ability to grow on cheap and abundant substrates. Quantitative, systems-level assessment of its metabolic activity is accordingly merited. Resource-balance analysis (RBA) models capture not only reaction stoichiometry but also enzyme requirements for catalysis, providing valuable tools for understanding metabolic trade-offs and optimizing metabolic engineering strategies. Here, in this work, we present systems-level measurements of R. toruloides metabolic flux based on isotope tracing and metabolic flux analysis. In combination with new proteomic measurements, these flux data are used to parameterize a genome-scale resource balance model rtRBA. We find that S. cerevisiae and R. toruloides grow at nearly indistinguishable rates using similar biosynthetic but dramatically different central metabolic programs. R. toruloides consumes one-fifth as much glucose, which it metabolizes primarily via the pentose phosphate pathway and TCA cycle unlike primarily glycolysis in S. cerevisiae . Overall, across these two divergent yeasts, protein abundances aligned more closely than metabolic flux. Resource balance modeling of these metabolic programs predicts superior theoretical yields but lower productivities in R. toruloides than S. cerevisiae for industrial chemicals, highlighting the value of rapid glucose uptake for productivity but respiratory metabolism for yields.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2026

AMMT Round Robin 316H HFIR Irradiation Test Matrix and Readiness for Insertion

The Round Robin 316H HFIR-2 irradiation campaign, conducted under the Advanced Materials and Manufacturing Technologies (AMMT) Program, is designed to evaluate the variability in irradiation response of laser powder bed fusion (LPBF) stainless steel (SS)-316H across multiple national laboratories, powder heats, and processing conditions, with comparison to wrought counterparts. A total of 11 materials are intended for irradiation in the High Flux Isotope Reactor (HFIR), targeting two irradiation temperatures (400°C and 600°C) and one level of irradiation damage (2 dpa). This irradiation campaign utilizes the standardized general tensile (GENTEN) capsule design to accommodate subsized tensile specimens of these materials. Thermal analyses were performed to ensure appropriate temperature control by design and uniformity within the capsules. All the capsules for this irradiation except one have been successfully assembled, welded, and tested and are ready for HFIR insertion. The last capsule requires re-build following a failed weld. All eight capsules are intended to be inserted in HFIR cycle 517, which is scheduled to start September 8, 2026. This report summarizes the design, material selection, and readiness for insertion of the capsules.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Mummified baboons reveal the far reach of early Egyptian mariners

The Red Sea was witness to important events during human history, including the first long steps in a trade network (the spice route) that would drive maritime technology and shape geopolitical fortunes for thousands of years. Punt was a pivotal early node in the rise of this enterprise, serving as an important emporium for luxury goods, including sacred baboons ( Papio hamadryas ), but its location is disputed. Here, we use geospatial variation in the oxygen and strontium isotope ratios of 155 baboons from 77 locations to estimate the geoprovenance of mummified baboons recovered from ancient Egyptian temples and tombs. Five Ptolemaic specimens of P. anubis (404–40 BC) showed evidence of long-term residency in Egypt prior to mummification, consistent with a captive breeding program. Two New Kingdom specimens of P. hamadryas were sourced to a region that encompasses much of present-day Ethiopia, Eritrea, and Djibouti, and portions of Somalia and Yemen. This result is a testament to the tremendous reach of Egyptian seafaring during the 2nd millennium BC. It also corroborates the balance of scholarly conjecture on the location of Punt.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2014

Determination of attenuation coefficients of heavyweight concretes containing colemanite by using {sup 133}Ba radioactive isotope source - Paper 140

As it is well known, colemanite is a boron ore. Boron is one of the most important underground richness of Turkey. Boron is employed as a constituent material of neutron shielding because of its high absorption of thermal neutrons. Barite ore includes BaSO{sub 4}, which is a well photon radiation absorbent. The literature about radiation attenuation in heavy concretes (including boron and barite ores) is fairly abundant. Abdo and Abdo et al. theoretically calculated and determined cross section by using XCOM computer program for photon and neutron radiations shielding. Akkurt et al. measured radiation transmission of heavy concretes including normal and barite aggregates for different γ-ray energies and calculated linear attenuation coefficients. Bashter et al. studied heavy concretes including hematite-serpentine, ilmenite-limonite as control absorber in nuclear reactors γ-rays and neutron particles shielding. Bashter and Bashter et al. studied heavy concretes including hematite-serpentine, ilmenite-limonite, basalt-magnetite, ilmenite, basalt, steel and magnetite for only photon radiation shielding and calculated linear and mass attenuation coefficients from 10 keV to 1 GeV. Dealmeid et al. studied radiation transmission and angular distribution for 25 MV X-rays by using LINAC. Demir and Keles measured radiation transmission of concrete including boron waste for 59.54 and 80.99 keV γ-rays. Kaplan reported about many heavy concretes studies in his book. Kitis et al. studied heavy concretes including barite aggregates for cyclotron shielding. Mollah et al. determined concretes including ilmenite and magnetite aggregates for neutron shielding by using {sup 252}Cf radioactive source and BF{sub 3} neutron detector. Yarar and Bayuelken studied concretes including colemanite aggregates for neutron shielding. In this paper, we have measured radiation transmission of thirteen different heavy concretes produced by using barite, colemanite and normal aggregates for 276, 303, 356 and 383 keV γ-ray energies of {sup 133}Ba radioactive isotope by using CdTe semiconductor detector and calculated linear and mass attenuation coefficients. Results were compared to X-ray mass attenuation coefficient values of NIST. The attenuation of γ-ray photons is governed primarily by the atomic number and density of the concrete as a shielding material. Photon attenuation depends on pair production, photoelectric effect and Compton scattering. In this study, linear attenuation coefficients (μ) decrease with the increasing the photon energy for these concretes. Heavy elements there is a fairly sharp minimum in the absorption cross section at about this energy region. It is indicated that attenuation coefficient is useful parameter for radiation shielding. As known, the linear attenuation coefficients of the concretes used in this study at these γ-ray energies are not available in the literature by using CdTe semiconductor detectors. The data shows the effects of including boron-rich colemanite or barium-rich barite in addition to normal aggregate in the concrete mixtures. As would be expected, the barite is desirable for photon attenuation, while the colemanite is desirable for neutron attenuation. The concretes including barite and colemanite would be preferred as materials for building construction of radiation oncology departments, radiology departments, nuclear reactors and radiation research laboratories against photon and neutron particle radiations. But other aggregate properties of colemanite such as abrasion resistance and freeze-thaw durability are also found to be important for concrete but they cannot be quantified at this study. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2018

GlyQ-IQ

GlyQ-IQ is a program that performs a targeted, chromatographic centric search of mass spectral data for glycans. It is designed for high resolution LC-MS datasets. The software uses a list of glycan targets to search for expected features in MS1 spectra. Features are characterized by monoisotopic mass, elution time, and isotopic fit score. Features are annotated by glycan family relationships and optionally by in-source fragmentation patterns.

Monroe, Matthew↗
DOE OSTI2016

Aluminum Metal Dissolution for Mark-18A Target Processing

The Mark-18A Program was established to preserve the unique materials found in Mark-18A targets for future research and development activity within the Department of Energy or other U.S. government agencies. The Savannah River Site has 65 Mark-18A targets to be provided for the recovery of high valued isotopes. The majority of the recovery work will be performed in the Savannah River National Laboratory (SRNL) Shielded Cells due to dose concerns. The recovered product will be transported to the Oak Ridge National Laboratory (ORNL) for future research and development tasks. The Mark-18A target material will be removed from their current configuration, dissolved, chemically separated, and calcined to a stable oxide. The baseline flowsheet is patterned after the flowsheet ORNL has previously used to process both Mark-18A and Mark-42 targets. However, it is necessary for SRNL to adapt its facility and the ORNL flowsheet to identify process improvements to successfully recover the actinide materials within the constraints of the SRNL Shielded Cells. These studies evaluated and demonstrated a chemical method for cladding removal using a mixture of sodium nitrate (NaNO 3 ) and sodium hydroxide (NaOH). Experiments provided better insights into dissolution of the Mark-18A aluminum cladding. In particular, these studies measured the effects of temperature on dissolution rate, flammable gas generation, and nitrite formation. Using the lessons learned from batch dissolution testing at SRNL, a continuous dissolution flowsheet was successfully demonstrated using prototypic-shaped materials and chemically-conservative materials. The SRNL target dissolution flowsheet departs from the baseline ORNL flowsheet in small but significant ways as discussed in the following paragraphs. These flowsheet changes facilitate faster initial reaction rates at a given temperature, enable cladding removal at a lower process temperature, and yield a more uniform dissolution rate throughout cladding removal. Multiple tests indicate that the total time at the dissolution temperatures will be about six hours for one fourth of a target. The shorter dissolution time provides the benefit of a single-shift dissolver operation. Off gas sampling determined that the quantities of flammable hydrogen (H 2 ) and ammonia (NH 3 ) expected in the process off gas will meet allowable flammable gas generation rates for the Shielded Cells Facility. The SRNL approach enables NaOH to be added at a constant rate, which simplifies operations in the shielded cells. In all six continuous dissolution tests, the process was easy to control with minimal operator interface. Analyses of process solutions throughout the experiments provide a better understanding of system behavior. The data show consistent behavior across several similar but different test conditions. Between the batch-test and continuous-test data, if future changes to the flowsheet are required, sufficient data exist to adequately predict how the system will respond to those changes. Future work will focus on design, fabrication, and evaluation of a system for dissolution of one-fourth of a Mark-18A target, which will be full-scale operations for SRNL. Based on these results, the process parameters are adequately understood and the system can be operated safely.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the complex aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present studies used current analytical tools to evaluate the various alternate designs for cycle length, scientific performance (e.g., neutron scattering), and steady-state and transient thermal performance using both safety limit and nominal parameter assumptions. The studies concluded that a new reference design combining a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone will allow successful conversion of HFIR. Future collaboration with the program will reveal whether the new reference design can be fabricated reliably and affordably. Following this feedback, additional studies using state-of-the-art developmental analytical tools are proposed to optimize the design of the fuel zone radial contour and the amount and location of both types of neutron absorbers to further flatten thermal peaks while maximizing the performance of the reactor.

Renfro, David G.↗
DOE OSTI2014

Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy’s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the “complex” aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present studies used current analytical tools to evaluate the various alternate designs for cycle length, scientific performance (e.g., neutron scattering), and steady-state and transient thermal performance using both safety limit and nominal parameter assumptions. The studies concluded that a new reference design combining a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone will allow successful conversion of HFIR. Future collaboration with the program will reveal whether the new reference design can be fabricated reliably and affordably. Following this feedback, additional studies using state-of-the-art developmental analytical tools are proposed to optimize the design of the fuel zone radial contour and the amount and location of both types of neutron absorbers to further flatten thermal peaks while maximizing the performance of the reactor.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Alpha Spectrometry Results for Groundwater Samples Collected in Northern Iraq and a Summary of the Environmental Setting of the Adaya Burial Site

The Radiation Protection Center (RPC) of the Iraqi Ministry of Environment continues to evaluate the potential health impacts associated with the Adaya Burial Site, which is located 33 kilometers (20.5 miles) southwest of Mosul. This report documents the radiological analyses of 16 groundwater samples collected from wells located in the vicinity of the Adaya Burial Site and at other sites in northern Iraq. The Adaya Burial Site is a high-risk dump site because a large volume of radioactive material and contaminated soil is located on an unsecure hillside above the village of Tall ar Ragrag. The uranium activities for the 16 water samples in northern Iraq are considered to be naturally occurring and do not indicate artificial (man-made) contamination. With one exception, the alpha spectrometry results for the 16 wells that were sampled in 2019 indicate that the water quality concerning the three uranium isotopes (Uranium-233/234, Uranium-235/236, and Uranium-238) was acceptable for potable purposes (drinking and cooking). However, Well 7 in Mosul had a Uranium-233/234 activity concentration that slightly exceeded the World Health Organization guidance level. Eight of the 16 wells are located in the villages of Tall ar Ragrag and Adaya and had naturally occurring uranium concentrations. Wells in the villages of Tall ar Ragrag and Adaya are located near the Adaya Burial Site and should be sampled on an annual schedule. The list of groundwater analytes should include metals, total uranium, isotopic uranium, gross alpha/beta, gamma spectroscopy, organic compounds, and standard water quality parameters. Our current understanding of the hydrogeologic setting in the vicinity of the Adaya Burial Site is solely based on villager's domestic wells, topographic maps, and satellite imagery. To better understand the hydrogeologic setting, a Groundwater Monitoring Program needs to be developed and should include the installation of twelve groundwater monitoring wells in the vicinity of Tall ar Ragrag and the Adaya Burial Site. Characterization of the limestone aquifer and overlying alluvium is needed. RPC should continue to support health assessments for the villagers in Tall ar Ragrag and Adaya. Collecting samples for surface water (storm water), airborne dust, vegetation, and washway sediment should be conducted on a routine basis. Human access to the Adaya Burial Site needs to be strictly limited. Livestock access on or near the burial site needs to be eliminated. The surface-water exposure pathway is likely a greater threat than the groundwater exposure pathway. Installation of a surface-water diversion or collection system is recommended in order to reduce the potential for humans and livestock to come in contact with contaminated water and sediment. To reduce exposure to villagers, groundwater treatment should be considered if elevated uranium or other contaminants are detected in drinking water. Installing water-treatment systems would likely be quicker to accomplish than remediation and excavation of the Adaya Burial Site. The known potential for human exposure to uranium and metals (such as arsenic, chromium, selenium, and strontium) at the Adaya Burial Site is serious. Additional characterization , mitigation, and remediation efforts should be given a high priority.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2020

Consistent Evaluation of the Prompt-fission Neutron Spectrum and Multiplicity for n+ 235,238 U and n+ 239 Pu

This report was written to satisfy a FY20 NCSP milestone on 235,238 U and 239 Pu. The milestone requires to “finalize a report assessing our methodology to evaluate prompt-fission neutron spectrum (PFNS) and multiplicity consistently”. More specifically, we study whether the code CGMF can reproduce ENDF/B-VIII.0 evaluated PFNS and average prompt-fission neutron multiplicities, $\bar{v}$, for 235,238 U and 239 Pu using one joint parameter set per isotope. If CGMF is shown to be able to reasonably reproduce ENDF/B-VIII.0 within its model-parameter space, this code could be used for future consistent evaluations of PFNS and $\bar{v}$. To answer this question, we explore here the parameter space of CGMF and its impact on calculated values and whether they are close to evaluated and experimental data. We also list experimental data that would enter a future evaluation and statistics method that could be used to obtain evaluated data and covariances. We conclude that values of $\bar{v}$ calculated by CGMF are reasonably close to ENDF/B-VIII.0 data, while more work on modeling the PFNS is needed (parameter optimization and model improvements) to reliably use it for evaluations.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Status of the n+ 234 U evaluation in the resolved resonance region (*), (**)

In natural uranium, the 234 U isotope represents only 0.0055%, however, this minor isotope can affect highly enriched uranium metal benchmark calculations. In fact, enriched uranium contains more 234 U than natural uranium as the result of the uranium enrichment process. Therefore, the n+ 234 U nuclear data evaluation is one of the milestones of the APPENDIX B within the Nuclear Criticality Safety Program (NCSP).

AMPX↗