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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Radiation Signature Training Device Gamma Equivalent Spheres

Radiation signature training device sources are a continuation of a previous effort to produce a highly enriched uranium equivalent sphere. The original research was designed and motivated to produce a Department of Transportation special form source that emulated the passive gamma signature of a solid metal highly enriched uranium sphere. In addition, the mass reduction (i.e., of the 235 U) minimized logistical costs (security, criticality safety, radiological safety and transportation requirements). Radiation signature training device sources expand on the original concept, redesigning the encapsulation in a modular fashion that allows a variety of loadings. The result is a reconfigurable special nuclear material source set that can vary greatly in the magnitude of the gamma flux and the isotopic distribution. Use of the source to test a variety of standards (N42.34-2015, N42.38-2015, N42.43-2015, N42.53-2018, N42.48-2018, DHS-TCS Backpack Based Radiation Detection Systems, DHS-TCS Handheld Instruments Used for the Detection and Identification of Radionuclides, DHS-TCS Vehicle Mounted Mobile Systems) will be discussed.

Blessinger, Christopher S.↗
DOE OSTI2022

Propagation of viral genomes by replicating ammonia-oxidising archaea during soil nitrification

Ammonia-oxidising archaea (AOA) are a ubiquitous component of microbial communities and dominate the first stage of nitrification in some soils. While we are beginning to understand soil virus dynamics, we have no knowledge of the composition or activity of those infecting nitrifiers or their potential to influence processes. This study aimed to characterise viruses having infected autotrophic AOA in two nitrifying soils of contrasting pH by following transfer of assimilated CO 2 -derived 13 C from host to virus via DNA stable-isotope probing and metagenomic analysis. Incorporation of 13 C into low GC mol% AOA and virus genomes increased DNA buoyant density in CsCl gradients but resulted in co-migration with dominant non-enriched high GC mol% genomes, reducing sequencing depth and contig assembly. We therefore developed a hybrid approach where AOA and virus genomes were assembled from low buoyant density DNA with subsequent mapping of 13 C isotopically enriched high buoyant density DNA reads to identify activity of AOA. Metagenome-assembled genomes were different between the two soils and represented a broad diversity of active populations. Sixty-four AOA-infecting viral operational taxonomic units (vOTUs) were identified with no clear relatedness to previously characterised prokaryote viruses. These vOTUs were also distinct between soils, with 42% enriched in 13 C derived from hosts. The majority were predicted as capable of lysogeny and auxiliary metabolic genes included an AOA-specific multicopper oxidase suggesting infection may augment copper uptake essential for central metabolic functioning. These findings indicate virus infection of AOA may be a frequent process during nitrification with potential to influence host physiology and activity.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2024

Radiological Characterization of Uranium Decay Products in LEU U-10Mo Before and After Heat Treatment

Fuel for the U.S. high-performance research reactor fleet is undergoing significant development as the United States moves away from using highly enriched uranium dispersion fuels. The proposed fuel is a high-density, low-enriched uranium (LEU, 19.75 wt% U 235) alloyed with ten weight percent molybdenum fuel (known as LEU U 10Mo). The LEU U 10Mo fuel samples studied in this report were made by down-blending highly enriched uranium with a master alloy, which is cast from molybdenum rods and depleted uranium, via vacuum induction melting and then casting.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Uranium integral fission product yields for a spectrally-shaped 14.1 MeV neutron source at the National Ignition Facility

This paper describes the experimental results for an energy tuning assembly created to modify the National Ignition Facility deuterium–tritium fusion neutron source into a notional thermonuclear and prompt fission neutron spectrum, which has applications in integral measurements, nuclear data benchmarks, and radiation effects on microelectronics. The Monte Carlo neutron transport utilized MCNP5 to estimate the ETA-modified fluence using the ENDF-B/VIII.0 and IRDFF-II continuous energy nuclear data libraries, and SCALE Sampler was used to estimate the systematic nuclear data covariance using ENDF-B/VII.1 and IRDFF-II in a 252-group structure. The experiment fielded eight activation foils and a highly enriched uranium sample. This provided fifteen reaction channels that were used in a forward-fit comparison to the modeled results and to unfold the neutron spectrum using STAYSL. Gamma-ray spectrometry was performed on the activation and highly enriched uranium foils, and the reduced Χ 2 between the modeled and experimental values was 1.21. The results from the STAYSL unfold, reduced Χ 2 = 1.62, indicated that the modeled neutron spectrum was achieved and the systematic nuclear data uncertainty associated with the neutron transport and activation product cross sections was representative of the experiment. Integral cumulative fission product yield data were collected for 37 mass chains with a combination of gamma-ray spectrometry and radiochemical analysis. Fission product analysis was generally in agreement with two models using a semi-empirical fit and the General Observables of Fission code, with the exception of mass chains 88, 109, 111, 112, 113, 129, 139, 142, 144, 151, and 156.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Detecting and characterizing special nuclear material for nuclear nonproliferation applications

Abstract There is an urgent need for new, better instrumentation and techniques for detecting and characterizing special nuclear material (SNM), i.e., highly enriched uranium and plutonium. The development of improved instruments and techniques requires experiments performed with the SNM itself, which is of limited availability. This paper describes the findings of experiments performed at the National Criticality Experiments Research Center conducted using new instruments and techniques on unclassified, kg-quantity SNM objects. These experiments, performed in the framework of the Department of Energy, National Nuclear Security Administration Consortium for Monitoring, Technology, and Verification, focused on detecting, characterizing, and localizing SNM samples with masses ranging from 3.3 to 13.8 kg, including plutonium and highly enriched uranium using prototype detectors and techniques. The work demonstrates SNM detection and characterization using recently-developed prototype detection systems. Specifically, we present new results in passive detection and imaging of plutonium and uranium objects using gamma-ray and dual particle (fast neutron and gamma-ray) imaging. We also present a new analysis of the delayed neutron emissions during active interrogation of uranium using a neutron generator.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Godiva IV central cavity neutron environment characterization with threshold neutron detectors

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

Whitman, Nicholas H.↗
DOE OSTI2026

Microreactor Assembly Transportation Cask Model Description for Criticality Safety Validation Basis Assessment (Rev. 3)

Criticality safety analyses are completed on a transportation cask used for microreactor assembly shipment to provide an example of model and analysis to industry for reproducing this type of study on their microreactor fuel shipment. The fuel assembly considered is based on a gas-cooled microreactor (GC-MR), which utilizes HALEU fuel in the form of TRISO particles and utilizes various design options considered in industry designs. Various versions of this GC-MR assembly were studied, with and without YH 2 moderator, providing similar conclusions. The shipment cask design is revised based on an existing ES-3100 design, developed by Y-12 for the transport of highly enriched uranium (HEU), but is enlarged to hold the GC-MR fuel assembly. Criticality safety analysis for the cask/GC-MR fuel assembly package was performed using the CSAS6 sequence of SCALE6.3.2, utilizing the ENDF/B-VII.1 based continuous energy neutron library, and the analysis strictly follows the guideline from NRC reference reports. Different scenarios, e.g. normal operation, undamaged cask with water flooded, damaged cask with optimal water moderation, have been analyzed and it could be concluded the package would always have a large margin of subcriticality even packed in an infinite array. Sensitivity and similarity analyses are also performed using the TSUNAMI sequence of SCALE6.3.2, and the similarity analysis uses all the experiments from the ICSBEP Handbook with Highly Enriched Uranium (HEU), Intermediate and Mixed Enriched Uranium (IEU) and Low Enriched Uranium (LEU) systems, together with additional ones that are sponsored by the DNCSH program, and selected IRPhEP experiments using TRISO fuel and graphite moderator. These similarity analyses indicate that the dry nominal design has no similar benchmark experiments (ck values greater than 0.8), which may become problematic if more assemblies are shipped together (or a fully loaded core is shipped) and margin to criticality is reduced. However, the cask models with flooded assemblies exhibited similarities to many experiments with c k values greater than 0.8.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Delayed Critical and Subcritical Experiments with Polyethylene Moderated Unreflected Thin 15 in. Diameter HEU Metal Plates

The thin ~15 in. diameter highly enriched uranium (HEU) metal plates were assembled to delayed criticality at the Oak Ridge Critical Experiments Facility (ORCEF) in 1969 with various thicknesses of polyethylene (varying from 1/16 to 2$\frac{3}{8}$ inches) between uranium metal plates. The average 235 U enrichment was 93.27 wt. %. These unreflected critical configurations contained 4$\frac{2}{3}$ to 20$\frac{5}{6}$ thin 15 in. diameter HEU metal plates (on loan from Los Alamos National Laboratory [LANL] and shipped to Oak Ridge National Laboratory [ORNL] on June 3, 1969). Depending on the thickness of polyethylene, the enriched uranium masses varying from 28,053 to 135,148 grams. Fractional plate sections consisted of the appropriate number of 60° pie sections. In addition to the measurement at delayed criticality, subcritical measurements were also performed by the inverse kinetic rod drop method. Prompt neutron decay constant measurements were also performed by the Rossi alpha and randomly pulsed neutron method using a time-tagged spontaneous fission californium neutron source; these are briefly reported here. At the time of these measurements in 1969, the thin HEU metal plates were in near-pristine condition with extremely little oxidation, allowing better descriptions of the uranium plates than the use of these plates in a heavily oxidized and deteriorated condition in recent reflected benchmark experiments at the LANL facility at the Nevada Test Site with these same thin highly enriched uranium metal plates. This report documents the experimental information for the measurements performed so that later researchers can perform the required uncertainty and calculational analyses and documentation to use these data for an International Nuclear Criticality Safety Benchmark Evaluation Program (ICSBEP) or a Nuclear Energy Agency (NEA) benchmark. Data from the experiments described should be acceptable for use as criticality safety benchmark experiments for the ICSBEP and the NEA nuclear criticality safety benchmark program once the uncertainty analysis on the measured neutron multiplication factors is completed. Additional data—such as the dimensional inspection reports, uranium isotopic information, and other relevant particulars—should be retrieved from the Y-12 Plant or LANL and incorporated in the final ICSBEP benchmark. Based on previous ICSBEP benchmarks with this enriched uranium metal at ORCEF, the uncertainties in $k_{eff}$ could be as low as ± 0.0002 for some configurations. Other experiments with smaller-diameter than 15 in. diameter HEU metal plates have been benchmarked in HEU-METFAST-001. The prompt neutron time decay measurements could be the basis for an International Reactor Physics Benchmark Program. Preparation of the present report is part of an effort at ORNL to document more than 15 undocumented critical and subcritical experiments enumerated in ORNL/TM-2019/18 and performed by ORNL at ORCEF and other US Department of Energy critical experiments facilities using more than 500 operational days of critical facility time. This work for this report publication was supported by the Nuclear Criticality, Radiation Transport and Safety NCSP Program at ORNL.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Prototype for Unclassified Algorithm

Continuing previous efforts to investigate and develop the Unclassified Radioisotope Algorithm, the goal of the FY19-FY20 effort was to develop a prototype detector system which uses the algorithm to confirm warhead attributes related to the presence of either weapons grade plutonium (WGPu) or highly enriched uranium (HEU). The final deliverable is a prototype attribute measurement system built with common, commercially available gamma radiation detector components, capable of confirming the presence of specific, complex radioactive sources of interest, without the collection and storage of gamma energy spectra. This is accomplished by processing each gamma pulse as it is received, applying weight values based on the energy and incrementing or decrementing scalar counters which can be compared with expected values to determine if the measured source is consistent with WGPu or HEU. This report documents the design of the prototype system as well as the development of the algorithm and performance testing results. While the previously conceptualized, simple algorithm resulted in a prohibitive amount of false positives, the goal for a simple attribute measurement system capable of verifying Ba-133 and Ra-226 (weapons grade plutonium and highly enriched uranium surrogate testing sources) at over 95% accuracy with sub 5% false positive rate was demonstrated.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2024

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

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

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

A general mechanism for gel layer formation on borosilicate glass under aqueous corrosion

Mineral and glass dissolution is a scientific topic deeply investigated but incompletely understood and of a great interest for the geochemical and materials science communities. If the interfacial dissolution/reprecipitation mechanism seems to be applicable to most of silicate minerals, the debate remains open concerning glass. Here we studied two model glasses, a ternary borosilicate (CJ1) and the same glass doped with 4.1 mol% of Al2O3 (CJ2). The two glasses were altered at 90°C, pH 9, and in conditions far and close to saturation with respect to amorphous silica, to determine the initial and residual rates. Moreover, a specific experiment was conducted for a short duration with a solution highly enriched with 18O and 29Si isotopes to understand how passivating gels form. SEM, TEM and ToF-SIMS characterization, along with Monte Carlo simulations were used to understand the rate limiting reactions at play and infer the role of Al. We show that Al yields a slower matrix dissolution in dilute conditions. However, it slows down he formation and the maturation of the passivating gel and favors alteration by partial hydrolysis of Si and Al entities followed by in-situ reorganization/relaxation into a porous network. Unexpectedly, CJ1 experienced both interfacial dissolution/reprecipitation and partial hydrolysis followed by in situ reorganization of the silicate network during the course of a single experiment. This study offers a unified concept that can pave the way for the future development of a predictive kinetic model based on a detailed description of bond breaking and bond forming as a function of glass composition and alteration conditions.

Gin, Stephane↗
DOE OSTI2020

Aqueous passivation of multi-principal element alloy Ni 38 Fe 20 Cr 22 Mn 10 Co 10 : Unexpected high Cr enrichment within the passive film

The aqueous passivation of a non-equiatomic Ni 38 Fe 20 Cr 22 Mn 10 Co 10 - at.% (Ni 40 Fe 20 Cr 20 Mn 10 Co 10 - wt.%) multi-principal element alloy (MPEA) was investigated in 0.1 M NaCl at pH 4 and compared to a conventional binary Ni 76 Cr 24 – at.% (Ni 78 Cr 22 – wt.%) alloy. The electrochemical behavior and oxide film characteristics were explored utilizing in-situ electrochemical and ex-situ surface-sensitive techniques. The passive film composition, thickness, and elemental valence states, as well as, the fate of each element were studied by in-situ atomic emission spectro-electrochemistry, ex-situ X-ray photoelectron spectroscopy, and atom probe tomography. The Ni 38 Fe 20 Cr 22 Mn 10 Co 10 MPEA demonstrated slightly better corrosion resistance compared to the binary Ni 76 Cr 24 , alloy. Passive films on the MPEA contained primarily Cr, and small amounts of Ni, Fe, Co and Mn, while dissolution of Ni, Fe, Co was observed. Ni 0 was enriched at the oxide/metal interface while Cr was depleted. Enrichment of Cr in the passive film occurred to a greater extent in the MPEA than for the Ni-Cr binary alloy. As a result, enrichment factors were determined and the origins of enrichment are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Alternate Methods for Cleaning Zirconium Plate

The United States High Performance Research Reactor Conversion Program identified the need to transition from highly enriched uranium fuel to lowly enriched uranium fuel for high power research and civil reactors. One of the fuel configurations selected for these reactors was a uranium (U) foil alloyed with 10% molybdenum (Mo). The U-10Mo fuel foil was then covered with a zirconium (Zr) interlayer and pressed into an aluminum cladding. Idaho National Laboratory prescribed a need to remove surface oxidation on the Zr prior to bonding it with U-10Mo fuel foil. To determine the best method of oxidation removal, a variety of polishing methods, including hand polishing with a diamond paste or Scotch-Brite™ pads, mechanically wet-polishing with abrasive belts, and chemical etching with hydrofluoric/nitric acid mixtures, were investigated on pre-polished Zircaloy 702 sheets purchased directly from the manufacturer. No change in surface roughness was detected using any of these methods relative to the as-received material. In fact, scanning electron microscopy detected grit from hand polishing with Amplex Grade 30 waste soluble diamond paste and Scotch-Brite™ pads. Visible staining was found, and fluorine detected via X-ray photoelectron spectroscopy on the flash etching samples. These contaminants were not seen on the as-received material. In addition to the need to remove the oxide layer on the Zr prior to co-rolling, there are concerns that the Neolube debonding agent on the co-rolling can may contaminate the Zr surface after co-rolling. The experimental results found that zirconium in contact with Neolube coated steel plates was easily cleaning with ethanol wiping after heat treated and no carbide inclusions remained on the surface. Overall, all cleaning methods tested did not significantly change the surface roughness of the plates or reduce the oxide layer present on the zirconium. Carbon was found on all the samples, before and after cleaning, except for the flash etch samples which detected fluorine instead. The insignificant changes seen in the cleaned plates relative to the as received plates question the need for cleaning prior to rolling. Further investigation in the requirements for cleaning at this step is recommended. Current methods for cleaning the surface after hot rolling with alcohol appear sufficient. After hot rolling, if Neolube does peel off the can and onto the zirconium coated foil surface, a dry abrasive or diamond paste polish is not recommended due to the chance of abrasive being embedded into the surface. However, the ability of an etchant to lift the lubricant off the zirconium surface indicates that a spot etch method could work to clean specific spots where Neolube is present, if the need arises.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Commercialization of High-Density High Assay Low Enriched Uranium Fuel Systems

The Office of Reactor Conversion and Uranium Supply (NA 231) at the National Nuclear Security Administration leads the conversion effort for the United States High Performance Research Reactors (USHPRR). These reactors are the final civilian reactors in the US to transition from High Enriched Uranium (HEU) to high assay low enriched uranium (HALEU). Each of these reactors represents unique capabilities and no currently available fuel system meets their needs for conversion. The Fuel Fabrication (FF) Pillar of the USHPRR project is responsible for the fabrication of experimental elements, conversion elements, and establishing a commercial economical production capability. FF is also responsible to share with the domestic and international community the theoretical knowledge gained. Other pillars within the USHPRR project provide the experimental and conversion fuel designs, assist the reactors with licensing activities, and ensure the entire fuel cycle is evaluated. Over the last decade, FF has worked with the production partners at Y-12 National Security Complex (Y-12) and BWXT Nuclear Operations Group, Research and Test Reactors (BWXT). Y-12 has begun processing the alloy feedstock for the conversion elements with a qualified process. BWXT has started the final fabrication of the experimental elements. Once the experimental elements are complete, BWXT will begin conversion element fabrication. The FF Pillar resides at Pacific Northwest National Laboratory (PNNL) and uses PNNL, universities, commercial vendors, and the DOE national laboratory system to evaluate process development activities to improve the process steps. FF supports the fabrication of two high density fuel systems, monolithic U-10Mo (Figure 1) and Uranium Silicide (Figure 2). The U-10Mo fuel system is further along the development process. FF assists in long term planning with the production partners. This includes ramping production of the elements from experimental quantities to annual steady state needs. As part of the ramp up, opportunities to improve yield and product quality are identified to ensure the fuel systems are cost effective.

Catalan, Michael A. [BATTELLE (PACIFIC NW LAB)]↗
DOE OSTI2022

Delayed onset of discontinuous precipitation-based phase transformation in U10Mo alloys doped with Silicon

A uranium-10 wt% molybdenum (U-10Mo) alloy is one of the primary candidates for metallic fuels that would use low-enriched uranium in place of highly enriched uranium, to support nuclear nonproliferation efforts. Optimal performance of a U-based metallic nuclear fuel can be achieved by retaining the high-temperature, body-centered cubic (bcc) allotrope (γ-U) at room temperature, which can be accomplished in the U-10Mo alloy. However, presence of minor alloying elements can influence the final constitution of room-temperature phases in the U10Mo alloy, specifically, formation of α-U phase which results in anisotropic behavior of the fuel in reactor. Further, through a detailed transmission electron microscopy analysis, the present study reports the constituent phases that are present in a U10Mo alloy containing ~0.1 wt% Si after it is subjected to homogenization heat treatment and thermomechanical processing. For comparison, results from an undoped U10Mo alloy are also included. The experimental results reveal that γ-UMo solid solution is the major phase in a hot-rolled, Si-doped U10Mo alloy metallic fuel foil, along with U 2 MoSi 2 C, UC, and U 2 Mo, after isothermal annealing at 460 °C for 10h. In contrast, after the same heat treatment, the undoped U10Mo alloy metallic fuel had formed a noticeable amount of α-U along prior γ-UMo grain boundaries through discontinuous precipitation (DP, area fraction: ~27.9%) with characteristic lamellar morphology, together with γ-UMo, UC, and U 2 Mo. This result indicates that doping with Si could mitigate the DP reaction in U10Mo alloy and prevent formation of undesirable α-U. This work sheds light on optimizing Si-doping–dominated microstructure in U10Mo fuels and facilitates designing and tuning of microstructures of U10Mo alloys for tailoring the final designed performance of the fuel under irradiation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗