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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 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Performance testing of a prototype Pd-Ag diffuser

The fusion fuel cycle has gained significant attention over the last decade as interest in fusion programs has increased. One of the critical components of the fusion process is the tritium fuel cycle. The tritium fuel cycle is designed to supply and recycle process tritium at a specific throughput rate. One of the most important processes within the tritium fuel cycle is the clean-up of the of the process tritium. This step will initially separate the hydrogen isotopes (H2, D2, and T2) from the rest of the process gas using Pd-Ag diffusers or permeators. The Pd-Ag diffuser is an integral component for any tritium purification system; whether part of the United States’ defense mission or fusion programs. Domestic manufacturers of Pd-Ag diffusers are extremely limited and only a few manufacturers exist. Johnson-Matthey (JM) Pd-Ag diffusers (permeators) have previously been evaluated for the separation of hydrogen isotopes from non-hydrogen gas species in the process. JM is no longer manufacturing Pd-Ag diffusers and a replacement vendor needs to be identified to support future needs. A prototype Pd-Ag diffuser has been manufactured by Power and Energy, and is considered a potential replacement for the JM diffuser for tritium service. New diffuser designs for a tritium facility for any fusion energy applications must be characterized by evaluating their operating envelope prior to installation in a tritium processing facility. The prototype Pd-Ag diffuser was characterized to determine the overall performance as a function of the permeation of hydrogen through the membrane. The tests described in this report consider the effects of feed gas compositions, feed flow rates, pump configuration and internal tube pressure on the permeation of H2 through the Pd-Ag tubes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Design of the High Rigidity Spectrometer at FRIB

A High Rigidity Spectrometer (HRS) has been designed for experiments at the Facility for Rare-Isotope Beams (FRIB) at Michigan State University (MSU). The HRS will allow experiments to be performed with the most exotic neutron-rich isotopes at high beam energies (≳100 MeV/u). The HRS consists of an analysis beamline called the High-Transmission Beamline (HTBL) and the spectrometer proper called the Spectrometer Section. Here, the maximum magnetic rigidity of the HRS is 8 Tm, which corresponds to the rigidities at which rare-isotope beams are optimally produced at FRIB. The resolving power, angular acceptance, and momentum acceptance are set to match the anticipated scientific program. An ion-optical design developed for the HRS is described in detail, along with the specifications of the associated magnet and detector systems.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Status Report on NEAMS PROTEUS/ORIGEN Integration

The US Department of Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program has contributed significantly to the development of the PROTEUS neutron transport code at Argonne National Laboratory and to the Oak Ridge Isotope Generation and Depletion Code (ORIGEN) depletion/decay code at Oak Ridge National Laboratory. PROTEUS’s key capability is the efficient and scalable (up to hundreds of thousands of cores) neutron transport solver on general, unstructured, three-dimensional finite-element-type meshes. The scalability and mesh generality enable the transfer of neutron and power distributions to other codes in the NEAMS toolkit for advanced multiphysics analysis. Recently, ORIGEN has received considerable modernization to provide the high-performance depletion/decay capability within the NEAMS toolkit. This work presents a description of the initial integration of ORIGEN in PROTEUS, mainly performed during FY 2015, with minor updates in FY 2016.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

The Mk-18A Target Recovery Program: A Collaborative Effort - 19186

The National Nuclear Security Administration (NNSA) Office of Nuclear Materials Integration (OMNI) recently developed a strategy and subsequently developed a project scope for recovering valuable, indispensable radioactive isotopes from Mark-18A (Mk-18A) fuel targets currently stored at the Department of Energy's Savannah River Site. The Savannah River National Laboratory, the Oak Ridge National Laboratory, and OMNI in a joint effort with the Department of Energy-Savannah River Office and the Department of Energy-Office of Science developed a comprehensive plan for the Mk-18A fuel target recovery effort. The cooperation between the various organizations and the direction provided by OMNI allowed this effort to remain focused on the desired final deliverable while accomplishing the early goals both on schedule and within the budget constraints provided by OMNI. To date, specific hardware for the recovery effort has been manufactured, various flowsheets outlining the processing of the materials have been developed, and cooperation between the various organizations is remaining streamlined. The recovery and processing of these targets will provide valuable, unique isotopes for use by various organizations both within the DOE and outside of the DOE. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

The Los Alamos Neutron Science Center Spallation Neutron Sources

The Los Alamos Neutron Science Center (LANSCE) provides the scientific community with intense sources of neutrons, which can be used to perform experiments supporting civilian and national security research. These measurements include nuclear physics experiments for the defense program, basic science, and the radiation effect programs. This paper focuses on the radiation effects program, which involves mostly accelerated testing of semiconductor parts. When cosmic rays strike the earth's atmosphere, they cause nuclear reactions with elements in the air and produce a wide range of energetic particles. Because neutrons are uncharged, they can reach aircraft altitudes and sea level. These neutrons are thought to be the most important threat to semiconductor devices and integrated circuits. The best way to determine the failure rate due to these neutrons is to measure the failure rate in a neutron source that has the same spectrum as those produced by cosmic rays. Los Alamos has a high-energy and a low-energy neutron source for semiconductor testing. Both are driven by the 800-MeV proton beam from the LANSCE accelerator. The high-energy neutron source at the Weapons Neutron Research (WNR) facility uses a bare target that is designed to produce fast neutrons with energies from 100 keV to almost 800 MeV. The measured neutron energy distribution from WNR is very similar to that of the cosmic-ray-induced neutrons in the atmosphere. However, the flux provided at the WNR facility is typically 5×107 times more intense than the flux of the cosmic-ray-induced neutrons. This intense neutron flux allows testing at greatly accelerated rates. An irradiation test of less than an hour is equivalent to many years of neutron exposure due to cosmic-ray neutrons. The low-energy neutron source is located at the Lujan Neutron Scattering Center. It is based on a moderated source that provides useful neutrons from subthermal energies to ~100 keV. The characteristics of these sources, and ongoing industry program are described in this paper.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Characterization of Aluminum Metal Powders for Plutonium-238 Program

As part of radioisotope thermoelectric generator (RTG) production for power sources in deep space exploration, Oak Ridge National Laboratory (ORNL) produces plutonium-238 ( 238 Pu). To produce 238 Pu, neptunium-237 ( 237 Np) targets are fabricated at ORNL and subsequently irradiated at the High Flux Isotope Reactor (HFIR) at ORNL and the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). Specifically, current flowsheets utilize neptunium dioxide ( 237 NpO 2 ) targets. Aluminum (Al) powder is blended with 237 NpO 2 target material to improve thermal conductivity during irradiation. The current composition of 237 NpO 2 /Al pellets (i.e., cermets, or ceramic–metallics) is 20% 237 NpO 2 , 70% Al, and 10% void space. Al powders utilized by the 238 Pu program are high-fired under vacuum before blending into targets. The primary objective of this work is to measure the physical, chemical, and thermal properties of Al before and after the high-firing process. Properties of interest include the morphology, surface texture, particle size, crystal phase(s), and thermal conductivity of the material. These properties are measured with existing materials characterization equipment at ORNL, including powder x-ray diffraction (pXRD), scanning electron microscopy (SEM) with energy-dispersive spectroscopy (EDS), laser flash analysis (LFA), and thermogravimetric analysis (TGA). A secondary objective of this work is to compare the Al powders before and after the high-firing procedure to determine the effects of high-firing on chemical, physical, and thermal properties and to determine the efficacy of the high-firing process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

The ARGUS Solution Reactor and Molybdenum Production: A Summary Report Based on Open Literature

The (Aqueous Homogeneous Reactor) AHR was conceptualized at Los Alamos for work in advancing weapon design with initial experiments completed at Los Alamos National Laboratory (LANL) and Oak Ridge National Laboratory (ORNL) in the early 1940s and 1950s. The reactor design was discovered to be an efficient mean of production for medical radio isotopes (e.g. molybdenum-99 and other fission products) due to its streamlined process of extracting target radioactive sources from the fissile solution via a separation system. As the demand for medical radioisotopes has increased so has the need for a safe and low cost isotope source, the most commonly used of which is technetium-99m (99m Tc), daughter product from molybdenum decay. Studies indicate solution reactors pose a significantly lower hazard to the workers, public and environment, compared to other modern reactors (International Atomic Energy Agency 2008). The inherent safety and ability to selectively extract radionuclides from its fuel solution make AHRs particularly well suited for medical isotope production. This discussion narrows its focus to one of the longest running aqueous homogeneous reactor systems, ARGUS. A study of the conversion of ARGUS to a low-enriched uranyl (LEU) sulfate fuel solution was part of the Global Threat Reduction Initiative (GTRI) program. The goal of this paper is to provide a summary of the ARGUS reactor system based on open literature available.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2017

Special Actinide Recovery from Mark-18A Target Material

The US Department of Energy (DOE) manages an inventory of materials that contains a range of long-lived radioactive isotopes that were produced from the 1960s through the 1980s by irradiating targets in production reactors at the Savannah River Site (SRS) to produce special heavy isotopes for DOE programmatic use, scientific research, and industrial and medical applications. During the late 1960s, one of the production reactors was configured to operate in a very high flux mode, and >8 kg of 242Pu plutonium contained in Mark-18A (Mk-18A) targets was irradiated to produce 252Cf for use in neutron source applications. Twenty-one targets were processed at the Oak Ridge National Laboratory (ORNL) to recover the 252Cf, heavy curium (i.e., curium rich in 246Cm through 248Cm), and 244Pu. Sixty-five of the Mark-18A targets were not processed and are currently in wet storage at SRS. DOE’s Office of Nuclear Materials Integration (ONMI) established a Mk-18A Target Material Recovery Program (MTMRP) to separate and preserve the unique materials in these targets. Evaluations are also under way to determine whether additional rare long-half-life fission product isotopes present in the targets in much smaller quantities, such as 93Zr, can be economically recovered. These evaluations include assessment of the quality, value, and demand of the materials; process modifications required to facilitate recovery; and the incremental costs for their recovery from the Mk-18A targets.

Patton, Bradley↗
DOE OSTI2016

Diphosphine-Protected Au 22 Nanoclusters on Oxide Supports Are Active for Gas-Phase Catalysis without Ligand Removal

Investigation of monodispersed and atomically-precise Au nanoclusters provides a route to understand the roles of coordination, size, and ligand effects in Au catalysis. We have explored the catalytic behavior of a newly-synthesized Au 22 (L 8 ) 6 nanocluster (L = 1,8-bis(diphenylphosphino) octane) with in situ uncoordinated Au sites supported on TiO 2 , CeO 2 and Al 2 O 3 . Stability of the supported Au 22 nanoclusters was probed structurally by EXAFS and HAADF-STEM, and their adsorption and reactivity for CO oxidation were investigated by IR absorption spectroscopy and temperature programed flow reaction. Low temperature CO oxidation activity was observed for the supported pristine Au 22 (L 8 ) 6 nanoclusters without ligand removal. Isotopically labeled O 2 was used to demonstrate that the reaction pathway occurs through a redox mechanism, consistent with the observed support-dependent activity trend: CeO 2 > TiO 2 > Al 2 O 3 . Substantiated by density functional theory (DFT) calculations, we conclude that the uncoordinated Au sites in the intact Au 22 (L 8 ) 6 nanoclusters are capable of adsorbing CO, activating O2 and promoting CO oxidation reaction. Thanks to the presence of the in situ coordination unsaturated Au atoms, this work is the first clear demonstration of a ligand-protected Au nanocluster that are active for gas phase catalysis without the need of ligand removal.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

Final Report for the Workshop for Applied Nuclear Data Activities

The Workshop for Applied Nuclear Data Activities (WANDA) was held at the Elliot School of International Affairs at George Washington University from 22 to 25 January 2019. The purpose of WANDA was to bring subject matter experts from the national laboratories, universities and industry together with government program managers and their advisors to develop collaborative plans of action (e.g., roadmaps ) to address outstanding issues in nuclear data that effect applications ranging from nuclear energy to national security and nonproliferation to isotope production. WANDA was the latest in a series of collaborative, cross-programmatic workshops focused on nuclear data needs for applications that started in 2015 with the Nuclear Data Needs and Capabilities for Applications Workshop (NDNCA) and continuing with the Nuclear Data Roadmapping Enhancement Workshop (NDREW) in 2018 focused on nuclear data needs for applications. NDREW and WANDA played an important role in providing guidance for the recently created Nuclear Data Interagency Working Group (NDIAWG) comprised of program managers from a panoply government agencies. The NDIAWG issued two Funding Opportunity Announcements (FOA) in FY2017 and FY2018 which are supporting collaborative efforts to address high-priority nuclear data needs.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

Monthly Atmospheric 13C/12C Isotopic Ratios for 11 SIO Stations (1977-2008)

Stable isotopic measurements for atmospheric 13C/12C and 18O/16O at global sampling sites were initiated by Dr. C.D. Keeling and co-workers at Scripps Institution of Oceanography (SIO) in 1977. These isotopic measurements complement the continuing global atmospheric and oceanic CO2 measurements initiated by Keeling in 1957. This work is currently being continued under the direction of R.F. Keeling, who also runs a parallel program at SIO to measure changes in atmospheric O2 and Ar abundances (Scripps O2 Program). A more complete set of 13CO2 data is found online at http://scrippsco2.ucsd.edu/data/atmospheric_co2.htmlFor access to the data files, click this link to the CDIAC data transition website: http://cdiac.ess-dive.lbl.gov/trends/co2/iso-sio/iso-sio.html

12C↗
DOE OSTI2020

Los Alamos National Laboratory Release of Cesium-137 Corrective Action Plan

This Corrective Action Plan (CAP) was developed by a team chartered by John Sarrao, Deputy Director for Science, Technology, and Engineering (DDSTE), to review and respond to the findings of the Joint National Nuclear Security Administration/Triad National Security (NNSA/Triad) Investigation Team (JIT) Joint Investigation Report for the “Sealed Source Recovery at the University of Washington Harborview Training and Research Facility Results in Release of Cesium-137 on May 2, 2019,” dated March 30, 2020. On May 2, 2019, International Isotopes, Inc. (INIS), a subcontractor to Triad National Security, LLC, inadvertently breached a sealed cesium-137 source at the University of Washington (UW), Harborview Medical Center Research and Training (HRT) building in downtown Seattle while attempting to recover the source for the National Nuclear Security Administration (NNSA) Off-Site Source Recovery Program (OSRP). The source breach resulted in contamination of personnel and the HRT building, as well as a release of radioactive material to the environment

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

The Texas A&M Radioisotope Production and Radiochemistry Program

The main motivation of the project at Texas A&M University was to carry out the production of critically needed radioisotopes used in medicine for diagnostic and therapy, and to establish an academic program in radionuclide production and separation methods. After a lengthy battle with the Texas A&M University Radiation Safety Office, the Texas Department of State Health Services granted us a license for the production of radionuclides in July 2015, allowing us to work in earnest in our project objectives. Experiments began immediately after licensing, and we started the assembly and testing of our target systems. There were four analytical/theoretical projects and two experimental target systems. These were for At-211 production and for Zn- 62/Cu-62 production. The theoretical projects were related to the production of Mo-99/Tc-99m using (a) a subcritical aqueous target system and (b) production of Tc-99m from accelerator-generated Mo-99 utilizing a photon-neutron interaction with enriched Mo-100 targets. The two experimental projects were the development of targetry systems and production of At-211 and Zn-62/Cu-62 generator. The targetry system for At-211 has been tested and production of At-211 is chronic depending of availability of beam time at the cyclotron. The installation and testing of the targetry system for the production of Zn-62/Cu-62 has not been finalized. A description of the systems is described. The academic program in radionuclide production and separation methods was initiated in the fall of 2011; due to the lack of a radiochemistry laboratory, it was suspended. We expect to re-start the academic program at the Texas A&M Institute for Preclinical Studies under the Molecular Imaging Program.

43 PARTICLE ACCELERATORS↗
DOE OSTI2021

Demonstration of Defense Waste Processing Facility (DWPF) Higher Fissile Content Glass

The Accelerated Basin De-inventory (ABD) Program has been proposed as an alternative for future spent nuclear fuel (SNF) and nuclear material processing at the Savannah River Site (SRS). This approach would change the baseline H-Canyon (HCAN), Concentrate, Storage, and Transfer Facility (CSTF), and Defense Waste Processing Facility (DWPF) operations. The ABD Program would require that all domestic and foreign research reactor SNF currently at SRS be dissolved, stored, and then transferred to CSTF without the recovery of uranium. Preliminary assessments in the ABD Program plan have shown that ~5000 extra SRS high-level waste (HLW) canisters would be produced if the fissile mass loading remains at the current 897 g/m 3 limit; however, increasing the limit to 2500 g/m 3 would result in ~520 extra canisters. Thus, the ABD Program plan requires an increase of the DWPF fissile mass loading limit to 2500 g/m 3 to minimize canister production. DWPF considers the following isotopes in the calculation of fissile mass loading: U-233, U-235, Pu-239 and Pu-241.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Status, Challenges, and Plans for Consistent Modeling and Evaluation of Fission Data: Nubar, PFNS, and FPY [Slides]

This presentation discusses the importance of consistent evaluations for fission observables. Currently, prompt fission quantities (e.g., neutron multiplicity, neutron energy spectrum, and fission product yields) are evaluated independently of one another, either with separate models or data-only/non-model evaluations (few – if any – shared model inputs) and inconsistencies can arise in evaluated data (e.g. NSE 190, 258 (2018)). They have found that consistent, model-based evaluations lead to more physical constraints on free parameters in the model, more robust predictions for other observables and other isotopes, that there are correlations between the uncertainties on different observables, and that uncertainties can be propagated to unmeasured quantities. They are working on developing models to the point where fission observables are calculated at the quality needed for an evaluation. They are striving to consistently calculate fission observables, such as $\bar{v}$, PFNS, and FPY, to a quality suitable for an evaluation. Current evaluations for fission observables use separate models for each observable (with minimal shared input) or are purely data driven. LANL models, such as the Monte Carlo CGMF and deterministic BeoH, have different strengths; the similar fission fragment initial conditions and decay models leads to consistency between the two codes (FPY can be connected to $\bar{v}$ and PFNS, even when calculated by different codes). Work is underway by Lovell to optimize the models and Neudecker is working to perform detailed uncertainty quantification. The PFNS, in particular, presents a significant challenge.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Mixing Study for the Modification of H-Canyon Tank 31 and 32 Recirculation Lines Via M-Star®

As a part of the Accelerated Basin De-Inventory (ABD) program Tanks 31 or 32 will be re-purposed to support increasing the number of annual Material Test Reactor (MTR) and High Flux Isotope Reactor (HFIR) dissolutions. The proposed plan will allow Tanks 31 or 32 to be used as the dissolver cold chemical solution makeup tank and storage tank. The dissolver cold chemicals are 50% nitric acid, process water, mercuric nitrate, and gadolinium nitrate. Tanks 31 and 32 are 9 ft. (outer diameter) x 36 ft, horizontal, and can hold approximately 56,000 liters (15,000 gallons) each. One full volume of Tanks 31 or 32 can support 4 HFIR 6.4D equivalent batches. Both tanks are equipped with liquid level instrumentation. Tank 31 is equipped with 1 pump used for transfer and recirculation while Tank 32 is equipped with two pumps, one for transfer and the other for recirculation. Each of the pumps is equipped with sample taps. The tanks do not have specific gravity instrumentation, an agitator, or a sampler. Piping modifications will also be performed to supply the tank with cold chemicals. The current dissolver chemical composition is 5.0 -8.5M HNO3 and ~0.2 g Gd/L. The recirculating pump connected to Tank 31 is rated for a flowrate 175 gpm while the pump connected to Tank 32 is rated for a flowrate 50 gpm.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Sum-of-Fractions Methodology for Actinides in Water- and Polyethylene-Moderated and -Reflected Systems

Sum-of-fractions is a method intended to make sure a subcritical margin for aqueous solutions and slurries of fissionable isotopes exists. The method indicates that a system is subcritical if the sum of the ratios of the mass of each isotope (in a mixture) to its individual minimum subcritical mass limit is less than or equal to one. Historically, the basis of the sum-of-fractions has been derived from allowances given in the American National Standards Institute (ANSI)/ American Nuclear Society (ANS)-8.15-1981. However, the allowance was removed in ANSI/ANS-8.15-2014 due to a lack of technical basis. A methodology was developed to assess the validity of using the sum-of-fractions for water- or polyethylene-moderated systems for the following nuclides: 232 U, 233 U, 234 U, 235 U, 237 Np, 236 Pu, 238 Pu, 239 Pu, 240 Pu, 241 Pu, 242 Pu, 241 Am, 242 m Am, 243 Am, 242 Cm, 243 Cm, 244 Cm, 245 Cm, 246 Cm, 247 Cm, 249 Cf, and 251 Cf. The methodology uses available benchmark data for mixtures of 233 U, 235 U, and 239 Pu to establish the calculational margin, and a mass limit reduction to establish the margin of subcriticality. Water- or polyethylene-moderated and -reflected mixtures containing the nuclides are evaluated with the code system, SCALE 6.2.4. Including the calculational margin, subcritical mass limits for each nuclide were computed for optimally water- or polyethylene-moderated and fully reflected systems. These masses were used to create nuclide mixtures in which the sum of the mass to subcritical mass limit ratios is one. The various nuclide mixtures were modeled over a range of moderation and demonstrate the keff does not exceed the calculational margin. For additional assurance of subcriticality, a significant mass reduction is applied to each computed minimum critical mass of the nuclides without adequate benchmark data consistent with the method in ANSI/ANS-8.15-2014.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

A Plan to Qualify New Fuel for the High Flux Isotope Reactor for Material Minimization

The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) is one of five high power research reactors the Office of Material Management and Minimization (M3) Program, Office of Conversion is working to convert from using highly enriched uranium (HEU) fuel to using low-enriched uranium (LEU) fuel. This effort stems from the primary objective within the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA) to achieve permanent threat reduction by minimizing, and when possible, eliminating weapon-usable nuclear material around the world. Under M3’s Office of Conversion, the U.S. High Performance Research Reactor (USHPRR) Project is pursuing fuel qualification and licensing of LEU fuels to support the high-performance reactors. All high-performance reactors except HFIR will be converted to LEU monolithic uranium-molybdenum alloy fuel. HFIR will be evaluated for conversion to LEU using a uranium silicide fuel, namely, U3Si2-Al dispersion fuel. The mission of the USHPRR Project is to develop the technology needed to reduce, and eventually eliminate, worldwide use of HEU in civilian applications. The goal is to develop the technical means needed to use low enriched uranium (LEU) instead of HEU fuel in research and test reactors without significant penalties in performance, economics, or safety of the reactors. The USHPRR Project has four major elements, called Pillars: Fuel Qualification (FQ) managed at Idaho National Laboratory (INL), Fuel Fabrication (FF) managed at Pacific Northwest National Laboratory (PNNL), Reactor Conversion (RC) managed at Argonne National Laboratory (Argonne), and Cross-Cutting (CC) managed at Savannah River National Laboratory (SRNL). FQ is responsible for the qualification of the fuel type. RC is responsible for supporting reactor conversion analysis and overseeing licensing submittals leading to conversions of domestic reactors to LEU fuel. For the FQ effort, FQ (INL) worked in collaboration with RC (Argonne) and ORNL to develop the plan for the uranium silicide fuel qualification for HFIR. The resulting HFIR Fuel Qualification Plan provides the general approach for the USHPRR team to move the selected uranium silicide fuel design for HFIR conversion through qualification. Authorization and use in HFIR will be approved through the DOE’s Office of Science. Uranium silicide fuel was previously qualified in NUREG-1313 at an approximate maximum heat flux of 1.4 MW/m2 and a maximum fuel section temperature of about 130°C. In addition to the different regulator process utilized by DOE, these upper limits will be exceeded in HFIR; therefore, further testing will be necessary to ensure the fuel can meet HFIR qualification requirements. The HFIR fuel loading may exceed 4.8 gU/cm3 which was determined in the NUREG-1313 safety evaluation to be acceptable for use in non-power NRC-licensed reactors provided there exist no other safety considerations. In addition, the uranium silicide fuel will need to be qualified in a HFIR-specific design. This plan includes the currently available information from the USHPRR Project Functions and Requirements document and expands these requirements to ensure that planned tests have traceable results providing evidence that the requirements have been met. Data collection methods are discussed as well as the process to show that the requirements have been met. This document is designed to provide a pathway for researchers to obtain data necessary and at the appropriate quality level for HFIR fuel qualification

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗