High-Assay Low-Enriched Uranium Research Articles

The Bimodal Epithermal Astronuclear Reactor (BEAR): A Long-Lived, Universal Space Nuclear Powerplant

A neutronic model of a Nuclear Thermal Propulsion reactor based on the NERVA Peewee design using High Assay Low Enriched Uranium (HALEU) fuel in a Zirconium Carbide matrix has been thoroughly characterised and continuously modified over the last few years at the CSNR. The initial design (Emu, for the flightless bird) created in the summer of 2020 and presented at the NETS 2021 conference, has undergone significant adjustments in this time. Two major investigations have been undertaken and evolved over time; one into the requisites for the provision of electrical power to the propelled spacecraft via the circulation of a noble gas mixture in a closed Brayton cycle, and another into the use of planetary regolith as a reflector, reducing neutron leakage at what would otherwise be the reactor’s end of life in vacuo, rekindling it for an extended period of time as well as providing a very easy ultimate disposal solution, replacing the control drums with rods. It was determined that there are no insurmountable obstacles to performance in either mode.

The weapons potential of high-assay low-enriched uranium

The weapons potential of high-assay low-enriched uranium

Startup and Multiphysics Analysis of a Compact kW-Class Thermal Spectrum Microreactor

Self-regulating compact nuclear microreactor concepts are being developed for use in space and at terrestrial remote sites. The current emphasis is on the development of high-assay low-enriched uranium–fueled reactors that rely on metal hydrides to achieve size, specific weight, and power parity with the legacy highly enriched uranium systems. In the case of terrestrial applications, metal hydride–moderated systems can also improve economic feasibility. The neutronic design of metal hydride–moderated high-temperature reactor cores is complicated by the need to optimize and synchronize delayed spatial and temporal reactivity feedback from outer reflectors. Zebra is a thermal spectrum proof-of-principle core designed for demonstrating the reactor dynamics of hydride-moderated spectrum reactors, which is studied in this work. The reactor, including the fuel, yttrium hydride moderator, heat pipes, central control rod, structural supports, and the beryllium reflector, weighs ~425 kg. The design shares several features common with nuclear criticality test geometries routinely used in National Criticality Experiments Research Center (NCERC) experiments such that a prototype of the reactor can be readily assembled and tested. Additional features, such as heat pipes and hydrogen barrier clads, were added to minimize the potential for large thermal gradients that in turn could induce hydrogen loss or structural deformation during extended periods of operation at ~1000 K. In this work, normal and off-normal state performance of the Zebra rector core was analyzed using MCNP and Abaqus-based Reactor Multiphysics (MARM) software. Up-to-date nuclear, thermal, and mechanical performance data were used to characterize the performance of the yttrium hydride moderator. Steady-state analyses established that at a postulated power between 20 and 50 kW(thermal), the reactor core is nearly isothermal irrespective of quality of conduction coupling between heat pipes and fuel plates. Additionally, heat pipe failure modes, including simultaneous failure of all heat pipes in a quadrant of a reactor, were examined to bound the maximum credible temperature spike in the reactor core for extreme off-normal operating conditions. Finally, we detail the startup design challenges for the hydride-moderated thermal core, and analyze load-following cases to achieve “self-regulation” using the Dynamic Analysis of Reactor Transients module of MARM. The reactor is self-regulating with a reactivity temperature coefficient of ~−1 pcm/K at the operating point based solely on nuclear cross-section feedback. It can be further strengthened using additional spectral shift neutron absorbers and incorporating design features that enhance core expansion. This work captures that hydride-moderated systems are feasible for various self-regulating applications once systematic checks are verified in order to achieve a well-engineered core design.

METAL: Methodology for liquid metal fast reactor core economic design and fuel loading pattern optimization

The design of Liquid Metal-cooled Fast Reactor (LMFR) cores encompass two levels of design. The first is the physical core design, which is concerned with the design of the fuel assembly geometry in the context of a full core. The second is the fuel loading design, which is an in-core fuel management problem concerned with the design of fuel enrichment and core loading pattern. In this paper, an optimized fuel design search is developed to improve core loading patterns. As part of the implementation, the multiphysics simulation suite LUPINE has been enhanced to allow for fuel shuffles and an in-core fuel management optimization methodology has been added with the objective of reducing the fuel cost. The design methodology is referred to as Methodology for Economical opTimization of Applied LMFR (METAL), and this methodology is demonstrated using the popular Super Power Reactor Innovative Small Modular (SPRISM) sodium-cooled fast reactor (SFR) as a reference core. One challenge currently facing the deployment of LMFRs is the availability of TRansUranium (TRU) and high assay low enriched uranium (HALEU) driver fuel. The two forms of fuel are expensive and not widely available, which poses a challenge on the startup of LMFR cores. To address the availability of driver fuel, and to lower fuel costs, low enriched uranium (LEU) blankets are investigated as replacements to the natural uranium or depleted uranium blankets typically used. The advantage of this solution is the wide availability of LEU and its ability to lower the amount of TRU or HALEU driver fuel needed. The design objectives, constraints, and optimization algorithm for a SFR are identified. Then, METAL is applied to design a SFR core fueled by TRU driver fuel assemblies and LEU blanket assemblies. To demonstrate the advantage of introducing LEU blankets, METAL is also used to design another SFR core following the same objectives and constraints, but using naturally enriched blankets. By comparing the two cores, it was found that the introduction of LEU blankets results in a ∼28% reduction in the driver fuel mass requirements. Upon development of a levelized fuel cycle cost (LFCC) model, the 28% reduction in driver fuel mass corresponds to a 10% decrease in the LFCC. Additional advantages of using LEU blankets include reducing the core conversion ratio (CR), and improving the assembly radial power peaking factor (RPPF), which enhances the safety and non-proliferation performances of the SFR core.

Glass formulation and lab-scale testing of glasses designed for in-can Melter and in-container Vitrification of high-assay low-enriched uranium aqueous polishing Raffinate waste

Glasses were designed for processing a nuclear waste from aqueous polishing of high-assay low-enriched uranium using either In-Can Melter or GeoMelt® In-Container Vitrification™ (ICV) technologies, which operate at temperatures of Tp ≤ 1100 °C and Tp ≤ 1450 °C, respectively. Due to the different operating conditions, the melt and glass properties were optimized differently for each technology. Each glass was designed to optimize for maximum waste loading while simultaneously satisfying processing (e.g., crystallization, viscosity, conductivity) and product quality (e.g., durability, hazard characteristic, crystal content) constraints. The raffinate waste contains high nitric acid (4 M) and low total solids (9 g∙L−1) concentrations. Feed preparation processes were tested to facilitate concentration and nitrate destruction/removal while controlling redox of the melter feed. A successful lab-scale feed process including sugar addition and spray-drying was performed to generate an adequate melter feed for ICV processing.

Neutronic analysis of high burnup thorium-HALEU fuels in PHWR

Recent developments have been made in a novel fuel for pressurised heavy water reactors: high burnup oxide fuels composed of a mixture of thorium and high assay low enriched uranium. Several claims are made regarding the performance of such fuels in terms of economics, waste and non-proliferation, among others. This article analyses this class of fuel using an infinite lattice study in SERPENT. First, a set of fuel compositions in the thorium/uranium enrichment design space that satisfy the high burnup target is produced. Then, a representative thorium-based fuel is compared to a uranium-only fuel with the same discharge burnup, as well as current standard fuel, in the context of the various performance claims.

The Bimodal Epithermal Astronuclear Reactor (BEAR): A LONG-LIVED, Universal Space Nuclear Powerplant

A neutronic model of a Nuclear Thermal Propulsion reactor based on the NERVA Peewee design using High Assay Low Enriched Uranium (HALEU) fuel in a Zirconium Carbide matrix has been thoroughly characterised and continuously modified over the last few years at the CSNR. The initial design (Emu, for the flightless bird) created in the summer of 2020 and presented at the NETS 2021 conference, has undergone significant adjustments in this time. Two major investigations have been undertaken and evolved over time; one into the requisites for the provision of electrical power to the propelled spacecraft via the circulation of a noble gas mixture in a closed Brayton cycle, and another into the use of planetary regolith as a reflector, reducing neutron leakage at what would otherwise be the reactor’s end of life in vacuo, rekindling it for an extended period of time as well as providing a very easy ultimate disposal solution, replacing the control drums with rods. It was determined that there are no insurmountable obstacles to performance in either mode.

Evaluating Hydrogen-based Moderators in High-Temperature Gas-cooled Reactors with 5 wt% Enriched Uranium Annular Fuel Rods

Abstract Small Modular Reactors (SMRs) based on high temperature gas-cooled reactor (HTGR) technology are being developed for providing high-temperature process heat and high-efficiency (>40%) electrical power generation. However, most of the HTGR-SMR concepts require high assay low enriched uranium (HALEU) fuel, with enrichments typically above 10 wt% 235U/U, to get sufficiently high burnup levels and fuel lifetime. The need for HALEU is primarily a consequence of the low volumetric density of fissionable material in tri-structural isotropic (TRISO) fuel particles, and also the use of graphite as a moderator/reflector. A previous study has shown that a modified prismatic HTGR fuel assembly with hydrogen-based moderator (7LiH) and cylindrical fuel elements of 5 wt% 235U/U enriched uranium can greatly reduce fuel consumption of a HTGR. However, such a design concept could lead to positive temperature reactivity coefficients (TRCs), making reactor control more challenging, with reduced passive safety. The purpose of this study is to evaluate variations of the hydrogen-based moderator in this alternative fuel assembly concept to identify configurations that achieve negative TRCs, thus improving passive safety characteristics. Calculation results demonstrate that negative TRCs can be achieved with reduced hydrogen mass such that natural uranium consumption is substantially less than that of the TRISO fuel concept, with comparable or higher core life. This study also explores the option of using 7LiOH and NaOH as hydrogen-based moderators, instead of 7LiH, thus allowing operation at higher temperatures, where hydrogen TRCs are lower.

Modeling, analysis, and optimization of complex nuclear processes and facilities via computational methods: The HALEU process case study

Improving and adapting industrial systems to timely meet changing programmatic and market demands is an important goal to achieve, including when operating and maintaining complex nuclear processes and facilities. However, changes to these complex systems are costly, particularly when they are already in place and bounded to stringent requirements and constraints such as when handling radioactive material and contaminated equipment. These conditions often exist when treating spent nuclear fuel remotely within shielded nuclear radiation chambers, commonly referred as hot cells, to condition nuclear material and/or fabricate products for utilization in other nuclear enterprises such as in the manufacture of advanced nuclear fuel. The illustrative case considered here is the production of high assay low enriched uranium (HALEU) products supporting the deployment of advanced nuclear reactors. For the HALEU program, resources invested were and are being systematically analyzed so that these investments are maximized in a facility that is nearly 60 years old. A methodology that has effectively enabled optimized and improvements in the Spent Fuel Treatment (SFT) program, and consequently the HALEU program, involves discrete event simulation as addressed in this article. The quantification of multiple productivity metrics, including material processing rates, cycle times, bottlenecks, number of material transfers as well as equipment, workstation, and material handling utilization, has resulted in a myriad of diverse discoveries and data-informed decisions regarding process layout and constituent, labor levels and schedules, selection of new process units, storage needs, and other critical process configurations. This article describes such a computational capability being applied for decision-making, illustrates its application to an actual process and program, provides illustrative results, and argues how computational methods for the modeling, analysis, and optimization of complex processes and facilities does lead to informed decisions derived from data and not only from intuition.

Assessment of Reactor Physics Characteristics of Prismatic Fuel Concepts With a Hydrogen-Based Moderator for Use in a Fluoride Salt-Cooled Small Modular Reactor

Abstract Fluoride-salt-cooled high temperature reactor (FHR) effectively combines the solid fuel and moderator design of high-temperature gas-cooled reactor (HTGR) technology with the fluoride salt coolant (LiF-BeF2, FLiBe) of molten salt reactor (MSR) technology, enabling low-pressure (∼1 atm, 101.325 kPa), and high-temperature (∼700 °C) operations. The design and operational features of the FHR make it a potentially attractive option for a small modular reactor (SMR), provided that it can be modified and made physically small and operate at a low-enough power level (<350 MWth/<150 MWel). Most FHR-SMR designs use high-assay low enriched uranium (HALEU) fuel in the form of tri-structural isotropic (TRISO) fuel particles, combined with the use of a graphite moderator. However, there are alternative design concepts for an FHR-SMR that may offer superior performance characteristics, while utilizing an alternative fissile fuel supply option. In this exploratory study, lattice physics calculations were performed with Serpent to evaluate an alternative FHR-SMR prismatic fuel block design concept using coated annular fuel pellets instead of TRISO-particle fuel compacts, along with the use of hydrogen-based solid moderator rods made of 7LiH. In initial studies, it was found that fuel blocks with 120 moderator rods made of 7LiH tended to have large positive temperature reactivity coefficients (TRCs), which is undesirable for safety reasons. However, reducing the number of moderator rods to 90 or 54, while increasing the number of fuel rods and coolant holes led to low or negative temperature coefficients. For a prismatic fuel block design with 54-7LiH moderator rods, the isothermal temperature coefficient of reactivity (Isothermal TRC), with simultaneous changes in the fuel (F), graphite (G), hydrogen (H), and coolant (C) temperatures, ranges between −0.159 mk/K and −0.054 mk/K, depending on the operating temperature and fuel burnup. Such alternative FHR-SMR fuels could achieve a single-batch core life of ∼10 years with low enriched uranium (LEU) fuel, and ∼45 years with HALEU, in a 350-MWth reactor core.

Preliminary Conceptual Design of Nuclear Thermal Rocket Reactor Cores Using Ceramic Fuels with Beryllium or Composite Neutron Moderators

Several preliminary conceptual designs of nuclear thermal rocket reactor cores are presented that use tin-bonded monolithic ceramic [mononitride (UN), monocarbide (UC), and uranium dioxide (UO2)] fuel plates or pins with molybdenum-tungsten alloy clad. Neutron moderation is provided by a block of Be metal or composite materials using metal hydrides such as ZrH1.6 or YH1.6 with different matrices (MgO or Be). Mainly high-assay low-enriched uranium is considered, but highly enriched uranium is also assessed for a few configurations. Nominal core thermal power is 300 MW corresponding to about 66 kN (15 klbf) of thrust, and with minimal modifications, 500 MW may be possible (25 klbf of thrust). Depending on the configurations, the amount of 235U needed for criticality is 30 to 90 kg, and reactor weight is 2.5 to 3.8 tonnes.

A Review of Candidates for a Validation Data Set for High-Assay Low-Enrichment Uranium Fuels

Many advanced reactor concept designs rely on high-assay low-enriched uranium (HALEU) fuel, enriched up to approximately 19.75% 235U by weight. Efforts are underway by the US government to increase HALEU production in the United States to meet anticipated needs. However, very few data exist for validation of computational models that include HALEU, beyond a few fresh fuel benchmark specifications in the International Reactor Physics Experiment Evaluation Project. Nevertheless, there are other data with potential value available for developing into quality benchmarks for use in data- and software-validation efforts. This paper reviews the available evaluated HALEU fuel benchmarks and some of the potentially relevant benchmarks for fresh highly enriched uranium. It then introduces experimental data for HALEU fuel irradiated at Idaho National Laboratory, from relatively recent irradiation programs at the Advanced Test Reactor. Such data should be evaluated and, if valuable, collected into detailed benchmark specifications to meet the needs of HALEU-based reactor designers.

Microstructure and microchemistry changes at U-10Mo fuel/AA6061 cladding interfaces with varying hot isostatic pressing conditions

Monolithic uranium - 10 wt.% molybdenum (U-10Mo) is a promising high-assay low-enriched uranium fuel system for nuclear reactors used in research, medical isotope production, and remote power applications. During manufacturing, an aluminum alloy (AA6061) cladding is bonded to the U-10Mo fuel plate via hot isostatic pressing, during which diffusion and fuel/cladding chemical interaction occurs at plate edges. The microstructure and microchemistry changes that occur at fuel/cladding interfaces are important to understand as a function of process parameters to develop a reliable fuel fabrication process and to meet the desired specifications. Here, microstructural and microchemical changes are studied using complementary electron microscopy and atom probe tomography for two manufactured fuel plates with varied hot isostatic pressing (HIP) parameters. Results highlight that modifying thermomechanical processing parameters significantly changes the interaction layer thickness between U-10Mo and AA6061 by an order of magnitude. In addition, the distribution and concentration of elements (i.e., Al, Si, Mg) from cladding to fuel are investigated.

Extraction of Neptunium, Plutonium, Americium, Zirconium, and Technetium by Di-(2-Ethylhexyl)-Iso-Butyramide (DEHiBA) at High Metal Loadings

ABSTRACT Increased focus on carbon neutral energy has generated a resurgence of interest in nuclear power, and in particular advanced reactors which are likely to utilize high assay low enriched uranium (HALEU). This in turn could increase the economic attractiveness of recovering still partially enriched uranium from used nuclear fuel. Concomitant to development of advanced reactors, advanced reprocessing schemes should be developed which address the disadvantages to well established reprocessing schemes. The present study focuses on using di-(2-ethylhexyl)-iso-butyramide (DEHiBA) under high metal loading conditions for the reprocessing of used nuclear fuel. The elements examined in the study include the dominant transuranic actinides (Np, Pu, Am) as well as the often-problematic Tc and Zr. By increasing the concentration of the extractant from the more commonly reported 1.0 M – 1.5 M, the extraction of hexavalent actinides is substantially increased, while maintaining effective rejection of tri, tetra, and pentavalent actinides, particularly in the presence of high loadings of uranium. The extraction of Zr by 1.5 M DEHiBA is noted to be negligible by comparison to tributyl phosphate (TBP), however the coextraction of Tc with U is observed to be nominally twice the quantity that is extracted by TBP indicating a need for effective Tc management.

An Assessment of Heterogeneous Effects on System Reactivity for Criticality Safety Analyses with LEU+ and HALEU Materials

To develop the criticality safety basis for any system, process, or package, the worst-case configuration of materials resulting in the maximum system reactivity must be determined. It is commonly accepted that in terms of the maximum system reactivity, at the lower enrichments used in current commercial practice (i.e., 5 wt% 235U), a heterogeneous configuration is bounding of a homogeneous mixture of fissile and moderating materials. However, a common assumption made is that with increasing enrichment, a homogeneous system can be bounding. With increased industry interest in utilizing higher enrichments for commercial applications with low-enriched uranium (LEU+) (≤10 wt% 235U), and high assay low-enriched uranium (HALEU) (≤20 wt% 235U) materials, it has become increasingly important to verify any assumptions and to have a better understanding of the expected system behavior at these higher enrichments. The SCALE code system was used to assess the effects of heterogeneity on system reactivity with varying enrichments and system configurations for a UO2 and water system, typical of a transportation package criticality analysis. The purpose of this assessment was to provide insight on the effect of material heterogeneity on system reactivity with increasing enrichment. The results of this study confirm that for systems with a higher hydrogen-to–fissile material (H/X) ratio, the homogeneous mixture of material may be bounding for HALEU materials. However, for systems with a lower hydrogen-to–fissile material ratio (H/X ≤ 200), a heterogeneous configuration of contents is expected to be bounding for most LEU materials. Overall, for any LEU system, including HALEU material, heterogeneous reactivity effects should always be considered.

Conceptual Design of a Micro Nuclear Reactor for Canadian Arctic Communities

Small Canadian arctic communities rely on diesel generators for their electricity needs. Providing such generators with fuel year round presents logistical challenges because of inclement weather and the long transportation distances involved. This work presents the conceptual design of a 10-MW(thermal) microreactor that can be used to provide 3.5 MW of electricity as well as district heating to arctic communities. The reactor has a lead-cooled and graphite-moderated core with 13 vertical fuel channels containing high-assay low-enriched uranium fuel enriched to 10%. The core is enclosed in a unpressurized reactor vessel and is passively cooled through natural convection. Stirling engines are used to drive the electrical generators. The hot cylinders of the Stirling engines are located in the unpressurized reactor vessel and are heated directly by the primary coolant. Preliminary neutronic and thermal-hydraulic analyses of the core indicate that the design is technically feasible and that the reactor can function for 2 years and 9 months without refueling.

Centrus keeps contract to produce high-assay low-enriched uranium

Centrus keeps contract to produce high-assay low-enriched uranium

Multiphase Alloy Nuclear Waste Forms Developed for Pyrochemical U-10Mo Scrap Recovery Waste Streams

The US High Performance Research Reactor (USHPRR) Conversion program is developing an Al-clad Zr-bonded U-10Mo nuclear fuel foil to enable the conversion of research reactors to high-assay low-enriched uranium (HALEU) alloy fuel. Fabrication scrap from the fuel manufacturing will be electrorefined to recover the HALEU, which will generate waste streams consisting of Zr, Mo, and residual U retained in the stainless steel anode basket used in the electrorefiner. Four alloys were formulated with different relative amounts of Zr, Mo, and 316L-SS to represent the expected range of waste compositions. Laboratory-scale ingots were made and metallurgically characterized to assess differences in the microstructures and distributions of the surrogate waste metals. Different amounts of the same dominant phases with similar compositions were formed in most materials: a γ-austenite based matrix, two ZrFe2 intermetallic phases that could host residual uranium, an FeCrMo sigma (σ) intermetallic phase, a chi (χ) phase, and a secondary austenite (γ2) phase. Electrochemical tests were conducted to compare the corrosion behaviors in acidic and alkaline brine solutions for a range of simulated environmental redox conditions. Corrosion of the Mo-rich γ and γ2 solid solutions and the σ and χ intermetallic phases occurred. The multiphase alloy waste form formulated with the lowest molybdenum and highest zirconium contents showed the highest corrosion resistance and has sufficient amounts of ZrFe2 to immobilize trace amounts of residual uranium in the waste stream. The alloy formulation strategy and testing approach will be presented with key results.Work conducted at Argonne National Laboratory is operated for the U.S. Department of Energy, Office of Science under contact DE-AC02-06CH11357.

Evaluations of the Effect of Heterogeneity in HALEU Systems Using Modified Critical Benchmarks

This work uses evaluated criticality safety benchmarks to examine the effect of heterogeneity in high-assay low-enriched uranium (HALEU) systems. A subset of benchmarks selected based on their composition parameters is slightly modified to artificially introduce heterogeneity, and the difference in the effective multiplication factor between the homogeneous benchmark and the modified model is determined. The difference in the multiplication factor is evaluated for correlations against both the moderator ratio and the enrichment of the benchmarks, and the correlations are examined using established statistical methods. In several cases, statistically significant correlations are observed and discussed, and the lack of expected correlations is also discussed.

Preliminary conceptual design of fast neutron spectrum nuclear thermal rocket cores using monolithic uranium nitride fuel

This paper presents a few nuclear thermal rocket (NTR) reactor core preliminary conceptual designs based on the use of monolithic uranium nitride (UN) fuel plates or pins clad with tungsten (W), alloyed or unalloyed as needed. High-assay low-enriched uranium (HALEU) as well as high-enriched uranium (HEU) are considered. Nominal core thermal powers are between 300 and 1100 MW corresponding to thrusts between 66 kN and 240 kN. The estimated thrust-to-reactor-weight ratios (T/WRx) achievable with the HEU plate and pin configurations are between 3.8 and 10.9. The T/WRx achievable with the HALEU plate and pin configurations are lower, 1.9–4.0, than those achievable with HEU. It is noteworthy that even a relatively modest increase in fuel enrichment above the HALEU limit enables a significant increase in T/WRx. It must be emphasized that these results do not yet account for uncertainties and that the appropriateness of the assumed fuel thermal design limits for normal operation (i.e., peak centerline temperature of 3100 K) will need to be confirmed. Finally, just like for the other UN-based fuels currently considered for NTR applications (i.e., CERMET and CERCER), the proposed UN fuel systems need significant testing in prototypic conditions to confirm the appropriate performance for safe NTR applications. However, the monolithic fuel systems may have, overall, a higher technology readiness level (TRL), and, consequently, shorter fuel development and validation time may result.