ATR Conversion Update for USHPRR Stakeholders Meeting
This presentation provides on update on the ATR Low Enriched Uranium (LEU) Fuel Development project to the United States High Performance Research Reactor (USHPRR) Program.
42 ENGINEERING↗
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Joint Sample Analyses of Nuclear Forensic Materials Provided by the Republic of Kazakhstan: U. S. Laboratory Results
This document serves as an interim report to summarize sample analyses performed by Lawrence Livermore National Laboratory (LLNL) and Los Alamos National Laboratory (LANL) on a series of five nuclear forensics samples provided by the Institute of Nuclear Physics (INP) in the Republic of Kazakhstan. The sample set contains four uranium oxide powders and one low-enriched uranium fuel pellet. The samples were provided as part of a broader collaboration that involved a set of joint sample analyses conducted by INP and the US National Laboratories. The joint analyses are being conducted using well-developed analytical plans. These activities are designed to support the advancement of nuclear forensic science and capacity building in all three institutes in both countries. This report builds upon an earlier preliminary summary of the joint interactions and will be augmented by a final report. The final report will summarize the results and value of all sample analyses performed at the three institutes (INP, LLNL, and LANL). The final report will also provide an intercomparison of the results, analytical methods and best practices employed, as well as outline future collaborative nuclear forensics activities that are being developed in the Kazakhstan region.
and nuclear chemistry↗
Neutronics analysis of Full Size Plate–1 Experiment using MC21
Qualification and testing of Low Enriched Uranium fuel is desired so that it can be used in research reactors instead of High Enriched Uranium. The Full Size Plate - 1 (FSP-1) experiment involves irradiating a low enriched metallic fuel in the Advanced Test Reactor to assist in this general qualification. Prior to irradiation, an in-depth neutronics analysis is needed to ensure the safety requirements of the Advanced Test Reactor are met. To accomplish this, the MC21 software was used to analyze the FSP-1 experiment, which will allow a test run to be performed with this new fuel in the Advanced Test Reactor Critical facility. By verifying and validating the results of the MC21 software with comparisons to other codes and the preliminary irradiation in the Advanced Test Reactor Critical facility, the FSP-1 experiment will be ready to irradiate using the Advanced Test Reactor. This work is concerned with the neutronics analysis of the FSP-1 experiment using the MC21 software.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Radiation Dose Modeling for Niowave’s Accelerator Driven Uranium Target Assembly 3
Molybdenum-99 is a high-value radionuclide commonly used for medical purposes within the United States. The National Nuclear Security Administration (NNSA) seeks to reliably produce the radioisotope 99 Mo without the use of highly enriched uranium. NNSA’s Office of Material Management and Minimization (M3) provides funding and government laboratory expertise to private companies to expedite the production process domestically and currently funds designs that use low-enriched uranium or other 99 Mo production pathways. Several production designs are being explored across the industry, including uranium fission and photonuclear conversion of 100 Mo targets. Niowave Inc. seeks to produce 99 Mo via a high-energy electron accelerator that strikes a lead-bismuth eutectic target that ultimately produces a consistent neutron flux. The neutron flux then interacts in a subcritical reactor core configuration to produce fission in low-enriched or natural uranium targets. These fissionable targets are then processed to extract 99 Mo. The purpose of this work is to estimate the neutron and photon dose response across Niowave’s proposed facility for worker safety during operation. Owing to the size of the proposed Niowave facility and necessary shielding, unbiased Monte Carlo radiation transport is impractical, and variance reduction methods are required. This work focuses on the weight window variance reduction method to produce high confidence dose response results within a Monte Carlo radiation transport code. Specifically, an adjoint-informed weight window methodology was created to improve the dose response estimates for accelerator-driven subcritical reactor designs. This adjoint-informed methodology was implemented for Niowave’s proposed design and improved dose results at far-field locations across the facility. Acceptable dose rate contours for the proposed facility were generated across the facility and are presented in this work.
07 ISOTOPE AND RADIATION SOURCES↗
Depletion Benchmark of the AFIP-7 Experiment in the Advanced Test Reactor
Reactor physics depletion benchmarks for low-enriched uranium fuel are limited in number. In particular, there is very limited data for LEU benchmarks for U-10Mo (Uranium-10% Molybdenum) plate fuel developed for use in U.S. high-performance research reactors (USHPRR). USHPRR includes the Advanced Test Reactor (ATR), Advanced Test Reactor Critical Facility (ATR-C), High Flux Isotope Reactor (HFIR), University of Missouri Research Reactor (MURR), Massachusetts Institute of Technology Reactor (MITR), and National Bureau of Standards Reactor (NBSR) at the National Institute of Science and Technology. These reactors are fueled with high-enriched uranium dispersed fuel in a silicon/aluminum matrix. In support of conversion to a HALEU fuel, qualification of U-10Mo formed into a monolithic foil is being performed. Fuel qualification involves irradiated fueled specimens in the ATR. The irradiation tests provide an opportunity to benchmark depletion capabilities of reactor physics codes in support of the ATR operation, as well as develop benchmarks that can be used by other institutions to benchmark other reactor physics codes. This report documents the development of a benchmark model of the irradiation of the ATR Full -size plate In center flux trap Position 7 (AFIP-7) experiment.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Development of a thermal creep model for aluminum alloy 6061 cladding in U-10Mo monolithic fuel plates
Plate-type fuel elements consisting of a high-density, low-enriched uranium (LEU) U–10Mo-based fuel foil encapsulated in an aluminum alloy (AA) cladding are fabricated using the hot isostatic pressing (HIP) technique. During the HIP process, the fuel plate system is heated to 560 °C, then cooled to room temperature. This heat cycle significantly affects the mechanical properties of the aluminum cladding, and experimental investigations have shown that, post-HIP bonding, the mechanical properties of the aluminum cladding transition from those of AA 6061-T6 to something closer to the O temper. More specifically, the ultimate strength of the cladding decreases while its ductility increases, making it challenging to capture the changes in mechanical behavior and material properties. Understanding the residual stresses generated during the HIP process is critical for assessing the fuel plate’s integrity under various temperature, pressure, and irradiation. To simulate the HIP bonding process, the elastic, plastic, and thermal properties of the cladding are assumed to be similar to those of AA 6061-O temper. However, the primary challenge lies in the lack of available data for the creep model of the AA 6061 cladding during this transient process of HIP. The present study focuses on developing a computational model that predicts the creep behavior of the aluminum cladding in the fuel plates during the HIP process, as cladding creep significantly influences the residual stresses generated in U-10Mo fuel plates during HIP fabrication. Furthermore, as HIP bonding occurs at high temperatures that are nearing the melting point of aluminum, the present work considered a temperature-dependent Arrhenius-type creep model. In particular, a hyperbolic sine creep model is employed to estimate the creep properties of the as-fabricated aluminum cladding. In conclusion, the residual stresses predicted in the U-10Mo fuel when using the newly calibrated creep model closely align with the experimental measurements, validating the model’s accuracy.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Proceedings for the Workshop on Applied Nuclear Data Activities 2025
The 2025 Workshop for Applied Nuclear Data Activities (WANDA) covered four topic areas in nuclear data: Nuclear Data and Deterrence, Nuclear Data Prioritization for Fusion, High-Assay Low-Enriched Uranium and Novel Moderators for Advanced Reactors, and Data Preservation and Data Workflows. The intention of this workshop is to connect different communities that are invested in nuclear data and have their own unique sets of needs for the purposes of sharing information, fostering collaboration in areas of shared interest, and leveraging synergistic capabilities. The attendance of federal program managers at these workshops is essential in creating awareness of the needs of their respective communities and in providing information to better guide funding investments. In each of these topical sessions, a general description of the nuclear data needs and/or capabilities was presented, along with discussions of existing capabilities that could be leveraged, potential synergistic needs or resources, and challenges that must be overcome for the application space to progress. There are many synergistic nuclear data needs among these application spaces. The discussion largely focused on increasing the accuracy of the nuclear data and better quantifying the data uncertainties that have the greatest impact on applications. The full-day session on data processing and data workflows was by nature intended to be synergistic and applicable to all technical sessions at WANDA. A notable common theme that was highlighted across all sessions was the need for accelerated delivery of nuclear data products across complex and time-consuming workflows.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Rapid isotopic analysis of uranium microparticles via SP-ICP-TOF-MS
Inductively coupled plasma – time-of-flight – mass spectrometry (ICP-TOF-MS) was employed for the isotopic analysis of uranium particles of varying 235 U enrichment levels. Here, a single particle (SP)-based introduction scheme was employed such that individual particles, in a suspension, were analyzed. The uranium oxide microparticles were comprised of depleted uranium (DU, 235 U/ 238 U of 0.0017316(14)), natural uranium (NU, 235 U/ 238 U of 0.0072614(39)), and low enriched uranium (LEU, 235 U/ 238 U of 0.051025(15)). The percent relative difference of the SP-ICP-TOF-MS measured isotopic ratios compared to the expected values for the DU, NU, and LEU particle populations were 8.75, 0.12, and 1.23 %, respectively. After characterization, the DU and NU particles were doped within a complex sample matrix (Arizona Test Dust) containing Fe, Ti, Al, and Si particles, among others. Then, the suspension was analyzed via SP-ICP-TOF-MS and the detected particles were classified as DU or NU based on their measured 235 U/ 238 U ratio. In the same analysis, the matrix particles (i.e., Al, Fe, and Ti) were detected, demonstrating the simultaneous nuclide detection provided by the measurement platform. The presented SP-ICP-TOF-MS methodology for uranium particle characterization proved to be a high throughput method for detecting and isotopically discerning uranium particles with varying enrichment levels, in a complex matrix.
Stanberry, Jordan S. [Oak Ridge National Laborator↗
Kinetics and Mechanism of Ce(IV) Phase Transfer by the Neutral Extractants Tributyl Phosphate and N,N -Di(2-ethylhexyl)isobutyramide
Tributyl phosphate and N,N-di(2-ethylhexyl)isobutyramide (DEHiBA) are uranium-selective extractants that could be used to recover low-enriched uranium from commercial used nuclear fuel. The economic viability of liquid-liquid extraction processes for chemical separations depends on both thermodynamic and kinetic considerations. Here we used microfluidic droplet extraction to measure interfacial phase transfer rate constants for the extraction and stripping of Ce(IV), a non-radioactive actinide surrogate. The Damkohler numbers for each system were calculated and the phase transfer processes were determined to be diffusion-limited under all conditions, contrary to prior studies using constant interface stirred cells and single drop methods. A well-extracted 2:1 DEHiBA/Ce(IV) complex in the organic phase was determined from batch distribution studies, suggesting a different extraction mechanism for Ce(IV) compared with poorly extracted Pu(IV). Finally, phase disengagement rates were found to be rapid in both systems, in agreement with other studies.
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
Rapid 235 U/ 238 U determination by matrix assisted ionization–time-of-flight mass spectrometry
Matrix-assisted ionization (MAI) of inorganic analytes is a nascent research domain that holds promise for rapid, potentially facility-deployable analytical applications. Here, we present results of MAI uranium isotopic analysis ( 235 U/ 238 U) obtained on the timescale of minutes utilizing simple sample preparation and an ambient ionization time-of-flight mass spectrometer (ToF MS). Experimental MAI-ToF MS characterization of uranium Certified Reference Materials (CRMs) was used to establish method calibration and validate quantitative 235 U/ 238 U determination spanning depleted, natural, and low-enriched uranium isotopic compositions. Secondary standard analyses with total uranium mass loadings of 5–500 ng per analysis yield accurate calibrated 235 U/ 238 U results and relative uncertainties of 4.7–17.2% (approx. ±95% confidence level), with weighted-mean uncertainties approaching 1.5%. This method permits accurate determination of uranium isotopic composition in a sample with uranium content as low as 200 pg for equal atom 235 U: 238 U. Instrument detection limits constrain the minimum uranium mass required to identify the presence of highly enriched uranium (HEU ≥20% 235 U) as only 500 pg using the method presented here. MAI-ToF MS quantitation of relatively extreme isotope ratios ( 235 U/ 238 U ≤ 0.01) is limited by detection of minor 235 U (LoD 100 pg 235 U/analysis ≈ 10 ng total U/analysis), and subsequent method optimization is anticipated to further reduce these limits. These findings underscore the potential of MAI-ToF MS for isotopic characterization of uranium and other inorganic species for both basic and applied science.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Higher Power Design Concepts for NASA's Kilopower Reactor
The successful testing of the Kilopower reactor during the KRUSTY (Kilopower Reactor Using Stirling TechnologY) experiment significantly reduced the risk to fly fission power systems by demonstrating stable reactor operation through nominal and severe simulated mission scenarios. The experiment validated the neutronics, heat transfer, and power conversion systems needed for 1 kilowatt of electrical power production from the Kilopower reactor. The need for higher power reactors to support human exploration missions to the moon and Mars has become increasingly important due to the urgency to put boots on the moon by 2024 and have a sustainable presence in the following years. This desire has prompted NASA to continue the development of the Kilopower reactor to extend the power up to 10 kilowatts of electricity in support of a lunar base. These 10 kilowatt units are expected to be used as standalone units or be ganged together to create a modular power grid for propellent production, human habitats, and robotic exploration to name a few. The Kilopower reactor was originally designed to produce electrical power from 1 to 10 kilowatts using the same highly enriched uranium fuel, sodium heat pipes, and Stirling convertors at the proper scale. Consideration has also been given to the use of low enriched uranium fuel for these missions and will be studied along with the other aspects of the reactor. This paper will focus on the design concepts and trades associated with the scale up of the Kilopower power conversion system and heat transfer system to support human exploration of the moon and Mars.
Gibson, Marc↗
Calculations of Subcritical Limits in ANSI/ANS-8.1 [Slides]
ANSI/ANS-8.1 standard “Nuclear Criticality Safety in Operations with Fissionable Materials Outside of Reactors” provides subcritical limit (SCL) values for several single- and multiple-parameter limits for fissile nuclides, including, uniform aqueous solutions and metals containing nuclides of 233 U, 235 U, and 239 Pu, uranium-water lattices, homogeneous uranium-water mixtures of dry/damp oxides, uniform aqueous solutions of low-enriched uranium, and uniform aqueous mixtures of Pu(NO 3 ) 4 containing 240 Pu
ENDF↗
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↗
The University of Missouri Research Reactor (MURR) LEU Fuel Element Flow Test Preliminary Design
As part of the U.S. National Nuclear Security Administration’s (NNSA) mission to eliminate or minimize the civilian use of weapons-grade highly enriched uranium (HEU, ≥ 20 wt% U-235) fuels, the NNSA Material Management and Minimization (M3) Office of Reactor Conversion and Uranium Supply is collaborating with six U.S. high performance research reactors (USHPRR), including one critical facility, to convert from the use of HEU to low-enriched uranium (LEU, < 20 wt% U-235) fuel. Primary conversion objectives for the USHPRR are to develop LEU fuel element designs that will ensure safe reactor operations and maintain the existing experimental facilities performance. The work is being conducted through many interrelated activities by stakeholders across organizations.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Preliminary Benchmark Uncertainties for Deimos, a HALEU-Fueled and Graphite-Moderated Advanced Reactor Testbed
Many advanced reactor concepts will make use of various uranium fuels with levels of enrichment higher than previously seen in current light water reactors. In particular, High-Assay Low Enriched Uranium (HALEU), that is uranium enriched to 235 U ≈ 20 w/o%, is planned to be used in over ten new reactor concepts. HALEU is attractive for advanced reactors as it enables longer intervals between refueling. Unfortunately, little to no experience with HALEU is available in experimental literature raising concerns for not only licensing advanced re actors but also fabrication and transportation of HALEU fuels. This is where Deimos, a Los Alamos National Laboratory internal project, comes in. Deimos is a new critical experiment scheduled for FY24 at the National Criticality Experiments Research Center (NCERC). Deimos is a graphite moderated, graphite and beryllium reflected critical experiment making use of HALEU TRi-structural ISOtropic (TRISO) fuel from the Compact Nuclear Power System (CNPS). This transaction entails a brief description of efforts to benchmark Deimos for inclusion into the International Criticality Safety
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
TRISO-form HALEU-fueled Experiment for Transport Applications (THETA) Serpent 2 Benchmarking
A number of new advanced reactor designs use TRI-structural ISO-topic (TRISO) fuel. The uranium enrichment for the TRISO fuel is in the 5- to 19.75-weight percent range, which is considered High-Assay Low Enriched Uranium (HALEU). The Nuclear Regulatory Commission and vendors both have a strong interest in benchmarking these advanced reactor designs and having confidence in the predictive capabilities of the computer codes used to model them. In the recent years, Serpent 2 gained several capabilities focused on the reactor applications making it attractive for many reactor vendors.
Critical Experiment↗
Assessment of Uranium Isotope Impacts
This report examines the benefits and challenges associated with recycling recovered uranium (RU) from the used nuclear fuel (UNF) discharged from a once-through sodium-cooled fast reactor (SFR) with high-assay low-enriched uranium (HALEU) which has a higher residual U-235 content in the RU. However, there are three minor uranium isotopes that influence the value of the RU. There will be trace quantities of U-232 in RU, which can increase dose rates and, in turn, increase the cost of fuel fabrication. Additionally, the U-234 and U-236 isotopes are found in higher concentration in RU than in NU, and act as neutron poisons, requiring higher U-235 concentrations to compensate for the neutron poisoning effects of these isotopes.
07 ISOTOPE AND RADIATION SOURCES↗
Fuel Behavior Implications of Reactor Design Choices in Pressurized Water SMRs
Small pressurized water reactors (PWRs) can feature boron free operation, natural circulation mode, reduced height assemblies and/or long refueling cycles. This paper attempts to explore core design optimization for each of these design evolutions. In consequence, five core design layouts are developed incorporating boron free operation with continuous control rods insertion, natural circulation with low burnup/low power density design, natural circulation with high burnup/low power density design, forced circulation with standard core power density design, and forced circulation with high power density design. These cores’ performance is compared to a standard 4-loop PWR. The design process aims to improve the fuel cycle cost under safety constraints through core design optimization using CASMO4E/SIMULATE3 reactor physics codes and FRAPCON4.1 fuel performance assessment tool. Core modeling assumes standard 17x17 PWR fuel assemblies loaded with low enriched uranium (LEU) up to 5wt% or LEU+ (i.e., below 10wt% enrichment) pellets with gadolinium oxide (Gd2O3) as the burnable poison. Satisfactory core and fuel performances are obtained for all the designed cores under steady state and considered overpower transients. For low power density operation, long cycle lengths are achieved reaching a 2.5- and a 5-year cycles and peak rod-average burnup is pushed to 83 MWd/kgU. Other cycle lengths are maintained at 18 months. Boron free operation exhibits the ability to achieve longer cycle lengths at the cost of higher peaking factors leading to high local power and fuel temperatures which prevents sizable power uprates and is deemed uneconomical. Fuel assembly height reduction allows coolant velocity retrofit which enables higher core power density without violating structural integrity of the fuel assembly. As a result, a core power density of 123 kW/l is reached where total cladding hoop strain becomes the limiting parameter.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗