Examination of Moderator Intrusion into 30B UF6 Cylinder for Nuclear Criticality Safety
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Lang, Alex [ORNL] (ORCID:000900050202983X)↗
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Credible Criticality Safety Margin in the 30B Package with LEU+ UF 6 and Hypothetical Water Ingress
The commercial nuclear industry is pursuing advancements in fuel and reactor design that increase the uranium enrichment above 5 wt. % 235 U. These advancements will necessitate the ability to transport bulk quantities of UF 6 at increased enrichments. Currently, the 30B cylinder is the primary container used by the industry for UF 6 storage and transportation and has a long history of successful shipments. This container can support up to 2,277 kg of UF 6 at a maximum enrichment of 5 wt. % 235 U. Previous evaluations have assessed the potential impact of criticality safety for 30B transport at higher enrichments but assumed moderator intrusion would not require evaluation. Although current regulations allow for the exception of moderator intrusion for UF 6 packages through the design and quality control of the package content, this exception is limited to enrichments up to 5 wt. % 235 U. Thus, an investigation of moderator intrusion into a 30B cylinder should be performed. Moderator intrusion into a 30B cylinder is a unique condition for criticality safety evaluation in transportation because of the violent chemical reactions that occur between UF 6 and H 2 O. The resulting intrusion is strongly dependent on the breach size, breach location, breach interface solid/ullage of the UF 6 content, the UF 6 distribution (which is temperature dependent), and temperature/pressure conditions which are dynamic in accident conditions. Additionally, the complexity of the HF-UO 2 F 2 -H 2 O interface that occurs during the event can influence the potential solubility of uranium in the system, as well as influence the amount of UF 6 reaction with H 2 O. With the operating experience from Orano Federal Services LLC, the UF 6 chemical expertise from Oak Ridge National Laboratory (ORNL), and the criticality safety expertise from ORNL, this paper evaluates the neutronic conditions (i.e., k eff ) that apply the understood chemistry and experimental conditions that occur during moderator intrusion of a breached 30B cylinder under postulated accidents. This report examines the historical evaluations of UF 6 transport and expands these evaluations for the enrichments expected for nuclear industry advancement. These simulations primarily feature a homogeneous mixture of UF 6 and H 2 O as an infinite media system and as a sphere with water reflection given an impurity limit of 0.5 wt. % UF 6 . This report demonstrates that this H/U limit is valid up to 8 wt. % enrichment for both H 2 O and HF as the moderating mixture. The water-reflected homogeneous mixed spheres evaluated demonstrate the amount of safety margin applied by restricting the 30B cylinder impurity limit to 0.5 wt. %. Additionally, this report evaluates moderator ingress scenarios of a 30B cylinder. The 30B cylinder simulations vary the cylinder orientation, the mass of UF 6 in the system, the mass of H 2 O in the system, and how much H 2 O has reacted with UF 6 . The moderator ingress of a 30B cylinder was evaluated for the water-reflected homogeneous mixed spheres to demonstrate the amount of safety margin applied by restricting the 30B cylinder impurity limit to 0.5 wt. %. These simulations are standard practice and are independent of the UF 6 -H 2 O reaction. This work also explored more complex simulations that layer the H 2 O over the UF 6 , which incorporates the amount of reacted UF 6 as a separate layer and considers solubility limits of UO 2 F 2 in H 2 O as it is produced. This approach is intended to simulate the postulated event of a large hairline crack occurring on a submerged cylinder. Finally, this work considers the potential of H 2 O mixing into the UF 6 solid heterogeneously using a sponge-like model. The water ingress into the UF 6 is treated as random size spheres, and the reaction products form into a layer over the spheres.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Thermal cross section measurements at the RPI LINAC [Abstract]
Recently, a cold moderator was designed and developed for use at the RPI LINAC. This cold moderator proved to easily and safely couple to an existing neutron producing target, while enhancing neutron flux below 0.02 eV by cooling polyethylene down to 29 K. This cold moderator capability allowed for significantly improved counting statistics below 0.02 eV not previously possible due to a poor signal to background ratio. Additionally, testing was performed to characterize the energy resolution of the new cold moderator system and found the system easily capable of resolving resonances in Ta-181 at 4 and 10 eV, while also clearly resolving the Bragg edges found in Be metal below 0.01 eV. Following the design and development of a cold polyethylene moderator, a series of thermal total cross section measurements were performed for polyethylene, polystyrene, Plexiglas and yttrium hydride in the thermal region. These measurements serve to help validate thermal scattering law (TSL) evaluations in the 0.0006 – 20 eV energy range. For polyethylene and polystyrene, two sets of experiments were performed – one with the Enhanced Thermal Target (ETT) and another with the ETT plus the new cold moderator capability (ETTC). The yttrium hydride and Plexiglas measurements were only performed with the ETTC. The measurements for polyethylene help to validate the data processing methodology when using the ETTC, while extending the measured range of polyethylene down to 0.0007 eV. Two different Plexiglas, Plexiglas G and Plexiglas G-UVT, and two different concentrations of yttrium hydride, H/Y = 1.85 and 1.68, were measured. Overall, all materials had generally good agreement with their ENDF/B-VIII.0 TSL evaluations, though some discrepancies were noticed. In the case of the yttrium hydride, the high energy oscillations in the hydrogen cross section and the low energy Bragg edges in the yttrium cross section were clearly seen. These measurements represent the first total cross section measurements that encompass the entire thermal region from 0.0006 – 3 eV for polystyrene and yttrium hydride.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Thermal Cross Section Measurements At The RPI LINAC
Recently, a cold moderator was designed and developed for use at the Rensselaer Polytechnic Institute Linear Accelerator (RPI LINAC). This cold moderator proved to easily and safely couple to an existing neutron producing target, while enhancing neutron flux below 0.02 eV by up to a factor of 8 by cooling polyethylene down to 29 K. This cold moderator capability allowed for significantly improved counting statistics below 0.02 eV not previously possible due to a poor signal to background ratio. Additionally, testing was performed to characterize the energy resolution of the new cold moderator system and found the system easily capable of resolving resonances in Ta-181 at 4 and 10 eV, while also clearly resolving the Bragg edges found in Be metal below 0.01 eV. Furthermore, following the design and development of a cold polyethylene moderator, a series of thermal total cross section measurements were performed for polyethylene, polystyrene, Plexiglas and yttrium hydride in the thermal region. These measurements serve to help validate thermal scattering law (TSL) evaluations in the 0.0005 – 1 eV energy range. For polyethylene and polystyrene, two sets of experiments were performed – one with the Enhanced Thermal Target (ETT) and another with the ETT plus the new cold moderator capability (ETTC). The yttrium hydride and Plexiglas measurements were only performed with the ETTC. The measurements for polyethylene help to validate the data processing methodology when using the ETTC, while extending the measured range of polyethylene down to 0.0005 eV. Two different Plexiglas, Plexiglas G and Plexiglas G-UVT, and two different concentrations of yttrium hydride, H/Y = 1.85 and 1.68, were measured. Overall, all materials had generally good agreement with their ENDF/B-VIII.0 TSL evaluations, though some discrepancies were noticed. In the case of the yttrium hydride, the high energy oscillations in the hydrogen cross section and the low energy Bragg edges in the yttrium cross section were clearly seen. These measurements represent the first total cross section measurements that encompass the entire thermal region from 0.0005 – 1 eV for polystyrene and yttrium hydride.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Comparison of the Neutronic Performance of IRP-1 and IRP-2 at SNS
The Spallation Neutron Source (SNS) operated from start-up, in May 2006, through December, 2017 before replacing the Inner Reflector Plug (IRP). This component is the most costly and time-consuming “regular maintenance” replacement foreseen in the operation of the First Target Station (FTS). The IRP includes beryllium reflector blocks, steel and aluminum support structures, aluminum moderator vessels (which in turn include gadolinium and cadmium neutron poisons) and water coolant systems. The lifetime of this multi-million dollar component is driven by poison and decoupler burnup in moderator vessels. The moderators include these neutron poisons to best match neutron beam performance with scattering instrument needs. The burnup that comes with use changes the performance of the moderators until they no longer server their intended purpose. The reasonable desire to maximize the IRP lifetime means that the moderator performance at the Beginning Of Life (BOL) is significantly “over poisoned” and different from the performance at moderator End Of Life (EOL). Upon installation of the second IRP article, IRP-2, the SNS Neutronics team characterized the performance of the neutron beamlines in order to assess the differences between them, primarily by comparing the EOL performance of IRP-1 to the BOL performance of IRP-2. While IRP-2 was largely similar to IRP-1 in its conceptual design, there were some differences in the specific design of IRP-2 to support enhanced manufacturability, and the change from IPR-1 to IRP-2 coincided with the replacement of the light water in the IRP cooling loop with heavy water, as had been intended in the SNS design but not implemented during construction because of heavy water availability.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Final Design for Thermal/Epithermal eXperiments (TEX) with Chloride Absorbers to Provide Validation Benchmarks for Y-12 Electrorefining Facility
Uranium electrorefining operations at Y-12 require validation for chlorine absorption and reflection. Plutonium chloride solution operations at Los Alamos National Laboratory require chlorine absorption validation. The growing need for chlorine validation is evident to the wider criticality safety community, with multiple attendees at the recent TEX 2.0 meeting at Lawrence Livermore National Laboratory (May 2023) requesting validation for chlorine (Idaho National Laboratory (INL), Institute de radioprotection et de surete nucleaire (IRSN), Savannah River Nuclear Site (SRNS), LANL, and Y-12). Los Alamos National Laboratory has recently performed, and benchmarked, an experiment titled Chlorine Worth Study (CWS) for the internal operations at the lab, but due to the difficulties in precisely characterizing the material compositions of the chlorine absorbers it is advantageous to perform a complimentary study with a different chlorine-based absorber material. Furthermore, having a uranium-based vs plutonium-based experiment provides a separate and important validation basis for criticality safety and nuclear data evaluation. The original final design report for the study of chlorine absorption using the TEX-HEU experimental base was presented in 2022, but used the same chlorine-bearing materials that were found to be difficult to characterize in the LANL benchmark. A complete redesign of the experiment has been performed looking at alternative absorber materials in various forms to produce an experiment that is fully characterizable. This report presents five novel chlorine experiments using the TEX-HEU test bed which provides direct comparison to the Y-12 and INL/Terrapower application needs, utilizing sodium chloride (NaCl) absorber plates. The absorber plates will consist of granulated NaCl (≥99.5% pure), which will be fully encapsulated in aluminum tins, providing a simple but effective chlorine-based absorber material that can be completely characterized. Of the five configurations presented in this report, it is expected that two or three configurations will be down selected for the actual experiment, with the other configurations being alternates. Three of the configurations are in the standard configuration, where the absorber is placed directly on the HEU fuel plates, and two in the sandwich configuration, where the absorber is surrounded by polyethylene moderators to force additional neutron thermalization prior to reaching the absorber. There are two thicknesses of NaCl absorber plates: 3/16” and 1/4” active thicknesses (i.e. not including the encapsulation). Both variations of the absorbers have an active absorber radius of 6” and a total radius of 7.5” to match the diameter of the HEU plates, with the outer 1.5” being aluminum encapsulation. The high-density polyethylene (HDPE) moderators are of the thicknesses: 27/16”, 7/4”, 1/8”, 11/16”, and 3/4”. The final configurations have six (one sandwich and one standard configuration), eight (one sandwich and one standard configuration), and 18 total fuel layers (standard configuration). The standard and sandwich configurations were designed such that the differences in moderator thicknesses are 1” HDPE, which were already procured for the original CED-2. The proposed configurations were precisely tuned to closely match the sensitivity profiles and neutron spectra of the Y-12 upset cases and were also compared to the INL/Terrapower upset cases. The assessment of experimental uncertainties of the non-absorber components was predicted to be 0.00114 Δk eff . The assessment of uncertainties resulting from the absorbers was predicted to be 0.00029 Δk eff . This results in a total uncertainty of 0.00118 Δk eff . Many of the largest uncertainties, namely the moderator densities, may be reduced with precision dimensional inspection of the components. The 1” HDPE moderators as well as the HDPE reflectors from the original CED-2 were incorporated in the final designs presented here. Additional HDPE moderator plates must be fabricated to complete the configurations. NaCl absorber plates will by fabricated at LLNL. The total additional cost is expected to be $\$$54,250 for the remaining components. Precise inspection, including dimensional, mass, density, and impurity, is recommended for all components. LLNL estimates that these costs are around $\$$12,000. It is expected, based on previous TEX-HEU experiments, that three weeks of experimental facility time is needed to complete the experiments.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Determination of energy correction coefficient for neutron albedo dosimeter. Three approaches
To accurately determine dose, albedo TLD dosimeters require application of a neutron energy correction factor (NECF). Here, in this paper, the results from three different methods were used to determine NECFs. In the first method, NECF values were calculated as a ratio of the known “conventional true” dose and the neutron dose reported by albedo dosimeter(s) irradiated on a phantom. The results from this direct method were utilized for verification of two other methods. The second method used a specially designed device called the Neutron Area Monitor (NAM) model 5. The third method of NECF determination was based on the response of a BF 3 tube under different levels of moderation. The results of this work showed that the Navy-developed Neutron Area Monitor (NAM) model 5 measures NECF values correctly when irradiated with 252 Cf (bare and D 2 O-moderated). This was an important result, because it validated the use of the NAM-5 in nuclear power and marine propulsion reactors, as their neutron energy spectrum is approximated by the field of a Cadmium-coated 30 cm diameter steel sphere of D 2 O moderated 252 Cf. In the case of a broad neutron energy spectrum the 6 Li response to the small amount of slow neutrons outweighs its response to the fast neutrons overwhelmingly responsible for the neutron dose rate. Based on the obtained results we can conclude that NAM-5 provides correct NECF determination if the fraction of slow neutrons in the neutron dose rate is more than 20%. A satisfactory correlation between the NECF determined by NAM-5 and the multi-moderated BF3 tube method was observed, but the multi-moderated BF 3 tube values were ~30% lower than those obtained by the NAM-5 at the same locations for broader neutron energy spectrum locations.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Pre-screening to guide coronary artery calcium scoring for early identification of high-risk individuals in the general population
Abstract Aims To evaluate the ability of Systematic COronary Risk Estimation 2 (SCORE2) and other pre-screening methods to identify individuals with high coronary artery calcium score (CACS) in the general population. Methods and results Computed tomography-based CACS quantification was performed in 6530 individuals aged 45 years or older from the general population. Various pre-screening methods to guide referral for CACS were evaluated. Miss rates for high CACS (CACS ≥300 and ≥100) were evaluated for various pre-screening methods: moderate (≥5%) and high (≥10%) SCORE2 risk, any traditional coronary artery disease (CAD) risk factor, any Risk Or Benefit IN Screening for CArdiovascular Disease (ROBINSCA) risk factor, and moderately (>3 mg/24 h) increased urine albumin excretion (UAE). Out of 6530 participants, 643 (9.8%) had CACS ≥300 and 1236 (18.9%) had CACS ≥100. For CACS ≥300 and CACS ≥100, miss rate was 32 and 41% for pre-screening by moderate (≥5%) SCORE2 risk and 81 and 87% for high (≥10%) SCORE2 risk, respectively. For CACS ≥300 and CACS ≥100, miss rate was 8 and 11% for pre-screening by at least one CAD risk factor, 24 and 25% for at least one ROBINSCA risk factor, and 67 and 67% for moderately increased UAE, respectively. Conclusion Many individuals with high CACS in the general population are left unidentified when only performing CACS in case of at least moderate (≥5%) SCORE2, which closely resembles current clinical practice. Less stringent pre-screening by presence of at least one CAD risk factor to guide CACS identifies more individuals with high CACS and could improve CAD prevention.
Ties, Daan (ORCID:0000000164729886)↗
Yttrium Hydride Post Irradiation Examination Plan
The overarching goal of the Department of Energy Office of Nuclear Energy (DOE-NE) microreactor program is to develop technologies for the deployment of civilian microreactors by stakeholders1. Microreactors are expressed as advanced transportable nuclear reactors operating at low power (<20MWth) but high temperatures (>600°C), as well as plug-and-play and inherently safe designs. One prerequisite of a microreactor is the compactness, so that a truck can transport the reactor under safe conditions with the current road infrastructure1,2. The compactness of these reactors likely can be attainable by use of solid components for the essentials of the nuclear core, such as fuel, core heat removal components, reflectors, and moderators. Among these essentials, where fuel enrichment must remain < 20% to meet High Assay Low Enrichment Uranium criteria (HALEU), the largest contribution to the compactness is offered by use of solid moderators which benefit from light atomic weight elements, such as hydrogen, carbon, and beryllium2. Among these, hydrogen-bearing materials, such as metal hydrides, are superior to other options from the lowest critical mass standpoint. Noting that, factors other than critical mass should be considered for a specific reactor design. Yttrium- or zirconium-based metal hydrides have been down-selected due to their neutronic performance. In addition to the neutronic perspective, maintaining hydrogen within the metal hydride is important at the high operating temperatures proposed by advanced reactors. Yttrium hydride (YHx) is, therefore, a proposed moderator material that offers better hydrogen retention at higher operating temperatures than zirconium hydrides due to higher retention and thermal stability of hydrogen in the metal3. The irradiated materials properties of metal hydrides, in this case YHx, must be assessed for the qualification of these moderators. Material testing and inspection processes must illustrate that the effects of dimensional and property changes on thermophysical and mechanical properties do not cause any significant changes on the microreactor safety, and the moderating power is maintained within design limits. Thus, the effect of irradiation on the thermophysical and mechanical properties must be determined. This post-irradiation examination (PIE) plan specifically aims to determine these properties for YHx following Advanced Test Reactor (ATR) irradiation.
99 GENERAL AND MISCELLANEOUS↗
Comparison of Nuclear Thermal Propulsion Reactor Cores Using Different Fuel Forms, Neutron Spectra, and Uranium Enrichments
A comparison of 300 MWt HEU and HALEU NTP cores using different fuel forms is presented to demonstrate trends and potentially inform fuel development. Compared to HEU, HALEU cores require additional in-core moderation. Beryllium was chosen as the reference in-core moderator material; metal hydrides may also be considered. Typical NTP fuel forms such as carbide and carbonitride solid solutions, carbide-graphite composites, CERMETs and CERCERs were considered together with less typical fuel forms such as UC and U metal. Be-moderated HALEU cores are heavier than HEU cores (1.9-2.7 and 0.9-2.0 metric tons, respectively). Furthermore, HALEU loading is very sensitive to the amount of in-core moderation which translates into tight fabrication tolerances. HALEU cores requires more reactivity control mechanisms than HEU cores to compensate for temperature feed-back. Using enrichments only slightly above HALEU (e.g., 25%) may provide a compromise between the administrative preference for using HALEU and the engineering preference for simplifying the reactor systems by, for example, not using in-core moderators.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
PLN-6268 Yttrium Hydride Post Irradiation Examination Plan
The overarching goal of the Department of Energy Office of Nuclear Energy (DOE-NE) microreactor program is to develop technologies for the deployment of civilian microreactors by stakeholders1. Microreactors are expressed as advanced transportable nuclear reactors operating at low power (<20MWth) but high temperatures (>600°C), as well as plug-and-play and inherently safe designs. One prerequisite of a microreactor is the compactness, so that a truck can transport the reactor under safe conditions with the current road infrastructure1,2. The compactness of these reactors likely can be attainable by use of solid components for the essentials of the nuclear core, such as fuel, core heat removal components, reflectors, and moderators. Among these essentials, where fuel enrichment must remain < 20% to meet High Assay Low Enrichment Uranium criteria (HALEU), the largest contribution to the compactness is offered by use of solid moderators which benefit from light atomic weight elements, such as hydrogen, carbon, and beryllium2. Among these, hydrogen-bearing materials, such as metal hydrides, are superior to other options from the lowest critical mass standpoint. Noting that, factors other than critical mass should be considered for a specific reactor design. Yttrium- or zirconium-based metal hydrides have been down-selected due to their neutronic performance. In addition to the neutronic perspective, maintaining hydrogen within the metal hydride is important at the high operating temperatures proposed by advanced reactors. Yttrium hydride (YHx) is, therefore, a proposed moderator material that offers better hydrogen retention at higher operating temperatures than zirconium hydrides due to higher retention and thermal stability of hydrogen in the metal3. The irradiated materials properties of metal hydrides, in this case YHx, must be assessed for the qualification of these moderators. Material testing and inspection processes must illustrate that the effects of dimensional and property changes on thermophysical and mechanical properties do not cause any significant changes on the microreactor safety, and the moderating power is maintained within design limits. Thus, the effect of irradiation on the thermophysical and mechanical properties must be determined. This post-irradiation examination (PIE) plan specifically aims to determine these properties for YHx following Advanced Test Reactor (ATR) irradiation.
99 GENERAL AND MISCELLANEOUS↗
Electrode scale and electrolyte transport effects on extreme fast charging of lithium-ion cells
A combination of cell testing and electrochemical modeling is used to investigate extreme fast charging (XFC) performance for cells with a low loading of 1.5 mAh.cm -2 and moderate loading of 2.5 mAh.cm -2 . Cells with a low loading of 1.5 mAh.cm -2 withstand XFC performance remarkably well even up to 9C CC charging with high charge acceptance and very little lithium plating. For a moderate loading of 2.5 mAh.cm -2 , the 6C CC charge capacity is poor with significant amounts of lithium plating. Insufficient electrolyte transport properties are shown to be a major limitation for XFC performance of moderate and high loading cells. Charging at elevated temperature is shown to be an effective strategy for moderate loading cells enabling good 10-minute charge acceptance and mitigating lithium plating. Finally, the electrochemical model is used to investigate strategies for enabling 4-6C CC charging for cells with a loading of 3-4 mAh.cm -2 . A combination of elevated temperature, reduced tortuosity, and electrolytes with improved transport are likely needed to enable XFC for high energy density cells.
25 ENERGY STORAGE↗
Electrochemical CO2 Reduction over Metal-/Nitrogen-Doped Graphene Single-Atom Catalysts Modeled Using the Grand-Canonical Density Functional Theory
Renewably driven, electrochemical conversion of carbon dioxide into value-added products is expected to be a critical tool in global decarbonization. However, theoretical studies based on the computational hydrogen electrode largely ignore the nonlinear effects of the applied potential on the calculated results, leading to inaccurate predictions of catalytic behavior or mechanistic pathways. Here, we use grand canonical density functional theory (GC-DFT) to model electrochemical CO2 reduction (CO2R) over metal- and nitrogen-doped graphene catalysts (MNCs) and explicitly include the effects of the applied potential. We used GC-DFT to compute the CO2 to CO reaction intermediate energies at -0.3, -0.7, and -1.2 VSHE catalyzed by MNCs each doped with 1 of the 10 3d block metals coordinated by four pyridinic nitrogen atoms. Our results predict that Sc-, Ti-, Co-, Cu-, and Zn-N4Cs effectively catalyze CO2R at moderate to large reducing potentials (-0.7 to -1.2 VSHE). ZnN4C is a particularly promising electrocatalyst for CO2R to CO both at low and moderate applied potentials based on our thermodynamic analysis. Our findings also explain the observed pH independence of CO production over FeN4C and predict that the rate-determining step of CO2R over FeN4C is not *CO2- formation but rather *CO desorption. Additionally, the GC-DFT-computed density of states analysis illustrates how the electronic states of MNCs and adsorbates change non-uniformly with applied potential, resulting in a significantly increased *CO2- stability relative to other intermediates and demonstrating that the formation of the adsorbed *CO2- anion is critical to CO2R activation. This work demonstrates how GC-DFT paves the way for physically realistic and accurate theoretical simulations of reacting electrochemical systems.
CO2 reduction↗
Ceramic Composite Inert Matrix Fuel Forms in High-Temperature Gas-Cooled Microreactors
Here, this work optimizes micro-prismatic high-temperature gas reactor (HTGR) designs to reduce the energy-normalized mass of spent nuclear fuel (SNF) and high-level waste (HLW) produced. The optimization was performed for the current graphite moderator and an inert matrix fuel (IMF) concept employing different composite moderators in a prismatic design architecture. The fuel matrix is magnesium oxide (MgO) with entrained tristructural-isotropic (TRISO) fuel. The moderator materials, including beryllium oxide (MgO-BeO) and beryllium (MgO-Be) at 40 vol % loading and yttrium hydride (MgO-YH x=1.9 ) and zirconium hydride (MgO-ZrH x=1.9 ) at 15 vol % loading, were entrained within the MgO host matrix. A generic graphite micro-prismatic HTGR is used as the baseline point design where the external dimensions are held constant. The composite moderator designs use 19.9% enriched uranium nitride TRISO fuel and hexagonal assemblies. For each IMF concept, an optimization study was performed to maximize the discharge burnup of the fuel by varying the TRISO packing fraction and the lattice pitch of the assemblies. The mass of SNF and HLW, other waste metrics, fuel cost, environmental impact metrics, and the activity of the SNF and HLW at 100 years and 100 000 years were calculated for the optimized IMF and graphite reference designs. The IMF results were subsequently compared to those of the graphite reference and the values for a light water reactor (LWR) and a small modular LWR. For the SNF and HLW, all the IMF concepts and the graphite reference produced less waste compared to the traditional LWR designs. However, the IMF concepts outperformed the graphite reference regarding the mass of SNF and HLW. For the other waste metrics, the IMF concepts showed reductions in fuel cost with improved environmental metrics relative to the graphite reference. Overall, the IMF concepts significantly reduced the SNF and HLW produced per unit of energy generated compared to traditional LWR designs.
TRISO↗
Melting rate correlation with batch properties and melter operating conditions during conversion of nuclear waste melter feeds to glasses
The rate of conversion of nuclear waste melter feed to glass is affected by the selection of melter feed materials and by melter design and operation. The melting rate correlation (MRC) is an equation that relates the glass production rate with two types of variables: (1) feed and melt properties: conversion heat, cold-cap bottom temperature, and glass melt viscosity; and (2) melter design and operation parameters: melter geometry, melter operating temperature, and gas bubbling rate. The MRC shows good agreement for an extended melting-rate data set of high-level waste (HLW) melter feeds and a data set generated for low-activity waste (LAW) melter feeds. Laboratory observation of heated melter feed samples is often used to assess the cold-cap bottom temperature of HLW melter feeds (moderately foaming feeds), but this technique appears inadequate for LAW melter feeds (vigorously foaming feeds). For LAW feeds, an adequate assessment of the cold-cap bottom temperature was achieved using evolved gas analysis, which allows identification of the collapse of primary foam for oxidized feeds. This assessment shows that the cold-cap bottom temperature for vigorously foaming LAW feeds is higher than that for moderately foaming HLW feeds. When the results of MRC are compared, LAW feeds are generally less sensitive to the bubbling rate and melt viscosity, and more sensitive to the cold-cap bottom temperature than HLW feeds. The MRC qualifies as a promising tool to support the selection of melter feed materials and melter operating conditions, which is determined from expensive independent scaled melter experiments, and sophisticated mathematical models.
Lee, Seung Min↗
Transcriptional, metabolic, physiological and developmental responses of switchgrass to phosphorus limitation
Abstract Knowing how switchgrass ( Panicum virgatum L . ) responds and adapts to phosphorus (P)‐limitation will aid efforts to optimize P acquisition and use in this species for sustainable biomass production. This integrative study investigated the impacts of mild, moderate, and severe P‐stress on genome transcription and whole‐plant metabolism, physiology and development in switchgrass. P‐limitation reduced overall plant growth, increased root/shoot ratio, increased root branching at moderate P‐stress, and decreased root diameter with increased density and length of root hairs at severe P‐stress. RNA‐seq analysis revealed thousands of genes that were differentially expressed under moderate and severe P‐stress in roots and/or shoots compared to P‐replete plants, with many stress‐induced genes involved in transcriptional and other forms of regulation, primary and secondary metabolism, transport, and other processes involved in P‐acquisition and homeostasis. Amongst the latter were multiple miRNA399 genes and putative targets of these. Metabolite profiling showed that levels of most sugars and sugar alcohols decreased with increasing P stress, while organic and amino acids increased under mild and moderate P‐stress in shoots and roots, although this trend reversed under severe P‐stress, especially in shoots.
59 BASIC BIOLOGICAL SCIENCES↗
Plutonium Production Reactor Progression Problems: Magnox Neutronics Benchmarks
The goal of this document is to provide a set of comprehensive benchmarking neutronics problems for gas-cooled, graphite-moderated reactors. The problems are designed to be representative problems that increase in complexity to test performance of various codes. For the simulation of the benchmarking problems, the Shift Monte Carlo code and MPACT deterministic code are chosen because of their state-of-the-art capabilities, which are optimized for various reactor designs. Because these codes have been developed primarily for light water reactor applications, their application to graphite-moderated, gas-cooled reactors has not been extensively explored. This report attempts to evaluate benchmarks for a Magnox-style reactor, one that is graphite-moderated and gas-cooled. Parameters that are calculated and presented in this benchmark include the eigenvalue, peaking factors, and isotope concentrations. Results show good agreement between Shift and MPACT, and demonstrate the capability of MPACT to model graphite-moderated systems well.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
WEC Irradiation Can Thermal Testing Report
With renewed interest for the next generation of small modular reactors and space-based nuclear reactors, metal hydrides have been a focus of research for moderating materials to reduce the need for highly enriched fuel to maintain small core sizes. Yttrium hydride (YH) is an attractive candidate for these metal hydrides due to its ability to retain hydrogen at elevated temperatures as well as the low neutron absorption cross section of yttrium metal. Given the necessity for reactors to operate at elevated temperatures for extended periods of time, understanding hydrogen retention in moderator channels in simulated operating thermal conditions is paramount in predicting reactor performance and the long-term stability of YH. Additionally, coating strategies on the moderator encapsulating will be explored to ascertain if it aids in retention of hydrogen within the capsule and moderator.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗