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36 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

The Prompt Fission Uranium Neutron Spectrum (PFUNS) Experiment: Critical Configurations and Irradiations

The objective of the Prompt Fission Uranium Neutron Spectrum (PFUNS) experiment is to reduce the uncertainty of the Prompt Fission Neutron Spectrum (PFNS) of 235 U above 8 MeV. The experiment was performed at the DOE National Criticality Experiments Research Center (NCERC) at the Nevada National Security Site. To meet the experiment objective, activation foils were placed in a central void region of a critical configuration consisting of concentric highly enriched uranium (HEU) metal hemishells. The set of activation foils were chosen based on threshold reactions to neutron energies across the fission spectrum, but especially those in the high energy tail of the fission spectrum. PFUNS was performed on the Planet critical assembly machine at NCERC and uses the Rocky Flats (RF) HEU hemishells. PFUNS has similarities to the Measurement of Uranium Subcritical and Critical (MUSiC) experiment conducted at NCERC in 2021, which also used RF hemishells, but contains a large central cavity to allow for a sample plate to be inserted. This large void means that much more HEU is needed to achieve a critical configuration (108 kg for PFUNS versus 59 kg for MUSiC). This work describes the 2024 experiment execution of four critical configurations and two irradiations for the PFUNS project.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Impact of U-10Mo HALEU Fuel Element Tolerances on the Massachusetts Institute of Technology Reactor safety and operational performance – Neutronics

The U.S. is coordinating efforts for the conversion of six U.S. high performance research reactors (USHPRR), including one critical assembly from highly enriched uranium (HEU) to low-enriched uranium (LEU). In order to continue the mission of these reactors, including the Massachusetts Institute of Technology Reactor (MITR), and achieve similar performance, high assay low-enriched uranium (HALEU) with a high-density metallic alloy of uranium with 10 wt% molybdenum (U-10Mo) is being considered. Following the preliminary design of the proposed MITR LEU fuel elements using the U-10Mo monolithic alloy, the impact of the fabrication specification was assessed. This work focuses on the analysis of select neutronics characteristics of the MITR LEU core as a function of the variation of the relevant fuel specification parameters (e.g., U-10Mo composition, fuel plate thickness, etc.). A separate article submitted to this journal addresses the impact on the thermal hydraulic performance. The analyses in these works are performed based on an all-LEU conversion management plan identified in previous work, in which only the proposed elements are utilized for achieving the conversion of MITR. The variations of two main neutronics characteristics are assessed as a function of the variability of the specification parameters resulting from the fabrication process: the MITR LEU core reactivity and the fuel cycle length. The main findings of this work show that the MITR core can meet the operational requirements during the LEU transition plan under the limiting fabrication parameter combinations considered. In addition, the analyses show that the dependency of the core neutronics characteristics on the specification parameters is highly linear within the specification tolerances. The rates of variation are reported in detail for each parameter and can serve as a powerful tool for future MITR fuel management in cases such as when HALEU supply is established that may allow additional cycle length or other operational benefits.

Conversion↗
DOE OSTI2026

Impact of Molten Gallium on the Microstructure and Corrosion Behavior of Aluminum and Uranium-Aluminum Alloys for Used Nuclear Fuel Reprocessing

Test reactors around the world utilize highly enriched uranium fuel to achieve high neutron fluxes for materials testing. Once spent, the remaining uranium is a valuable resource for subsequent fuel fabrication. However, some of these test reactor cores consist of curved plate-type fuel elements, fabricated using aluminum alloy 6061 (AA6061) cladding to encapsulate a uranium-aluminum alloy (UAlx) fuel matrix. These assemblies require non-standard reprocessing approaches for uranium recovery, as aluminum dissolves readily in acidic solutions, generating large volumes of waste and complicating downstream chemical separations. In this work, we investigate a novel chemical decladding strategy based on the interaction between AA6061/UAlx and molten gallium (Ga). Ga is known to induce severe degradation of aluminum metal through liquid metal embrittlement (LME), even at relatively low Ga concentrations. By penetrating the aluminum crystal lattice, Ga disrupts grain cohesion and facilitates fracture or dissolution of the aluminum matrix. Thermodynamic analysis of the Al–Ga binary phase diagram suggests that Ga may offer a viable pathway to selectively weaken or dissolve the AA6061 cladding, and potentially the aluminum component of the UAlx fuel matrix within. To this end, parametric experiments were performed at 50 °C and 100 °C across a range of Al–Ga atomic fractions. At lower Al fractions, the AA6061 was completely molten after 2 hours of exposure to the Ga metal. In contrast, samples with higher Al fractions (0.9 Al, 0.1 Ga) contained residual solids after 2 hours, which were characterized by microstructural examination using electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). These Al-Ga compositions were also evaluated using FactSage thermodynamic modeling to further elucidate the relationship between phase diagram behavior and LME.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Impact of U-10Mo HALEU fuel element tolerances on the Massachusetts Institute of Technology reactor safety and operational performance – Thermal hydraulics

The U.S. is coordinating efforts for the conversion of six U.S. High Performance Research Reactors (USHPRR) including one critical facility from highly enriched uranium (HEU) to low-enriched uranium (LEU). In order to continue the mission of these reactors, including the Massachusetts of Institute of Technology Reactor (MITR), and achieve similar performance, high-assay low-enriched uranium (HALEU) with a high-density metallic alloy of uranium with 10 wt% molybdenum (U-10Mo) is being evaluated. The impact of the fabrication specification and tolerances was assessed following the preliminary design of the MITR LEU fuel elements using the U-10Mo monolithic alloy. This research focuses on the analysis of fabrication specification impact on thermal hydraulics (TH) characteristic of the MITR LEU core as a function of the variation of the relevant fuel specification parameters (e.g., coolant channel gap thickness, fuel plate thickness, etc.). The analyses are performed based on an all-fresh LEU fuel conversion plan identified in a preliminary safety analysis report submitted to the Nuclear Regulatory Commission. The reactor power margin to the onset of nucleate boiling (ONB) is assessed under the limiting safety system settings (LSSS), where a scram occurs, to ensure there is sufficient margin to the reactor safety limit, which is defined by the onset of flow instability that occurs after the ONB. The best estimate plus uncertainty approach is employed to analyze this TH characteristic, which yields realistic results while maintaining adequate conservatism, utilizing a statistical uncertainty propagation method with the STAT7 code. The TH characteristic is analyzed as a function of the variability of the specification parameters resulting from the fabrication process. The main findings of this study show that the MITR core can meet the TH safety and operational requirements at the all-LEU initial core startup (cycle 1), selected transition cycles (most reactive cycle and most limiting cycle: cycle 3 and 5, respectively) and equilibrium (cycle 14) cores under all limiting fabrication parameter combinations considered. In addition, the analyses show that the dependency of the core power margin to ONB on those specification parameters that have the most direct impact on TH performance is non-linear but monotonically decreasing within the specification tolerances. The third order polynomial fit curves are reported in detail for selected limiting cases and can serve as a powerful tool for future MITR fuel management in cases such as when HALEU supply is established that may allow additional cycle length or other operational benefits.

Conversion↗
DOE OSTI2023

CHALLENGES IN THE DEVELOPMENT OF THE ELECTROREFINING PROCESS AT Y-12

In order to ensure future capabilities of the Y-12 site as older buildings retire, a new process was developed, the Metal Purification Process. The purpose of this new process is to provide a simpler and more efficient uranium metal purification that the previous multi-staged, complex chemical processes. The basis for this technology currently exists in the US DOE complex, but it has not been utilized on a large scale for uranium at higher enrichments. The system requires larger-geometry vessels and furnaces in order to meet the through-put requirements. The crux of the process is the electrorefining cell, consisting of several concentric liners and an unfavorable geometry cylindrical crucible. Within the crucible, a molten Li-CL-KCL-UCL3 salt electrolyte is present. ‘Dirty’ metal is placed into a loading basket and loaded into an electrorefining cell. Here, the metal reacts into the molten salt within the cell to form UCL3, while ‘clean’ metal simultaneously plates out from the salt electrolyte in the form of metal dendrites. These dendrites are then collected and removed from the cell. After removal from the cell, the dendrites are taken to a furnace to remove adhered salts. Then, the salt-free dendrites are moved to a separate furnace for consolidation into the final product form. The development of the Criticality Safety Evaluation for this complex system brought with it many NCS Challenges and lessons learned. These challenges include: design decisions for ensuring subcriticalty in the electrorefining cell during abnormal conditions, the interface of the main glovebox system with auxiliary systems such as the purification system and designated storage, implementation of mass tracking, and concerns from production and operation regarding the movement of material within the system.

36 MATERIALS SCIENCE↗
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 OSTI2021

Enhancing isotope mixing in U-10Mo downblend castings with electromagnetic stirring

Uranium alloyed with 10 wt percent molybdenum (U-10Mo) is the proposed fuel for use in the United States’ high-performance research reactors. The U-10Mo fuel is fabricated by a two-step casting process that downblends highly enriched uranium with depleted uranium (DU) and/or natural uranium (NU) and subsequently alloys the resulting high-assay low-enriched uranium (HALEU) with Mo. Currently, a two-step casting process is needed to meet the ingots’ U-235 enrichment homogeneity specifications. This work demonstrates the ability to provide more homogeneous U-235 distributions during downblend casting by using zirconia for crucibles rather than graphite. When graphite is used as a crucible, the electromagnetic field produced by the induction heater couples directly with the graphite. Zirconia is nonconducting in this study, and therefore does not couple with the field. The induction field couples directly with the metal, and causes electromagnetic stirring (EMS) in the molten metal pool. Eight downblend casting experiments were carried out in this work. Four were performed with zirconia crucibles and four with graphite crucibles. The U-235 enrichment was measured at nine discrete points in each casting using laser ablation multi-collector inductively coupled mass spectrometry (LA-MC-ICP-MS). The enrichment homogeneity was approximately eight times better in zirconia crucible castings than in graphite crucible castings as measured by the plate enrichment range and enrichment coefficient of variation. The results show a single casting step could be used to meet USHPRR enrichment specifications if electromagnetic stirring is present during casting. A related investigation was carried out to determine whether it is appropriate to measure enrichment on as-cast ingots or whether heat treated specimens should be used to accurately characterize enrichment. The results show that the as-cast microstructure plays a significant role in enrichment homogeneity. It is recommended that all enrichment measurements be done after the homogenization heat treatment typical of U-10Mo processing.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Purification of U from U-10Mo scrap generated during the fabrication of high performance research reactor fuel

A low enriched U-Mo alloy fuel is under development to replace highly enriched U fuels currently used in United States high performance research reactors. The alloy casting and fuel fabrication processes will generate scrap streams containing low enriched U (LEU) which must be recovered. Solvent extraction processes were designed using the Argonne Model for Universal Solvent Extraction (AMUSE) to purify solutions containing 20 and 50 g/L U. The feed for the solvent extraction processes was prepared from solutions generated from the dissolution of U-10Mo-Zr foils and U-10Mo-Zr-Al mini-plates. The U purification processes were demonstrated using two, 16-stage banks of miniature mixer-settlers. The solvent extraction experiments demonstrated that all design objectives for the U purification processes could be met. The U recovery in the product stream for each flowsheet was ≥99.9%. The flowsheet demonstrations also showed that the purity of the U Product will meet the requirements of the ASTM International C1462-21 specification for LEU metal enriched to less than 20% 235 U. In conclusion, the AMUSE modeling for both flowsheet demonstrations was validated by comparing predicted and measured concentrations of U, Mo, and Zr in the exit streams and stage samples at steady-state conditions in the mixer-settlers.

modified PUREX process↗
DOE OSTI2025

TEX-Chlorine Assemblies: Highly Enriched Uranium Plates with Sodium Chloride Absorbers Using Polyethylene Moderator and Polyethylene Reflector

This evaluation documents highly enriched uranium (HEU) experimental critical configurations with polyethylene moderators and sodium chloride absorbers conducted as part of the United States Nuclear Criticality Safety Program’s Thermal/Epithermal eXperiments (TEX) program. HEU-MET-MIXED-021 provides the benchmark evaluation of five TEX experiments designed to establish baseline configurations with HEU Jemima plates moderated by high density polyethylene (HDPE). The TEX-HEU experiments were designed to cover five different fission energy regimes by varying the thickness of the interstitial HDPE moderator, with varying fractions of thermal, intermediate, and fast fissions, and to be easily modified to accommodate test materials of interest. HEU-MET-INTER-013 documents the first TEX-HEU variation, incorporating hafnium in seven different experimental configurations. This evaluation covers an additional variant that incorporates absorber plates of compacted high-purity sodium chloride salt. These experiments were motivated by a criticality safety need for validation data for uranium purification by means of electrorefining with chloride salts, especially thermal and intermediate energy configurations resulting from moderator upset conditions, and their design was optimized by matching sensitivity profiles from application cases. All three experimental configurations are judged to be acceptable as benchmark cases. The main parameter varied between the configurations is the thickness of the polyethylene moderators and the sodium chloride absorbers between the HEU plates. Varying the thickness of the polyethylene tunes the neutron energy spectrum between majority thermal (Case 1 and 2) and intermediate (Case 3). The fission fractions, presented in Table 1, are determined calculationally. Case 3 is cross listed as HEU-MET-INTER-014.

42 ENGINEERING↗
DOE OSTI2023

TEX-HEU Baseline Assemblies: Highly Enriched Uranium Plates with Polyethylene Moderator and Polyethylene Reflector (IER-297 CED-4b Report)

This experiment provides the baseline experimental configurations for TEX-HEU. The purpose of the TEX-HEU design is to provide multiple configurations that span the entire neutron energy spectrum and can be easily modified to incorporate diluent materials. Future experiments utilizing the TEX-HEU design are planned to incorporate hafnium, lithium, and chlorine as diluent materials. Five experimental configurations were judged to be acceptable critical benchmark experiments.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

ZPPR-15 Small Reactivity Worth Experiments

In addition to measuring the worth of sodium voiding and the worth of simulated control rods, the ZPPR staff measured the Doppler worth of four standard samples, the worth of axial expansion and the worth of radial bowing in ZPPR-15. Doppler sample worths were measured in ZPPR-15A, ZPPR-15B and ZPPR-15D. The Doppler samples were natural UO 2 , depleted uranium metal, depleted U-10Zr alloy and 33% enriched UO 2 . These samples were placed in a sample capsule in the inner core of ZPPR-15 and heated to various temperatures between 300 K and 1100 K. The resulting changes in reactivity relative to the reference configuration were measured to determine the Doppler worths of the samples. Special segmented drawers were constructed to measure the worth of axial expansion in ZPPR15. The segments were connected to each other, and the back segment was connected to a spring-loaded cable. Pulling the cable introduced small gaps between drawer segments to simulate axial expansion, and releasing the cable eliminated those gaps. The worth of this simulated axial expansion was measured in ZPPR-15A, ZPPR-15B and ZPPR-15D. The worth of radial bowing was measured in ZPPR-15, but the experimenters regarded the result as unsatisfactory. Radial bowing was measured by two methods in ZPPR-17A, so the ZPPR-17A bowing measurements were analyzed in place of the unpublished ZPPR-15 radial bowing measurement. The first ZPPR-17A measurement consisted of rearranging the plates in 96 core drawers at the core-radial blanket boundary to simulate outward motion of the fuel during bowing. The second ZPPR-17A measurement used a new bowing oscillator to move the core plates a small distance vertically in a special drawer. The published uncertainties for the Doppler worth, axial expansion worth and radial bowing worth measurements are the statistical uncertainties in the measurements. In reality, the published uncertainty for each of these measurements is just one component of the total uncertainty. There are additional uncertainties specific to each measurement and a common uncertainty related to conversion from the natural measurement units, cents, to pcm for calculations. A full uncertainty analysis was performed for each measurement. The significant uncertainties were quantified, and a total uncertainty was determined for each measurement. The ZPPR-15 and ZPPR-17A experimental records were used to create detailed as-built Monte Carlo models of the Doppler worth, axial expansion and radial bowing measurement configurations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

HIGHER EFFICIENCY FISSION COUNTER FOR NEUTRON DETECTION

Fission chambers containing 235U have been used for detection of neutrons for many decades. In these chambers 235U is electroplated on a metal which is one plate of an ionization chamber. Since the range of fission products from induced fission in the chamber is extremely short, these deposits from electroplating are so thin that the amount of 235U results in low neutron detection efficiency and the chamber needs high internal 235U surface area to increase efficiency. A typical ~8 in. long, ~2 in. diameter fission counter only contains < 2 g of 235U total, resulting in very low detection efficiency per incident neutron impinging on the outer surface of the detector. However, because of their large electronic pulses from fission products compared to gamma rays it is very easy to discriminate against gamma rays impinging on the detector. These types of chambers are produced commercially and are used worldwide in nuclear research and nuclear power reactors. ORNL has made a higher efficiency fission ionization chamber by using many thin concentric annuli to increase the area. A higher efficiency fission counter would be useful and can be made by incorporating 235U into a glass or ceramic scintillator. For nearly 100 years the green color in antique glass was produced by incorporation of depleted uranium (up to 20 wt. %) into a glass. The concept for this scintillation-optical-based fission detector is to dope a scintillator (either ceramic or glass) with 235U. Glass scintillators are commercially available doped with as much as 8 wt. % 6Li for neutron detection. In the 235U doped scintillator the fission product pulses would be double the size of those in an ion chamber since all fission products produce light in the scintillator whereas in the ionization chamber half do not go into the ionizing gas volume. These higher efficiency fission chambers would be very useful for a wide variety of research, nuclear research, and nuclear power reactor applications. Because of their smaller size and weight, they should be useful for space application such as monitoring of space nuclear power reactors. This report reviews existing fission chamber technology and suggests the development of these small high-efficiency fission chambers. Other fission isotopes could be used for other research applications.

Mihalczo, John [ORNL] (ORCID:0000000254292681)↗
DOE OSTI2025

TEX-Chlorine: Highly Enriched Uranium with Chloride Absorbers to Provide Validation Benchmarks for Y-12 Electrorefining Facility

This report documents the final benchmark of the TEX-Chlorine (IER-499) Thermal Epithermal eXperiments (TEX) with highly enriched uranium with chlorine absorbers and high-density polyethylene reflectors and moderators. TEX-Chlorine is a variation of the TEX-HEU baseline assembly with the addition of sodium chloride absorber plates. This evaluation contains three experimental configurations that were performed on Comet at NCERC between July and August 2024. The three configurations, which were acceptable as benchmark cases, spanned from thermal (first two configurations) to fast (third configuration). All three cases were reviewed and accepted by the ICSBEP TRG in April 2025 and was submitted to the ICSBEP in August 2025 after receiving subgroup approval.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

TEX-Hf Assemblies: Highly Enriched Uranium Plates with Hafnium Using Polyethylene Moderator and Polyethylene Reflector

This report documents the integral experiment evaluation and publication for IER-532 (TEX-Hf), Thermal/Epithermal eXperiments (TEX) with highly enriched uranium (HEU) fuel and hafnium (Hf), moderated and reflected by polyethylene. The design of TEX-Hf is a variation of IER-297 (TEX-HEU), with the inclusion of hafnium as a diluent material. The experiment and evaluation include seven configurations, all of which were acceptable as benchmark cases. These benchmark cases provide validation for hafnium in the thermal, intermediate, and fast neutron energy regimes by maximizing the sensitivity in k eff to the hafnium cross sections. The evaluation was reviewed and accepted by the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Technical Review Group on April 17, 2024, and was submitted to the ICSBEP in August 2024 following subgroup approval.

42 ENGINEERING↗
DOE OSTI2025

Evaluation of U10Mo Fuel Plate Performance Modeling Over Hot Isostatic Press and Hydraulic Bending for MURR DDE Plates

The United States High Performance Research Reactor Program’s objective is to reduce the amount of highly enriched uranium currently implemented in research reactors. The conversion of these research reactors requires designing a monolithic U10Mo plate fuel, with the fuel plate geometry being dependent on each research reactor. The process of forming the plates includes a hot isostatic pressing (HIP) to manufacture a prototypic plate. In the case of the Missouri University Research Reactor (MURR) design demonstration element (DDE) plate manufacture, plates that have been through HIP are then curved using dies and a hydraulic press to impart the desired curvature. Both fabrication processes impart residual stresses into each fuel plate region, with the curvature of the plates taking some regions of the fuel plate up to their material yield stresses, accompanied by plastic strain. The amount of plastic strain and stress imparted onto each MURR DDE plate is determined by the radius of curvature, thickness of each region, and overall width of the fuel plates. Furthermore, this work aims to predict the yield stresses and strain using ABAQUS to simulate the proposed fabrication process of the MURR DDE plates, accompanied by discussion over the stresses and strains as to their relation to nuclear fuel performance and the impact they will have during early irradiation.

ABAQUS↗
DOE OSTI2023

Influence of processing on secondary phase formation and microstructural evolution at U-10Mo alloy and Zr interlayer interfaces

Thermo-mechanical processing of uranium-10 wt. % molybdenum (U-10Mo) fuel plates leads to microstructure changes at the U-10Mo/Zr interfaces. Secondary phases formed at this interface are particularly important to interfacial bond strength, process optimization, and maintaining structural integrity of the U-10Mo fuel plates during irradiation. In this work, we determined the phases and phase transformation products occurring at the interface of the U-10Mo fuel and Zr interlayer when the fuel plate is subjected to short and long hot isostatic pressure times. Interfacial morphology, structure and composition of phases formed, and relative hardness across the U-10Mo/Zr interfaces were studied using a multi-length scale, multi-modal characterization approach involving electron microscopy, atom probe tomography, and atomic force microscopy. Here, results highlight that the extent of phase transformations, secondary phase formation, and hardness variability across interfaces can be controlled by modifying processing parameters. Phase diagram construction and thermodynamic calculations were performed using the Thermocalc software to identify expected phases formed at interfaces during the maximum hold temperature of 560 °C experienced during HIP.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Tensile properties of co-rolled Zr/U-10Mo fuel at ambient and elevated temperatures

The United States High Performance Research Reactor (USHPRR) initiative is currently interested in replacing high-enriched uranium fuel in research reactors with high-assay low-enriched uranium fuel. Here, this work aimed to assist in that mission by providing tensile and bend test data on bare U-Mo and tensile test data on co-rolled Zr/U-10Mo plates at differing thicknesses (0.285 mm, 0.300 mm, and 0.465 mm) and temperatures (293 K, 473 K, and 623 K). Yield strength and ultimate tensile strength values are reported along with an in-depth comparison of how temperature, rolling direction, thickness, and the presence of the Zr-interlayer affect mechanical properties. Results showed that yield strength and ultimate tensile strength in Zr/U-10Mo foils lower as temperature increases by approximately 300 MPa from 293 to 473 K, and another 100 MPa from 473 to 623 K. Additionally it was found that with increased temperature, there is less differentiation in fuel necking and delamination behavior, as well as a greater perceived elongation to fracture. The elastic modulus of co-rolled mini-plates does not follow the volumetric rule of mixtures in terms of averaged strength and is actually lower than the reported moduli of both Zr and U-10Mo.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

BWR Spent Nuclear Fuel Acquisition and Testing to Support DOE-NE High Burnup Spent Fuel Data Project

The Office of Spent Fuel and Waste Disposition (SFWD) within the US Department of Energy (DOE) Office of Nuclear Energy (NE) established the Spent Fuel and Waste Science and Technology (SFWST) campaign to conduct research and development (R&D) activities related to the storage, transportation, and disposal of spent nuclear fuel (SNF) and high-level radioactive waste. The SFWST program was created within SFWD to address issues of extended or long-term SNF storage and transportation. Some near-term objectives of SFWST are to use a science-based, engineering-driven approach to: Support the enhancement of the technical bases to support the continued safe and secure dry storage of SNF for extended periods; Support the enhancement of the technical bases for retrieving SNF after extended dry storage; Support the enhancement of the technical bases for transporting high burnup (HBU) fuel and transporting low burnup fuel and HBU fuel after dry storage DOE-NE, in partnership with the Electric Power Research Institute, developed the High Burnup Spent Fuel Data Project to perform a large-scale demonstration and laboratory-scale testing of HBU pressurized water reactor (PWR) fuels (exceeding 45 gigawatt-days per metric ton of uranium [GWd/MTU]). Under this project, 25 sister rods—which are rods that have the same design, power histories, and other characteristics—were removed from assemblies at the North Anna Nuclear Power Station and sent to Oak Ridge National Laboratory (ORNL) in January 2016. ORNL performed detailed nondestructive examination (NDE) on all 25 rods. The NDE consisted of visual examinations, gamma and neutron scanning, profilometry and rod length measurements, and eddy current examinations. After completing the NDE, 10 of the sister rods were delivered to Pacific Northwest National Laboratory (PNNL) in a NAC International, Inc. legal-weight truck cask in September 2018 for destructive examination (DE). To date, SFWD work has focused on the PWR fuel that is part of the Sister Rod Test program. No boiling water reactor (BWR) fuel has been tested in the program, and the data needs that were identified for the PWR fuel have not been collected for BWR fuel. The goal to obtain six to nine BWR rods and test them at ORNL will support closing this important data gap. BWR fuel comprises approximately 56% of the total fuel assemblies currently in storage at nuclear power plants in the United States. BWR nuclear fuel and cladding designs and manufacturing are significantly different from PWRs. Differences include the following: BWR fuel pellets are larger than PWR pellets; Variations of Zircaloy-2 (including liners) are used instead of the Zircaloy-4 cladding materials used in PWRs; Clad manufacturing and stress-relief processes are different between PWRs and BWRs; The fuel rod dimensions are different because larger rod diameters and thicker cladding are used in BWRs; BWR fuel typically has lower internal rod pressures and sees vastly different operating conditions than PWR fuel (i.e., two-phase flow); BWR assemblies are “canned,” meaning each assembly is surrounded by a metal fuel channel; BWR cladding is often composed of an inner pure Zr liner that has widely different mechanical properties than the Zircaloy-2 alloy and exhibits a stronger affinity for hydrogen; The BWR SNF generally has more total hydrogen in the cladding/liner than typical PWR fuel; The construction of the PWR and BWR assemblies is vastly different; BWR rods are solidly attached to the assembly nozzles and experience a much different vibration and shock load than PWR rods, which are “floating” within a grid system attached to guide tubes, and the rods sit loosely on the bottom end plates. These numerous differences will affect the way the BWR SNF responds under dry storage preparation processes (e.g., vacuum drying) and during transportation. The results collected in the PWR experimental program must be compared with a subset of similar data collected on BWR SNF to establish a technical basis for whether the larger PWR database is sufficient to bound the BWR SNF end-of-life conditions as is currently assumed for several fuel/clad properties. Changes that occur in both fuel types at HBU could exacerbate any mechanical property differences. As the fuel burnup increases, several changes occur that might affect the performance of the fuel, cladding, and assembly hardware in storage and transportation. These changes include increased cladding corrosion layer thickness, increased cladding hydrogen content, increased cladding creep strains, increased fission gas release, and the formation of the HBU structure at the surface of the fuel pellets. The Nuclear Regulatory Commission (NRC) limits the current maximum rod-averaged burnup to 62 GWd/MTU due to these changes and the lack of data at higher burnups.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗