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At least 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

HFIR LEU High Density Silicide Dispersion Optimized Design Neutronics Analyses with PHAME

A high-fidelity neutronics model of the Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) with the low-enriched uranium (LEU) high-density silicide dispersion Optimized fuel design was updated and analyzed to generate reactor physics-based metrics to support follow-on thermal hydraulic and transient analyses of this design. The Python HFIR Analysis and Measurement Engine (PHAME) was also updated to enhance the automation capabilities of the framework developed and maintained to perform these reactor physics modeling and simulation efforts. The automated framework significantly increases the efficiency and reproducibility to design and thoroughly analyzes HFIR LEU core designs, changes, and uncertainties. Reactor physics metrics evaluated include but are not limited to fuel depletion, cycle length, fission rate density distributions, axial power peaking factors, kinetics data, reactivity coefficients, control element worths, heat deposition rates, and decay heat. These neutronics results provide essential input to follow-on steady state thermal, thermal hydraulic and reactor transient analyses, which are subject of other reports. The Optimized design operates at 95 MW to maintain HFIR’s current highly enriched uranium core performance level at 85 MW.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Participation in and Assessment of the Second DNCSH Public Workshop

The DOE/NRC Criticality Safety for Commercial-Scale HALEU Fuel Cycle and Transportation (DNCSH) project was established through the Inflation Reduction Act of 2022 (H.R. 5376) to support the US Nuclear Regulatory Commission (NRC) and industry in addressing critical experiment validation gaps that impede the licensing basis and regulatory approval of high-assay low-enriched uranium (HALEU) operations. An initial public workshop was held in February 2024 to address HALEU transportation validation gaps. The resulting call for proposals was released in April and resulted in funding for the execution and/or evaluation of 16 critical experiments. A second public workshop was held in August 2025 to address facility and operational validation gaps, precluding a second call for proposals. A list of attendees is provided in APPENDIX A, Table A-1. A total of 319 participants joined the meeting, which was hosted online via Microsoft Teams as well as in person. The slides from the meeting were uploaded online to the NRC’s Agencywide Documents Access and Management System (ADAMS). The meeting agenda is provided in Table 1-1. In preparation for the meeting, a study was performed to examine expected fissile forms for the fuel cycles of various fuel types at different stages of production and the apparent validation gaps. The resulting report, titled “Benchmark Gap Assessment for the Manufacturing of High-Assay Low-Enriched Uranium Fuels,” provided the foundation for the discussions that took place during the workshop. The discussions and the validation gaps in the report were used to develop the second call for proposals. The present report presents the feedback received before, during, and after the second workshop. All the data presented are based on voluntarily self-reported identification, opinions from workshop participants, and survey responses and are assumed to be as accurate as practically reasonable. The discussions during the workshop and the subsequent survey responses were intended to direct attention to industry-specific areas of interest and to collect feedback on the work performed to date by the DNCSH project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

HEU Metal Delayed Critical Experiments with 10 to 19 Inch Thick Graphite Reflectors

Approximately 100 graphite-reflected highly enriched uranium (HEU, 93.14 wt % 235 U) metal annular and cylindrical critical experiments were performed in the early 1960s at the Oak Ridge Critical Experiments Facility (ORCEF). This report presents details from experiment logbooks, experimental data sheets and the author's memory for 44 HEU metal (93.14 wt % 235 U) critical assemblies with graphite reflectors varying from 10 to 19 in. thick, outside diameters varying from 7 to 15 in., inside diameters varying from 7 to 13 in. and critical HEU metal masses varying from 20.4 to 69.0 kg. The data from the 44 experiments described in this report are acceptable for use as criticality safety benchmark experiments for the International Criticality Safety Evaluation Program (ICSBEP) once the uncertainty analysis on the measured k eff is completed. Based on previous ICSBEP benchmarks with this HEU metal at ORCEF, the uncertainties in the measured k eff are expected to be as low as ±0.0004. Preparation of this report is part of an effort at Oak Ridge National Laboratory (ORNL) to document more than 15 undocumented series of critical and subcritical experiments enumerated in Critical and Subcritical NEA Benchmark Possibilities for Measurements at ORCEF and Other US DOE Facilities (Mihalzo, ORNL/TM-2019/1188, 2019) and performed by ORNL at ORCEF and other US Department of Energy critical experiments facilities. More than 500 operational days of critical facility time were used, not including setup and dismantlement time. This documentation for a part of one series of graphite reflected highly enriched uranium metal critical experiments, that used 50 operational days of ORCEF time, was performed using funding received from the DOE Office of Nuclear Energy’s Nuclear Energy University Programs at the University of Tennessee Nuclear Engineering Department. This documentation was also supported by the Nuclear Criticality, Radiation Transport, and Safety programs at ORNL.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Godiva IV central cavity neutron environment characterization with threshold neutron detectors

Godiva IV is a cylindrical fast burst reactor comprised of approximately 65 kg of highly enriched uranium that is operated by Los Alamos National Laboratory and sited at the National Criticality Experiments Research Center at the Nevada National Security Site in Nevada in the United States. Godiva IV is typically operated at delayed critical and in the regime spanning from sub-prompt to super-prompt bursts. Godiva IV is used for sample irradiations, criticality safety demonstrations, dosimetry studies, and for studying super-prompt behavior. In preparation for both an upcoming experiment to reduce uncertainties in the prompt fission spectrum for 235U using threshold neutron detectors, and for future research using Godiva IV, it was desired to exercise the process of the selection of threshold neutron detectors/activation foils, radiation metrology, and the subsequent adjustment of the neutron spectrum. For this exercise, nine high purity threshold neutron detectors/activation foils were irradiated in a Godiva IV burst. The foils were then analyzed using a high-purity germanium detector in the NCERC counting laboratory to determine end of irradiation specific activities for available IRDFF-II reactions. This work summarizes the Godiva IV foil irradiation, radiation metrology results, and adjusted neutron spectrum. The results of this exercise ultimately characterized the neutron environment inside the sample irradiation cavity inside Godiva IV to a higher degree than previously performed, informed decisions for the upcoming larger scale experiment, and will inform future neutron spectrum characterizations at NCERC.

Whitman, Nicholas H.↗
DOE OSTI2020

Preliminary Evaluation of Driver Fuel Plates and Reduced Sodium Temperature to Increase Fast Test Reactor Experimental Neutron Irradiation Capabilities

All sodium-cooled fast test reactors operated so far have used essentially the same fuel geometrical ar-rangement, namely cylindrical fuel pins assembled into a triangular array in hexagonal fuel assemblies. Fuel compositions and dimensions may vary but the overall geometrical pattern remains identical. Fur-thermore, most sodium-cooled reactors have operated, and still operate, with core-average inlet and outlet sodium temperatures of, respectively, 350-400°C and 500-550°C. The present preliminary study revisits the standard fuel assembly geometry as well as operating tempera-tures to evaluate potential impact on fast test reactor neutron irradiation capabilities. The reference fuel is U-20Pu-10Zr. Pin (reference) and plate geometries are considered together with two sets of core-average sodium inlet/outlet temperatures: 350/500°C (reference) and 200/400°C. More prototypic environments could be obtained inside ad-hoc irradiation testing vehicles that are independent from the reactor primary coolant. Everything else being the same, fuel temperature is significantly lower in plates than in pins. Therefore, plates could in principle accommodate higher plutonium content, thus, reducing the need for uranium enrichment or allowing the use of lower quality plutonium (U-xPu-10Zr thermal conductivity and solidus temperature decrease as x increases, hence, necessitating additional thermal margins). For the reference pin configuration, lowering the inlet/outlet sodium temperatures from the reference 350/500°C down to 200/400°C provides additional thermal margins that can be used to increase the peak fast flux from about 4.5 × 10 15 n/cm 2 -s to 6 ×10 15 n/cm 2 -s for the same core power of 300 MW. Assuming 300 Equivalent Fuel Power Days (EFPD) of operation per calendar year, a (steel) test article could accu-mulate up to 75 dpa/year. The use of fuel plates provides even more thermal margins which may allow the peak fast flux to reach values as high as 8 × 10 15 n/cm 2 -s for both sets of inlet/outlet temperatures. In this environment, a test article could accumulate up to 95 dpa/year assuming 300 EFPD/year, hence greatly accelerating irradiation testing. Allowing a fast test reactor to operate over a wide range of inlet/outlet temperatures could add significant flexibility to its neutron irradiation capabilities. Steel present in the driver fuel assemblies is not expected to accumulate more than 90 displacements per atom (dpa) over its lifetime. Hence, steel alloys that are not suited for very high dpa but can operate over a wide range of temperatures (from low to high) would be appropriate for this application. Finally, because pin and plate assemblies fit on the same grid plate, a fast test reactor could in principle start-up with standard fuel pin assemblies and, later on, move to a plate-type Mark-II fuel to further in-crease its neutron irradiation capabilities while potentially reducing the level of uranium enrichment needed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Impact of Time Dependent Reactor and Sensor Physics on Core Power Synthesis

Online synthesis of the power distribution is critical in the operation and control of nuclear power reactors to ensure that the core is operating within safety margins, and to provide essential knowledge associated with the burnup of the fuel. In light water reactors (LWRs), power synthesis is achieved by using some a priori knowledge of the state of the reactor core and updating based on the signals coming from in-core sensors—namely, self-powered neutron detectors (SPNDs). This report aims to study the effects of fuel burnup and sensor degradation on the ability to accurately synthesize the power distribution in a LWR. Several modeling tools were used to simulate power synthesis based on the responses of SPNDs, with emitters made out of Rh or V. A representative pressurized water reactor low-enriched uranium (LEU) core was modeled using the Polaris/Purdue Advanced Reactor Core Simulator (PARCS) approach. The Monte Carlo N-Particle Transport 6 (MCNP6) code was used, as well, to calculate response functions between different segments of fuel to individual SPNDs; this is a crucial parameter for power synthesis. The Oak Ridge Isotope GENeration (ORIGEN) package in the Standardized Computer Analyses for Licensing Evaluation (SCALE) code was used to model the time-dependent isotopic transmutation in the SPND emitters. All these data were fed into a custom code that enacted the point-based iterative (PBI) method to simulate power synthesis. Developmental work was also performed on high-fidelity SPND models in the GEometry ANd Tracking 4 (Geant4) code, which enables higher-accuracy modeling of the current responses from SPNDs. In this work, five sets of time-dependent power synthesis test cases were conducted. In these test cases, systematic changes in the input conditions enabled an analysis of the effect of (1) slightly inaccurate a priori power distribution assumptions with respect to fuel burnup, (2) highly inaccurate a priori assumptions with respect to fuel burnup (such that burnup is not included in the a priori assumed distribution), and (3) differences between Rh and V SPNDs in terms of downstream consequences of the transmutation in the emitters. The authors discovered that one may permissibly have slightly inaccurate a priori assumptions of the fuel burnup (such that the level of burnup may be slightly under- or over-approximated by the accumulated burnup in approximately 9.3 full power days), but to not account for burnup at all in the a priori assumption leads to severe levels of error, approaching 25% at maximum. The authors also discovered that V SPNDs are extraordinarily robust in the low-enriched uranium fuel cycle considered in this modeling work, whereas Rh SPNDs undergo significant transmutation that can result in large errors in the synthesized power distribution.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Flattop Benchmark CED-2 (IER-423)

The final design for the Flattop HEU Core Benchmark is presented in this report. Physical measurements will be performed on Flattop components using modern instruments and techniques to define uncertainties necessary to meet contemporary ICSBEP standards. A preliminary re-evaluation of ICSBEP HMF-0028 (Flattop HEU Core) led to the development of IER-423 CED-1, wherein total relative uncertainties in $k_{eff}$ were provided based on a new detailed model. The estimated uncertainty on the total uncertainty is +100%/-0%, given a lack of documented uncertainties on mass, dimensions, and densities of predominantly the natural uranium (NU) and highly enriched uranium (HEU) components.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Impacts of LEU+ and HBU Fuel on Decay Heat and Radiation Source Term

Using low-enriched uranium with 235 U enrichment slightly greater than 5.0 wt % (LEU+) fuel is desirable for light water reactors (LWR) because it can enable longer cycles and/or smaller fresh fuel batches. This report analyzes the impact of LEU+ on back-end considerations, such as its impact on the spent fuel pool, and on the thermal and shielding performance of dry storage systems. For both pressurized water reactor (PWR) and boiling water reactor LEU+ fuel, this report concludes that additional cooling time or regionalized loading plans are required to maintain the same level of thermal and shielding performance of dry storage systems loaded with LEU fuel. For PWR fuel, this report concludes that the impact of LEU+ fuel on the time-to-boil of a spent fuel pool (SFP) is minimal and does not require considerable changes in SFP management.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Protecting People and Planet

To produce energy, civilian nuclear power plants require fuel. The majority of existing plants rely on solid ceramic fuels that contain concentrations of about 3 to 5% of the element uranium-235. This specific isotope is well suited to sustaining the type of chain reaction required for a nuclear reactor to harness the heat produced by nuclear fission. Title I of the United States Atomic Energy Act of 1954 defines “special nuclear material” as plutonium, uranium-233, or uranium enriched in the isotopes uranium-233 or uranium-235. This label identifies materials that at high concentrations can be used to produce nuclear weapons. The United States and broader international community are fully committed to ensuring the peaceful use of nuclear energy by committing to practices known as safeguards.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2025

Noninvasive Thermal Mass Flow Meter for Safeguards

This project addressed the long-standing nuclear safeguards problem of detecting in a timely manner the unauthorized production of significant quantities of highly enriched uranium in gas centrifuge enrichment plants (GCEPs).

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2026

Impact of Time-Dependent Reactor and Sensor Physics on Core Power Synthesis (Rev.1)

Online synthesis of power distribution is critical in the operation and control of nuclear power reactors to ensure that the core is operating within safety margins and to provide essential knowledge associated with the burnup of the fuel. In light-water reactors, power synthesis is achieved by using some a priori knowledge of the state of the reactor core and updating based on the signals coming from in-core sensors—namely, self-powered neutron detectors (SPNDs). This report examines the effects of fuel burnup and sensor degradation on the ability to accurately synthesize the power distribution in a pressurized water reactor (PWR), considering the typical low-enriched uranium (LEU, 3%-5% enrichment) fuel cycle as well as the higher enrichment LEU+ (5%-8% enrichment) fuel cycle. Several modeling tools were used to simulate power synthesis based on the responses of SPNDs, with emitters made out of Rh or V. A representative PWR LEU core was modeled using the Polaris/Purdue Advanced Reactor Core Simulator (PARCS) approach. The Monte Carlo N-Particle Transport 6 (MCNP6) code was used as well to calculate response functions between different segments of fuel to individual SPNDs; this is a crucial parameter for power synthesis. The Oak Ridge Isotope GENeration (ORIGEN) package in the Standardized Computer Analyses for Licensing Evaluation (SCALE) code was used to model the time-dependent isotopic transmutation in the SPND emitters. All these data were fed into a custom code that enacted the point-based iterative method to simulate power synthesis. Developmental work was also performed on high-fidelity SPND models in the GEometry ANd Tracking 4 (Geant4) code, which enables higher-accuracy modeling of the current responses from SPNDs. In this work, five sets of time-dependent power synthesis test cases were conducted. In these test cases, systematic changes in the input conditions enabled an analysis of the effect of (1) slightly inaccurate a priori power distribution assumptions with respect to fuel burnup, (2) highly inaccurate a priori power distribution assumptions with respect to fuel burnup (such that burnup is not included in the a priori assumed distribution), and (3) differences between Rh and V SPNDs in terms of downstream consequences of the transmutation in the emitters and the extended nature of the LEU+ fuel cycle in comparison with LEU. The authors discovered that one may permissibly have slightly inaccurate a priori assumptions of the fuel burnup (such that the level of burnup may be slightly underapproximated or overapproximated by the accumulated burnup in approximately 9.3 full power days), but to not account for burnup at all in the a priori assumptions leads to severe levels of error, approaching 25% at maximum (for LEU). The authors also discovered that V SPNDs are extraordinarily robust in both the LEU and LEU+ fuel cycles considered in this modeling work, whereas Rh SPNDs undergo significant transmutation that can result in large errors in the synthesized power distribution.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

VALIDATION OF KENO V.A AND KENO-VI IN SCALE 6.3 BETA 3 USING ENDF/B-VII.1 AND ENDF/B-VIII LIBRARIES

In early 2019, internal beta releases of SCALE 6.3 were identified for internal users, as well as a limited number of external users, to apply when testing key features. Recently developed AMPX-formatted libraries based on ENDF/B-VIII have also been generated for use in internal testing. In this effort, a criticality safety validation was performed using nuclear data libraries based on the ENDF/B-VII.1 and ENDF/B-VIII releases to quantify the changes caused by code development and to identify differences caused by the new nuclear data. As with past validations, the primary suite of experiments used is the Verified, Archived Library of Inputs and Data (VALID) maintained in the Reactor and Nuclear Systems Division at Oak Ridge National Laboratory. The VALID suite contains a total of 618 individual configurations, including systems with fast, thermal, mixed, and intermediate neutron spectra. Fissile species include a range of uranium enrichments, plutonium, mixtures of uranium and plutonium, and 233U. Metal, solution, and compound, mostly in pin array fissile forms, are included in the suite. The results presented here include the 252-group general purpose multigroup library based on ENDF/B-VII.1 and the continuous-energy libraries based on ENDF/B-VII.1 and ENDF/B-VIII.

Marshall, William (B.J.)↗
DOE OSTI2024

Flattop-HEU Benchmark Reevaluation Summary

The Flattop critical assembly was first constructed in the 1950’s at Los Alamos National Laboratory as a follow-on to the Topsy experiment. Flattop is composed of a sphere of special nuclear material (SNM) surrounded by a thick spherical reflector made of natural uranium (NU). Two SNM cores currently exist: a highly-enriched-uranium (HEU) core and a plutonium core. The reflector is composed of three parts: a stationary hemisphere and two movable quarter spheres. For fine reactivity control, there are three control rods of NU that are inserted into the stationary hemisphere from underneath the assembly. The final components that allow for reactivity adjustment are the glory hole pieces, mass adjustment buttons, and hemispherical caps. These pieces can be loaded in various configurations to change the available reactivity loaded in the system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Calculations of enrichment cascade performance using enrichment probabilities – a new method

A new method for calculating the performance of uranium enrichment cascades is presented. The method assigns a unique “enrichment probability” for each isotope to move up or down from the basic enrichment unit, allowing independent material balance calculations for each isotope. The formulation is much simpler than previous methods, which rely on isotopic ratios, and this method can be used when previous methods fail. This method gives the same results as published cases for 235 U enrichment and also gives good agreement with published data on minor isotopes. Some comparisons with measured data and other calculations are given. One case shows that the maximum 235 U enrichment that can be obtained by enrichment of reprocessed uranium (0.02% 234 U initial) is 82%. Another example shows a large difference in the minor isotopic content of material enriched in batches compared to continuous enrichment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Consequence analyses of sabotage-induced radiological releases in high-temperature helium-cooled prismatic microreactors

Here, this study analyzes the radiological dose consequences of sabotage-induced accidents at three high-temperature helium-cooled prismatic microreactors (HTPMs) with thermal power ratings of 1, 10, and 50 MWt. Each HTPM employs uranium oxycarbide tristructural isotropic fuel enriched to 19.75 wt% high-assay low-enriched uranium. Simulations were conducted to estimate reactor core inventory at the point of fuel discharge––when the effective multiplication factor reduced to less than 1––representing peak radionuclide inventory. Postulated sabotage scenarios leading to reactor shutdown were analyzed at two intervals: immediately post-shutdown (0 h) and 3 days after shutdown using the SCALE code for radionuclide inventories and the RASCAL tool for dose consequences. Results show that although HTPMs benefit from inherent safety features and robust fuel design, radiological consequences scale with reactor power because of increased source term inventories. Smaller microreactors exhibited proportionally lower dose consequences. To support the economic and regulatory feasibility of microreactor deployment, this study emphasizes the value of a risk-informed, performance-based approach, as supported by regulations like 10 CFR Parts 100 and 53 in the United States. Microreactor developers should perform site-specific assessments of potential sabotage or low-probability, high-consequence events, especially when considering minimal on-site or full off-site emergency response.

Consequence↗
DOE OSTI2020

Characterization of Uranium Foil Irradiations at the WSU TRIGA Reactor Using a New Reactor Model in SCALE [Slides]

A depleted uranium foil and a highly enriched uranium foil that were irradiated at the Washington State University TRIGA reactor were used to measure the relative production of 237 U and 239 U, two key isotopes of interest in nuclear forensics, as well as their decay daughters 237 Np and 239 Pu. Total fissions in the two foils were also estimated using 99 Mo measurement. A reactor model of the TRIGA was built in the SCALE neutron transport code using dimensions obtained from an existing MCNP geometry. The model was successfully used to simulate foil fissions and the reaction rate ratios in 238 U, using the laboratory-measured isotope ratios as a benchmark. The flux distribution in the foils was determined. Finally, using the continuous energy Tsunami-3d module of SCALE, a sensitivity/uncertainty analysis was performed to determine the effects on the reaction ratio caused by other reactions in the reactor and by uncertainties in the known cross-sections.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

R-Value Measurements Performed on Uranium Targets Irradiated with Fission Spectrum Neutrons FY 2021 for F2019 Project

The separation and characterization of two irradiated uranium targets, a depleted uranium (DU) and a highly enriched uranium (HEU) target, was conducted in April of 2021. The two targets were assembled at Los Alamos National Laboratory (LANL) and irradiated using the critical assembly at the National Criticality Experiments Research Center (NCERC). Splits of the dissolved targets were received by Pacific Northwest National Laboratory (PNNL) after which the PNNL and LANL teams chemically separated the solutions using independent separation schemes and analyzed the separated fractions for short lived actinides and fission products. Chemical separations at PNNL were traced with stable or radioactive tracers to allow for the determination of chemical yields, analyzing using either inductively coupled plasma optical emission spectroscopy (ICP-OES), inductively coupled plasma mass spectrometry (ICP-MS) or gamma emission analysis (GEA) depending on the nature of the tracer. Several other analytical techniques were used by PNNL including kinetic phosphorescence analysis (KPA) and thermal ionization mass spectrometry (TIMS) depending on the analyte’s need. Comparisons were made between current and historical PNNL and LANL, as well as literature values. Overall, there was agreement between the two laboratories for the bulk of analytes, with some notable exceptions such as 111 Ag, and 141,143,144 Ce. Included in these comparisons were the short-lived actinides 237 U, 239 Np, the fission products 89 Sr, 91 Y, 95/97 Zr, 99 Mo, 111 Ag, 115/115 mCd, 136/137 Cs, 140 Ba, 141/143/144 Ce, 147 Nd, 153 Sm, 156 Eu, and 161 Tb, providing both total atoms as well as the R-values. The data presented in this report represents the sixth NCERC irradiation of HEU and DU (FY13, FY14, FY15, FY17, FY18, FY21), and their subsequent separation and analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗