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All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Margin to Onset of Nucleate Boiling studies for MITR and NBSR Design Demonstration Experiments

The United State High Performance Research Reactor (USHPRR) program aims to eliminate more than 200kg of High Enriched Uranium (HEU) from commerce annually by converting five U.S. high-performance research reactors and one associated critical assembly to Low Enriched Uranium (LEU) fuel using a high-density alloy of uranium-10 wt% molybdenum (U-10Mo). The Massachusetts Institute of Technology Reactor (MITR) and the National Bureau of Standards Reactor (NBSR) are two of five research reactors selected for this program. Previous studies have addressed the thermal-hydraulics performance of the fuel Design Demonstration Elements (DDE) for these reactors under conservative assumptions in the Belgian Reactor (BR)-2. This report extends these previous studies by analyzing the margins to Onset of Nucleate Boiling (ONB) for these two DDEs under the conservative conditions used in the previous thermal-hydraulics analyses

42 ENGINEERING↗
DOE OSTI2023

Preliminary NBSR Design-Demonstration Element Thermal-Hydraulics and Structural Analyses

1.1 OBJECTIVE The United State High Performance Research Reactor (USHPRR) program aims to eliminate more than 200kg of High Enriched Uranium (HEU) from commerce annually by converting five U.S. high-performance research reactors and one associated critical assembly to Low Enriched Uranium (LEU) fuel using a high-density alloy of uranium-10 wt% molybdenum (U-10Mo). The National Bureau of Standards Reactor (NBSR) is one of five research reactors selected for this program. The objective of this report is to provide preliminary thermal-hydraulic and mechanical analyses of the hydrodynamic effects in the NBSR Design Demonstration Element (DDE) under conservative approximations for the plate power distribution. This report provides details on the modeling approach and the simulation results obtained, including pressure, flow velocity, temperature, and oxide layer over the design demonstration experiment for the irradiation cycles in the Belgian Reactor (BR)-2.

42 ENGINEERING↗
DOE OSTI2020

Irradiation Demonstration Element Design Parameters for MURR LEU U-Mo Fuel Conversion

This report contains the results of reactor design and performance calculations for conversion of the University of Missouri Research Reactor (MURR ® ) from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the U.S. Department of Energy National Nuclear Security Administration (NNSA) Office of Material Management and Minimization (M3) Reactor Conversion Program at the Argonne National Laboratory and the MURR Facility.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

MITR Low-Enriched Uranium Conversion Fluid-Structure Interaction Preliminary Design Verification

The U.S. National Nuclear Security Administration (NNSA) Office of Material Management and Minimization (M 3 ) has developed and is pursuing an integrated approach to address the persistent threat posed by unintentional proliferation of nuclear materials. The NNSA M 3 approach reduces the risk of highly enriched uranium (HEU) and plutonium falling into the hands of non-state actors by minimizing the use of and, when possible, eliminating weapons-usable nuclear material around the world. In this geopolitical context, most research and test reactors, both domestic and international, have completed or started a program of conversion from the use of HEU to low-enriched uranium (LEU) as fuel. Conversion of civilian research reactors from HEU to LEU, and the return of the HEU to the country of origin, are important components of the NNSA non-proliferation program. Worldwide, 71 reactors have been converted to the use of LEU fuel, and an additional 32 have been confirmed to be permanently shut down. The U.S., with 20 reactor conversions, is among the 39 countries on six continents where conversions have occurred. With recent conversions in Ghana and Nigeria, an important milestone was reached in completing conversion of all reactors on the continent of Africa to LEU fuel. Africa thus becomes the third continent to have completed conversion of all HEU reactors to LEU, following Australia and South America

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Impacts of LEU+ and ATF on Fresh Fuel Storage Criticality Safety

The use of increased fuel enrichment, which is still in the realm of low-enriched uranium (LEU) fuel, has been of interest to commercial light water reactor operators as part of the next iteration in fuel cycle technological advances and research and development. Using increased enrichment fuel, or high-assay LEU (HALEU), in power plants has clear benefits for being able to load cores with additional power-producing fuel. Although HALEU enrichments can range up to 20%, the more guarded approach of investigating enrichments above current fuels within 10% enrichment is referred to as LEU plus (LEU+) to reflect the less drastic change in operating conditions and requirements and similarity to current fuel cycles. Of additional interest and increasing maturity is the incorporation of accident-tolerant fuel (ATF) concepts, which are also applicable to the current fleet. This class of technologies involves changes such as cladding (e.g., chromium coating or FeCrAl) and fuel composition (e.g., chromia dopant) alterations to demonstrate improved fuel performance under accident scenarios. The ability to properly store fuel before and after residence time in the reactor is crucial to plant operation. Typically, this is done in either a new fuel vault (NFV) or spent fuel pool (SFP). Storing, loading, and unloading dozens of fuel assemblies within the same general area provides opportunities for obvious criticality concerns. These concerns are addressed with regulations to the subcriticality margin that the NFV and SFP must maintain in certain conditions. Adopting LEU+ fuel results in inherent reactivity increases, which are extremely relevant for safe fuel storage. Therefore, a clear understanding of the effects of LEU+ fuel and ATF on criticality safety margins to regulatory limits is required, as well as an understanding of the degree of absorber crediting under normal and accident conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Micro X-ray computed tomography examination of mini plate fuel with hot isostatic pressed aluminum cladding

In order to minimize proliferation risks and improve security of nuclear material, the United States high performance research and test reactors (USHPRR) program is tasked with converting nuclear reactors that are fueled with highly enriched uranium (HEU) fuels to operate with low-enriched uranium (LEU) nuclear fuels. One favorable LEU fuel configuration is plate fuel with a metallic uranium-molybdenum foil clad within an aluminum alloy (AA 6061). In this fuel, the aluminum cladding is bonded with a hot isostatic pressing (HIP) method to seal the cladding around the fuel meat. However, the HIP process parameters influence the cladding performance, as a defective or incomplete bond can cause a pathway for corrosion. Micro X-ray computed tomography (XCT), a nondestructive technique that provides volumetric imaging, can be applied to inspect fuel plate cladding at the engineering scale. In this work, XCT methodology was developed and successfully utilized to not only observe the bond line of unirradiated mini fuel plates, but to also identify subsurface abnormalities in the plates’ cladding. Importantly, in future work, this technique could be applied to fuel plates, pre- and post-irradiation, to quantify irradiation effects on cladding defects and bond line integrity.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Evaluation of flow-induced plate deflection for University of Missouri research reactor low-enriched uranium fuel element

The University of Missouri Research Reactor (MURR), located on the campus of the University of Missouri in Columbia, Missouri, is one of the six United States (U.S.) High Performance Research Reactors (USHPRR), including one critical facility, that are actively collaborating with the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA) Office of Material Management and Minimization (M3) Reactor Conversion Program to convert from highly enriched uranium (HEU, ≥20 wt% U-235) fuel to low-enriched uranium (LEU, <20 wt% U-235) fuel. A new type of very high-density LEU fuel based on a monolithic alloy of uranium and 10 wt% molybdenum (U-10Mo) is expected to allow conversion of some USHPRR, including MURR. In the design of its fuel elements, MURR is using thin parallel curved fuel plates separated by coolant channels. In this work, fluid-structure interaction (FSI) analysis of the MURR LEU fuel element is performed at the element level (as compared to the plate level analysis), which models all components of the LEU fuel element, including fuel plates and the supporting structures. Therefore, the effect of supporting structures on the flow distribution within the element and the fuel plate deflection are evaluated. In addition to the element nominal flow rate and dimensions, the tolerances in the geometry of the coolant channel and plate thickness, the effect of a comb on plate deflection, and the uncertainty of the flow rate per element are evaluated. For the LEU fuel plates, which are thinner than the current HEU plates, the predicted plate deflection is found to be small compared to the fabrication and assembly tolerances. Thus, the FSI-induced deflections are not expected to noticeably reduce the coolant flow rate or predicted safety margins in the limiting channels for the MURR LEU fuel element. In addition to the simulation work, a hydraulic performance test of the MURR LEU fuel element is currently being planned to support conversion to the use of LEU fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

PWR Core Analysis for Cycle Extension and Uprates with LEU+ Accident Tolerant Fuel and 80 GWd/Tonne Burnup Limit

The U.S. Nuclear Regulatory Commission has recently drafted a rule enabling fuel burnup increase in light water reactors up to 80 GWd/t. In conjunction with use of fuel enrichment up to 10%, and accident tolerant fuel (ATF), this is anticipated to facilitate 24-month cycles in PWRs, along with further power uprates. In this paper, PWR core analysis is performed for 20% increased PWR power output along with cycle extension up to 24 months, in combination with use of chromia-doped fuel and chromium-coated clad, considered to be the most near-term ATF concepts. In combination, these lead to challenging conditions with a core average discharge burnup of up to ~74 GWd/t, challenging even the 80 GWd/t burnup limit. Analysis is performed using the 2-step method with POLARIS (within SCALE) used for the lattice calculations and PARCS for the core calculations. Core designs are first baselined for current operating conditions (LEU, 62 GWd/t discharge burnup limit) and then derived that meet cycle constraints on power distribution and the updated lead pin discharge burnup limit while maintaining at least two batches of fuel in the core. Gadolina loadings in fuel pins of up to 8% are used, with enrichment zoning both within the core and, to a limited extent, within assemblies. Here, doped fuel with coated cladding can utilize the same core designs as the reference UOX cores, exhibiting slightly lower burnup due to higher fuel density, which also offsets the slight reactivity penalty from the doping and coating. For the analysis performed here, doped fuel enabled a core with 24-month cycle and 20% uprate to stay within the 80 GWd/t lead pin discharge burnup limit.

LEU+↗
DOE OSTI2020

Review of the Technical Basis for Properties and Fuel Performance Data Used in HEU to LEU Conversion Analysis for U-10Mo Monolithic Alloy Fuel

This report provides the technical basis for properties and fuel performance data used in conversion analysis for U.S. High Performance Research Reactors (USHPRR) that will convert from highly enriched uranium (HEU) to low-enriched uranium (LEU) using a new U-10Mo monolithic alloy fuel that is being qualified. The conditions that the fuel experiences changes between an HEU and LEU fuel element design due to many causes, including the density of the fuel, the presence of U-238 resonant absorber, changes to the plate and coolant channel dimensions, and changes in fuel management due to reactivity or optimization. These types of changes have been documented in operational and safety analyses conducted for conversions over decades for over 70 reactors. These conversions have all, or almost all, required recalculation of safety-related values as well as establishing operational characteristics of the core, including power distribution, reactor core power level, and cycle length between required fuel management. An overall objective of conversion is to change the reactor core design as little as possible while maintaining the reactors’ scientific, isotope production, medical, and engineering missions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Impacts of Irradiation Structural Behavior on Thermal Hydraulics Safety Analysis to Support MURR LEU Conversion

The University of Missouri Research Reactor (MURR) located in Columbia, Missouri is one of six U.S. High Performance Research Reactors (USHPRR), including one critical facility, that is actively collaborating with U. S. Department of Energy (DOE) National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Office of Reactor Conversion and Uranium Supply to convert from the use of highly enriched uranium (HEU; ≥20 wt% U-235) to low-enriched uranium (LEU; <20 wt% U-235) fuel. A new type of very high-density LEU fuel based on an alloy of uranium and 10 wt% molybdenum (U-10Mo) is expected to allow the conversion to LEU of MURR, as well as four other USHPRR. MURR has been working with the Reactor Conversion Pillar at Argonne to perform fuel element design and fuel cycle performance analyses, steady-state thermal hydraulics safety analysis, and accident safety analyses in preparation for the conversion of MURR and to support a preliminary safety analysis report for conversion to LEU fuels. Subsequent analyses have also been performed, including transition cycles where all-fresh LEU fuel elements are introduced upon conversion and progressing through reactor operations the core is brought to equilibrium. Thermal hydraulics safety analyses performed as part of the above have employed an assumption on channel gap reduction due to burnup-related phenomena including fuel swelling, irradiation creep, and oxide layer buildup. Recently, a series of structural analyses have been performed on the MURR LEU fuel plates and an element due to significant differences between the plate and element designs of the MURR HEU and LEU fuels. In addition, NUREG-1537 indicates that structural phenomena are to be evaluated. Two separate types of structural analyses were performed for the MURR LEU fuel element: fluid-structure interaction (FSI) and irradiation thermo-mechanical. The FSI analysis evaluated the effects of hydraulic forces on the MURR LEU fuel element to quantify the flow-induced plate deflection, and a minimal impact to the channel gap thickness was predicted under prototypic and bounding conditions. The irradiation thermo-mechanical analysis evaluated the effects of fuel swelling, irradiation creep, and thermal expansion for the MURR LEU plates and the element for prototypic thermal and irradiation conditions based on a high-fidelity approach multiphysics approach. Overall, this thermo-mechanical analysis predicts smaller gap thickness changes in previously limiting regions. Larger changes are predicted in the middle of channels, and for end channels where power density is not typically a maximum. An additional thermo-mechanical analysis was performed for the outermost HEU fuel plate, which showed a similar magnitude of deflection as the outermost LEU plate. Due to substantial differences between the channel gap reductions assumed for the previous safety analyses and those predicted by the irradiation thermo-mechanical analysis, a need to evaluate their impact on the thermal hydraulics safety analyses arose. This report presents the results from the steady-state safety analyses for normal operation as well as the accident analyses for the two most limiting accident scenarios.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Burnable absorbers in nuclear reactors – A review

Burnable absorbers can benefit nuclear reactors of virtually any design by providing reactivity control for extended fuel cycles, tritium production, burning of long-lived radionuclides, and reactor safety. When selecting the ideal burnable absorber type and its design, one must consider the resulting impact on the reactor’s fuel cycle design and cost, reactivity, thermal hydraulics, manufacturing, and radiation response. These selection criteria, as well as neutronic and thermophysical material property requirements, may be vastly different depending on whether the burnable absorber is intended for use in a commercial water-cooled reactor, a research reactor, or a next-generation advanced reactor system. A recent integration, synthesis, description of past and present technologies, and identification of existing gaps and areas of future research is lacking on these important topics. Here, this paper includes a fundamental description of the use of burnable absorbers and their impacts on reactivity, absorber depletion, self-shielding, basic thermophysical properties, and the use of burnable absorbers in next-generation nuclear applications..

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Light Water Reactor LEU+ Lattice Optimization

Commercial light water reactor (LWR) operators and fuel vendors in the United States are exploring potential changes to nuclear fuel that include low-enriched uranium plus (LEU+) designs to further improve operational economics (e.g., extend cycle length). LEU+ fuel is fuel with a maximum enrichment between 5 wt% and 10 wt% 235 U; it allows for higher assembly burnup but likely requires additional reactivity control, e.g., increased burnable absorbers. This report examines possible LEU+ fuel lattice design changes using the lattice physics code, SCALE/Polaris. An optimization driver called the Metaheuristic Optimization Tool (MOT) is used to automate domain space exploration and optimization of LEU+ lattice designs. Heuristics from previous LWR lattice optimization studies were used to construct the objective function and define the domain space for optimization. This work successfully demonstrated that the optimization algorithms of MOT can generate feasible, nonproprietary LEU+ lattice designs (GE14 10 × 10 and Westinghouse 17 × 17) that meet the constraints of traditional LWR lattices while extending cycle length.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Design and full core fuel performance assessment of high burnup cores for 4-loop PWRs

Increasing the fuel discharge burnup of current light water reactors (LWRs) promises reductions in fuel cycle and/or operations costs. By assuming a constant core power density, the economic gain is enabled by better fuel utilization and/or an increased capacity factor. In this effort to investigate greater than 62 MWd/kgU maximum rod average burnup for 110+ kW/l core power density, two core designs have been developed for a standard 17x17, 193 fuel assemblies pressurized water reactor (PWR). The levelized unit cost methodology is employed to evaluate fuel cycle, operation and maintenance, and capital cost impacts and to examine the economic viability of both core design pathways. Core design and optimization are performed using the commercial STUDSVIK code package. Fuel performance analysis is realized in full core configuration via auditing FRAPCON4.1, FAST1.2, and the high-fidelity code BISON. To provide a realistic assessment, the core design process takes into consideration best practices in current PWR core design. It features acceptable performance in terms of various core design constraints on maximum allowable peaking and boron concentration. Gadolinia (Gd2O3) is used as a burnable poison with a maximum of 9 wt% concentration while feeding 89 or 77 fuel assemblies in a 3-batch refueling scheme. Full core fuel performance simulation, which allows for characterization of relevant fuel temperatures, plenum pressures, stresses, and strains, is performed with respect to two bounding burnup levels. Such performance is potentially licensable for the 18-month high burnup core (<68 MWd/kgU peak pin), while it is more challenging for the 24-month high burnup core design pathway (<75 MWd/kgU peak pin). Maximum rod plenum pressure is identified as the most limiting fuel performance parameter. Here, while the scope of the present study focuses on the steady-state plus overpower conditions, the acceptability of the new discharge burnup has to be further assessed by considering uncertainties and impacts under accident scenarios in the future.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Post irradiation examination of a uranium-zirconium hydride TRIGA fuel element

Low-enriched (LEU) U-ZrH fuel, with a 235 U content less than 20% of the total uranium, is being evaluated for possible use in different types of reactors, including space nuclear systems, light water reactors (LWRs) and micro-reactors. As a result, it is beneficial to better understand the macrostructural and microstructural changes that occur in this fuel during irradiation. This paper reports the results of the post irradiation examination of an LEU U-ZrH fuel element (30 wt.% U, <20% 235 U) using neutron radiography, precision gamma scanning, chemical analysis, optical metallography and scanning electron microscopy combined with energy dispersive spectroscopy and wavelength dispersive spectroscopy, where the fuel element was irradiated in a Training, Research, Isotope, General Atomics (TRIGA) reactor. Results of microstructural characterization indicated some dehydriding and cracking of the U-ZrH fuel occurred during irradiation; an axial and radial burnup gradient existed in the fuel during irradiation, as measured by gamma scanning and chemical analysis; negligible microstructural changes transpired during irradiation, based on comparison of irradiated and as-fabricated U-ZrH fuel microstructures; and, negligible, fission product-rich, phases could be resolved in a U-ZrH fuel that was irradiated to a calculated 20% depletion of 235 U.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Local Power Impact Experiment Design for a New Fuel Type for use in the Advanced Test Reactor

The Advanced Test Reactor (ATR), and complimentary zero-power ATR Critical (ATRC) reactor, located at Idaho National Labs (INL), are undergoing conversion from Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU). Both have a variety of testing locations that can receive large variations in flux due to its unique serpentine design, consisting of five lobes (see Figure 1). Initial criticality and power distribution throughout the core are controlled by core-external outer shim control cylinders (OSCCs). Distinct test loops allow for testing at specific temperatures, pressures, and irradiation conditions. The ATR is one of the key nuclear engineering research and testing facilities within the DOE National Laboratory Complex, and the ATRC supports its operation [1]. Currently, the Office of Material Management and Minimization (M3) within the National Nuclear Security Administration of the DOE is working to convert the remaining research reactors, including the ATR, from 93% HEU fuel to 19.75% LEU fuel (LEU) to support non-proliferation [2]. Extensive materials testing at INL and internationally has demonstrated that a high-density uranium molybdenum (U 10Mo) alloy can meet the performance requirements of the remaining high powered research reactors. However, there are many technical challenges to address before the conversion to LEU can be successful, including the accurate characterization of the reactor core physics with LEU fuel. Reactor physics safety evaluations currently use Monte Carlo for the 21st Century (MC21), a continuous-energy Monte Carlo radiation transport code [3]. Existing MC21 models of the ATR and ATRC cores have a validation basis for use in neutronics analyses with HEU fuel. The models are used to support safety analyses that include comparisons to the safety requirements for the reactors. However, the use of the LOWE element in the ATR and ATRC is not currently covered by the current model validation basis. To deploy the new fuel type, extensive computational reactor physics support is necessary to support the use of LOWE in the ATR and ATRC. Therefore, LOWE requires a rigorous validation basis, aligned with that of HEU fuel, that takes advantage of the existing software tools and processes currently used for the ATR and ATRC. The experiment to validate of the MC21 models for determining power, the Power Impact Validation Experiment, will consist of two flux runs in the ATRC, one with fully HEU loading and one with a single LOWE element. Both flux runs will be instrumented with 20 sets of azimuthal fission wires and 3 sets of axial fission wires, as shown in Figure 4. Standard flux run methodology will be used [4]. Power Impact Validation Experiment data will be compared against MC21 calculated data, both for absolute fission rate accuracy and to determine the relative change in fission rates between the two runs. The results of the Power Impact Validation Experiment and subsequent evaluations will provide the validation basis for MC21 for use with LOWE elements. Key features of the Power Impact Validation Experiment include: (1) Two flux runs to allow for LOWE perturbed measurements to be compared to already validated measurements taken from a full core of HEU fuel, (2) Optimization of instrumentation to balance analytical needs with practical considerations (e.g., limited time window to count beta particles from fission products), and (3) Standard ATRC core loading, including both driver positions and flux traps, to minimize cost while remaining representative of typical ATR core loading.

42 ENGINEERING↗