Oxygen Isotope and Fluorine Impurity Signatures during the Conversion of Uranium Ore Concentrates to Nuclear Fuel
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37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
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VERIFICATION MEASUREMENTS OF THE MASS AND ENRICHMENT OF URANIUM OXIDE CARD SOURCES
Measurements were performed to determine the mass and enrichment of 10 uranium oxide (U3O8) card sources. The measurements and analysis were completed as a verification of the card sources in support of Oak Ridge National Laboratory’s nuclear material control and accountability program. Although these cards are not nationally accredited as a nuclear standard, they are used as a working reference for measurements, such as holdup. The uranium card source measurements were taken with a broad energy germanium detector and the Genie 2000 Gamma Acquisition and Analysis software. A complete characterization of each of the 10 uranium card sources was performed using the 4 characteristic full energy peaks of 235U. Using the In Situ Object Counting System software to determine the mathematical efficiency of the measurement, the mass of 235U in each card was determined. The 235U mass in each card ranged from 10.42 to 12.63 g with a systematic error between 0.76 and 0.95 g and a random error of 0.01 g for each card source. The Multi-Group Analysis for Uranium (MGAU) software and the Fixed-Energy, Response Function Analysis with Multiple Efficiency (FRAM) isotopic analysis software were used to determine the isotopic composition of the uranium cards. The measured enrichment was compared to the declared enrichment for each card, with uncertainties ranging from 2.7% to 4.3% for the MGAU analysis and 2.3% to 4.1% for the FRAM analysis. This is a good example of how a well-benchmarked mathematical calibration method can be useful in characterizing uranium sources.
Hunneke, Rachel↗
Preliminary Insights Into the Feasibility of Determining the Purification Date of Enriched Uranium by Direct Measurement of the 230 Th/ 234 U Ratio Using an All-Faraday Detector Configuration on the Neoma MC-ICP-MS
Rationale: Mass spectrometric measurement of the 230 Th/ 234 U ratio to calculate the purification age of enriched uranium is typically conducted via a combination of ion counters and faraday detectors, thus requiring an inter-detector calibration scheme. Here, our aim is to understand whether the pursuit of a simplified measurement scheme involving only faraday detectors is feasible. Methods: We investigate the possibility of determining U-Th model ages for two enriched uranium standards (NBL U630 and U850) by direct measurement of the 230 Th/ 234 U ratio (without chromatographic separation or isotope dilution) on a ThermoFisher Scientific Neoma MC-ICP-MS utilizing both solution and laser ablation (LA)-based sampling techniques and an all-faraday detector configuration. Results: For the solution mode analyses conducted on aliquots containing sub μg/mL total U, we produce composite average 230 Th/ 234 U model dates of May 19, 1988 (± 351 days), and March 26, 1961 (± 2.5 years) using the directly measured 230 Th/ 234 U ratios for the NBL U630 and U850 uranium standards, which have certified purification dates of June 6, 1988 (± 190 days), and December 31, 1957 (± 36.5 days), respectively. The ages produced by LA-based sampling of dried residues of the same standards deposited onto cotton TexWipes are less accurate and of poorer precision (June 23, 2004 ± 8.7 years for U630 and December 21, 1965 ± 7.9 years for U850) but still yield meaningful information in regards to the purification date. Conclusions: We believe that further refinement of the all faraday detector measurement approach to include development of a more robust Th/U relative sensitivity factor determination, signal cutoff selection, and data processing protocols will allow for this approach to be confidently applied to enriched uranium materials with unknown purification histories. Potential advantages of the method include the reduced sample handling and infrastructure requirements as well as the ability to simultaneously generate a broad picture of the uranium isotopic composition in tandem with the U-Th age determination.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Reactor Physics Characteristics of High Flux Isotope Reactor Core Designs with Low-Enriched Uranium Fuels
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Betzler, Benjamin↗
Electrothermal vaporization as a new method for pre-inspection check samples
The use of electrothermal vaporization (ETV) has not been readily applied to nuclear safeguards or pre inspection check (PIC) sample analysis, even though the technique is ideally suited for low level (pg) isotopic quantification due to its high analyte transport efficiency. Here demonstrate the ETV method for PIC safeguard analysis. The ETV system is an automated, commercial of the shelf (COTS) technology, with the capability to introduce solid samples to a mass spectrometer platform. The system has been shown to simultaneously determine uranium assay, and the major isotopic ratios for uranium and plutonium. The ETV unit has demonstrated analysis up to 50 samples a day autonomously. The utilization of this ETV methodology provide a secondary option to the current PIC analysis and offers high throughput, enhanced sensitivity, and broader isotopic detection platform.
Hexel, Cole↗
Acceleration of uranium beam to record power of 10.4 kW and observation of new isotopes at Facility for Rare Isotope Beams
The Facility for Rare Isotope Beams (FRIB) is a major nuclear physics facility for research with fast, stopped, and reaccelerated beams that was successfully commissioned in May 2022. A key capability of FRIB is the production of an acceleration of the uranium beam, but this capability requires the facility to work at the design limits of the lowest charge-to-mass ratio and the highest power density on the beam intercepting devices. This paper presents techniques for overcoming the significant challenges in accelerating the uranium beam, culminating in the demonstration of 10.4 kW on target, and the discovery of three new isotopes. The high-power uranium beam enabled us to produce and identify G 88 a , A 93 s , and S 96 e , within the first 24 h of operation. The successful uranium operation at FRIB sets a new record for accelerated uranium beam power above 10 kW and opens a new avenue of research with rare isotopes. Published by the American Physical Society 2024
43 PARTICLE ACCELERATORS↗
A Comprehensive Review of Selected Major Categories of Lithium Isotope Separation Techniques
The ability to generate materials with an enriched isotopic abundance, that is one that differs from natural abundance, is critical in terms of technology. Isotopes are used in a wide range of applications, including radioactive tracers, nondestructive tests, radiotherapy in human medicine, nuclear fuels, and isotope‐substituted compounds for chemistry and biology research, as well as environmental geochemical signature tracers. A commonly discussed research topic is the isotopic separation of various elements—uranium, hydrogen, lithium, and iodine, among many others. Among these isotopes, lithium isotopes are especially important in various applications including strategic areas. These isotopes are light and hence their separation techniques are not very obvious. Herein, all the main categories of Li isotope separation are reviewed in depth with a focus on electrochemical technologies for stable lithium isotope separation.
Murali, Arun↗
Tracking natural CO 2 migration through a sandstone aquifer using Sr, U and C isotopes: Chimayó, New Mexico, USA
The geochemical and isotopic characteristics of groundwaters in Chimayó, New Mexico, reflect processes that affect water quality in the Tesuque Aquifer, which overlies a leaking natural CO 2 source in a structurally complex region. In this study, select isotopes (δ 13 C, 87 Sr/ 86 Sr, 234 U/ 238 U) are applied to groundwaters to better understand CO 2 transport mechanisms and related water-rock interactions that impact Chimayó groundwater. Carbon stable isotope ratios of dissolved inorganic carbon (DIC; δ 13 CDIC = -15.10‰ to 4.50‰) identify a distinct source of upward-migrating CO 2 that interacts with the groundwater. Additionally, groundwater 87 Sr/ 86 Sr compositions (0.7098 to 0.7154) reflect intrusion of varying amounts of saline water associated with the high CO 2 source, while 234 U/ 238 U ratios corroborate presence of deep groundwater and suggest impacts from distinctive natural uranium sources are affecting groundwater. Previous work proposed two CO 2 transport mechanisms at the site: (1) dissolved in deep brine that underlies the aquifer and (2) in the gas phase; this study uses isotope mixing models to identify wells that are affected by these CO 2 transport mechanisms and demonstrates that both transport mechanisms are associated with impaired groundwaters. Overall, this study demonstrates that applying multiple isotope systems (δ 13 CDIC, 87 Sr/ 86 Sr, 234 U/ 238 U) is a dynamic tool for identifying and measuring the impact of CO 2 leakage from a sequestration site.
58 GEOSCIENCES↗
Molybdenum and Tungsten isotope compositions of UO 2 fuel pellets: Implications for isotopically enriched taggants
The addition of isotopically enriched taggants to material at the front end of the nuclear fuel cycle could be a powerful tool used to assist law enforcement authorities should material outside of regulatory control be found. Two potential candidates for this purpose are molybdenum (Mo) and tungsten (W) as both elements have five or more stable isotopes and are trace elements contained within nuclear fuel. So there is a concern that Mo and W could undergo isotope fractionation during processes like uranium enrichment and to date, it is unknown if nuclear fuels have natural Mo and W isotope compositions. If Mo and W isotopic variability is present in nuclear fuels, this would hinder the use of these elements as isotopic taggants because it would be difficult to discern the original taggant isotope composition with high confidence. Therefore, a set of 16 low enriched uranium (LEU) fuel pellets from US commercial producers was analyzed using multi collector-inductively coupled plasma mass spectrometry (MC-ICPMS) to determine Mo and W isotope compositions (i.e., 94 Mo/ 92 Mo, 95 Mo/ 92 Mo, 96 Mo/ 92 Mo, 97 Mo/ 92 Mo, 98 Mo/ 92 Mo, 183 W/ 182 W, 184 W/ 182 W, and 186 W/ 182 W). Relative to terrestrial standards, LEU fuel pellets have variable Mo and W isotope compositions, thereby complicating the use of these elements as isotopically enriched taggants. As such, this work demonstrates that the isotope composition of any potential taggant must be well characterized in the base nuclear fuel prior to any taggant addition. Furthermore, these results suggest that Mo and W are not ideal candidates for isotopically enriched taggants.
58 GEOSCIENCES↗
FRAM Version 7.1’s Bias
The Fixed-Energy Response-Function Analysis with Multiple Efficiency (FRAM) code was developed at Los Alamos National Laboratory to measure the gamma-ray spectrometry of the isotopic composition of plutonium, uranium, and other actinides. For FRAM versions 4 and earlier, the reported uncertainties of the results come from the propagation of the statistics in the peak areas only. No systematic error components are included in the reported uncertainties. For FRAM versions 5 and 6, we examined the FRAM analytical results of both the archival plutonium data and the data specifically acquired for the isotopic uncertainty analysis project and found the relationship between the bias and other parameters. We worked out the equations representing the biases of the measured isotopes from each measurement using internal spectral parameters, such as peak resolution and shape, region of analysis, and burnup (for plutonium) or enrichment (for uranium). The resulting biases were included in the reported uncertainties of FRAM v.5 and v.6.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Spatially Resolved Characterisation of Low Mass Fraction Uranium Glass Working Reference Materials
Here, we present the results of a study to generate reference glasses that reflect an environment analogous to historic nuclear fallout samples of interest for post-detonation nuclear forensics. The glasses were generated by melting and then quenching SiO 2 , Al 2 O 3 and CaCO 3 powders. Two suites of glasses with three distinct U isotopic ratios were successfully made with enrichments in the 235 U isotope (~ natural [0.72%], ~ 53% and 94%), but the bulk elemental data showed heterogeneity (~ 10% RSD) with U mass fractions ranging from 331.47 to 373.63 μg g -1 . Spatially resolved U isotopic measurements were performed using three mass spectrometry techniques (secondary ion mass spectrometry-single stage accelerator mass spectrometry [SIMS-SSAMS], large geometry [LG] - SIMS, and laser ablation-inductively coupled plasma-mass spectrometry [LA-ICP-MS]) across five National Laboratories. The results showed good agreement with the bulk U isotopic data for the low, medium, and high U mass fractions. We conclude that despite elemental heterogeneity, these samples can serve as useful working reference materials for spatially resolved nuclear fallout analyses, as well as for other related spatially resolved analyses.
36 MATERIALS SCIENCE↗
Destructive Analysis of TRISO Particles: Crush/Burn/Leach Followed by Davies-Gray Titration and IDMS
The accurate accounting of nuclear materials is a cornerstone of international nuclear safeguards. One emerging challenge in this domain is the fabrication of TRIstructural ISOtropic (TRISO) particle fuels. Although these innovative fuel forms are critical for advanced reactor applications, their robust refractory ceramics and coating compositions present significant obstacles to destructive analysis (DA) methods. Ensuring full and quantitative recovery from these particles is essential for accurate mass accountancy. The current study was initiated to address these challenges, first by validating a previously established destructive method developed by Oak Ridge National Laboratory (ORNL) for the quantitative recovery of uranium from TRISO particles and then following that process with uranium content determination through isotope dilution mass spectrometry (IDMS) and Davies-Gray titration. This study expands on the scope of a digestive method that was developed under the Advanced Gas Reactor Fuel Development and Qualification program and is currently implemented in both the Coated Particle Fuel Development Laboratory and Irradiated Fuels Examination Laboratory at ORNL. The success of the previous Advanced Gas Reactor work relied on developing a DA method to evaluate the fabrication process and reactor experiments. The methodology described in this report was designed to rigorously investigate the efficacy of the crush/burn/leach sample preparation of TRISO particles; it aims to quantify uranium recovery while also assessing the effects of TRISO constituents (e.g., silicon and zirconium) on analytical precision and accuracy. By comparing the results from the titration method and IDMS, we sought to determine whether existing analytical procedures accepted by the International Atomic Energy Agency (IAEA) could be effectively translated to TRISO fuel forms. The team employed an approach that involved processing replicate TRISO samples, optimizing the milling (i.e., crushing) step and performing serial leaches. The elemental composition of the analytical samples was examined to prepare for interference studies in the second year of this project. The integration of gamma spectrometry to verify residual uranium activity further strengthened the validation. Statistical methods were applied to the collected data to evaluate the uncertainties arising from sampling, sample preparation and uranium quantification. These uncertainties were then compared to the IAEA’s international target values (ITVs). Additional data collected in upcoming project work will strengthen the uncertainty estimates. Ultimately, it is hoped that this project will contribute materially to the body of work related to characterization of TRISO based fuels for the purpose of material accountancy and its applications to international safeguards.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Preliminary Thermal-Hydraulic Analysis of the UTA–2 Subcritical Linear Accelerator–Driven System
The National Nuclear Security Administration’s mission includes establishing a reliable supply of 99 Mo without highly enriched uranium. Oak Ridge National Laboratory (ORNL) supports this objective through collaborative research and development with industrial partners. Niowave Inc., a current partner, is currently designing a subcritical linear accelerator-driven system (ADS) and preparing for the US Nuclear Regulatory Commission’s licensing process. Niowave’s system has the potential to efficiently supply medical radioisotopes. The technology includes a superconducting electron accelerator and a pile of both natural and low-enriched uranium (LEU) targets. The process fissions uranium and many valuable isotopes can then be extracted from the targets. Niowave is currently iterating through conceptual and detailed design processes for several system sizes. This report discusses UTA–2, which is at the demonstration stage. UTA–2 will validate numerical modeling results with experimental measurements before progressing to the detailed design of UTA–3, the commercial-sized ADS. The thermal-hydraulic behavior of the UTA–2 core design was numerically investigated using STAR-CCM+, as described in this report. STAR-CCM+, a state-of-the-art computational fluid dynamics (CFD) software that was commercially developed by Siemens, has an extensive user base and a set of validation studies. It is also compliant with the American Society of Mechanical Engineers’ Nuclear Quality Assurance 1 standard. Three cases are investigated in this report. Case 1 quantified the temperature field within the UTA–2 assembly and water tank using only conduction as the method of thermal energy transport. This simplified approach was overly conservative and yielded wetted cladding temperatures above the coolant saturation temperature. In this case, the maximum temperature of the wetted cladding surface of the highest power rod exceeded the saturation temperature of water by 63.8°C. Because of the overly conservative approach taken in Case 1 and its negative subcooled margin, buoyancy-driven natural circulation flow physics were implemented in Case 2. Adding coolant motion significantly distributed the thermal energy of the system through convective heat transfer. This relatively small amount of convective heat transfer significantly reduced system temperatures and increased the subcooled margin from –63.8 to 61.3°C. This margin confirmed that no boiling was expected during normal operation of UTA–2 at 230 W. Case 3 had no additional physics models but considered an overpower event during which the power of each LEU and natural uranium rod was at its respective peak values. This resulted in a study with the total assembly power equal to 176% of the nominal power of 230 W considered in Cases 1 and 2. The resulting natural circulation flows were slightly enhanced. The subcooled margin decreased slightly to 46.3°C, which is still a significant margin to local boiling of the water within the tank. This margin confirmed that no boiling was expected during an abnormal operation of UTA–2 at 406 W.
07 ISOTOPE AND RADIATION SOURCES↗
Computational Thermal Hydraulics of a High-Performance Low-Enriched-Uranium Annular Target for HFIR Irradiation
Molybdenum-99 has historically been generated via isolation from fissioned highly enriched uranium (HEU) targets. Here, this isotope is in high demand due to its daily use across the world in radiopharmaceutical medical procedures. The primary objective of this work was to design and analyze an experimental target assembly containing one low-enriched-uranium (LEU) annular target for irradiation at the High Flux Isotope Reactor (HFIR). Efforts included incorporating spatially dependent energy sources from neutron and gamma interactions, quantifying thermal contact conductance at material interfaces, performing grid-independent studies, comparing turbulence models, and simulating various steady-state and transient scenarios relevant for irradiation qualification and eventual insertion. These models provide velocity, pressure, and temperature distributions in both space and time. Such results enable the selection of an appropriate irradiation location, fission rate density, and flow-limiting orifice size and demonstrate compliance with HFIR safety requirements such that insertion into the reactor can be approved. This analysis shows that across all scenarios, wetted surface temperatures remain below the coolant saturation temperature with no net vapor formation in the coolant. In every scenario, all components stay below 30% of the aluminum 6061 melting temperature. Computational fluid dynamics and system-level models predict peak target temperatures that agree within 4%, though the predicted axial location of the peak differs by about 10% of the heated length due to differences in flow development length. These results de-risk the irradiation of LEU (annular targets) and strengthen a domestic, HEU-independent 99 Mo supply by providing important fuel performance data to form the foundation for a robust licensing basis.
Molybdenum-99↗
230Th/234U radiochronometry for uranium materials by alpha spectrometry for nuclear forensics analysis
We have developed an alpha spectrometry method for 230Th/234U radiochronometry to determine the model separation age for uranium materials. This method offers an alternative to the more commonly used, but significantly more resource intensive multi-collector inductively coupled plasma mass spectrometry (MC-ICP-MS) while providing similar accuracy and precision. The development included radiochemical separation of uranium and thorium, calibration of the 232U tracer and the alpha spectrometry measurement. The method was validated by analyzing a certified reference material, CRM 125-A, which is a uranium oxide pellet assay, isotopic, and radiochronometric standard. The results were compared to those obtained by our routine MC-ICP-MS 230Th/234U radiochronometry analysis method to assess the performance parameters for alpha spectrometry compared to the state-of-the-art technique.
Macsik, Zsuzsanna [Los Alamos National Laboratory ↗
Evaluation of Hydrogen Gas Generation and Permeation in the 9979 Type AF Shipping Package
The 9979 Type AF Shipping Packaging is a cost-effective radioactive material package designed by Savannah River National Laboratory (SRNL) that consists of two primary components: a foamed outer drum for structural protection and an inner containment drum. The packaging was designed to transport Highly Enriched Uranium (HEU), Low Enriched Uranium (LEU), and other isotopes not exceeding a Type A quantity. These contents have the potential to generate flammable hydrogen gas during transport due to the degradation of hydrogenous materials (e.g. water vapor, plastics, etc) by high-energy alpha radiation. Since 10 CFR 71.43(h) prohibits the incorporation of packaging features explicitly designed for continuous venting, alternative justification was required to demonstrate that the requirements of 10 CFR 71.43(d) for negligible reactions between packaging components and contents were satisfied. An analysis was performed to demonstrate that the potential for hydrogen gas generation over a one-year period was limited by the effects of permeation through the packaging materials. The rate of hydrogen collection was evaluated for both the inner containment drum and the outer structural drum under 10 CFR 71.71 Normal Conditions of Transport (NCT). The analysis concludes that the Lower Flammability Limit (LFL) for hydrogen gas will not be reached in a one-year shipping period assuming a minimum void volume is maintained within the drum.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
SCALE 6.2.4 Validation: Nuclear Criticality Safety
The computational bias of criticality safety computer codes must be established through the validation of the codes to critical experiments. A large collection of suitable experiments has been vetted by the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and made available in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook). More than 600 cases from this handbook have been prepared and reviewed within the Verified, Archived Library of Inputs and Data (VALID), which is maintained by the Reactor and Nuclear Systems Division at Oak Ridge National Laboratory. The performance of the KENO V.a and KENO-VI Monte Carlo codes within the SCALE 6.2.4 code system is assessed using the VALID models of benchmark experiments. A range of nuclear cross section libraries based on Evaluated Nuclear Data File (ENDF)/B-VII.1 in both multigroup (MG) and continuous energy (CE) formats is considered. The critical experiments available to validate the KENO V.a code cover 15 broad categories of systems. These systems use a range of fissile materials, including a range of uranium enrichments, various plutonium isotopic vectors, and some mixed uranium/plutonium oxides. The physical forms of the fissile material also vary and are represented as metal, solutions, or arrays of rods or plates in a water moderator. The neutron energy spectra of the systems also vary and cover fast, intermediate, mixed, and thermal spectra. Over 550 of the total cases use the KENO V.a code for the four nuclear data libraries considered in this report.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Diversion Path Analysis: A Proposed Methodology to Develop an MC&A Approach for Liquid-Fueled Molten Salt Reactors
Nuclear material control and accounting (MC&A) is a critical element of both the US Nuclear Regulatory Commission (NRC) and US Department of Energy (DOE)’s domestic safeguards and security requirements. NRC licensees are required, under Title 10 of the Code of Federal Regulations (10 CFR) Part 74 to establish and maintain an MC&A program that captures and records the quantities and locations of special nuclear material (SNM) at the facility. Along with physical protection, MC&A is a key element of domestic nuclear material safeguards that enables the NRC to ensure that SNM is controlled and accounted for. SNM, per 10 CFR Part 74, refers to plutonium, 233 U, and uranium enriched in the isotope 233 U or 235 U, but does not include source material. Periodic physical inventories, coupled with material balance evaluations, are effective and demonstrated tools to account for and detect theft or diversion of SNM in facilities containing SNM in bulk material form (i.e., not in discrete, countable items). Historically in the United States, these types of facilities have included fuel fabrication, conversion, and enrichment facilities. In comparison, reactors have relied on item counting of assemblies and control of SNM while in containment (e.g., a sealed reactor pressure vessel) because, to date, reactor fuel has been in item form. In liquid-fueled molten salt reactors (MSRs), unlike traditional light water reactors (LWRs) or bulk facilities, bulk SNM quantities can change significantly during operation as a result of depletion and transmutation. This introduces challenges to the use of traditional periodic physical inventories and material balance evaluations to detect theft or diversion of SNM in reactors that use SNM in bulk material form. Liquid-fueled (i.e., salt-fueled) MSR facilities are MSRs that use SNM within a salt eutectic as the fuel. The SNM is in a bulk material form any time it is outside of fresh or spent fuel storage containers. Some examples of when SNM will be in bulk form in the facility are during addition of fuel to the reactor system, while fuel is circulating in operation, and while fuel is in a drain tank. Periodic physical inventories and material balance evaluations can likely be effectively applied to many portions of an MSR facility, including all areas where depletion and transmutation are not significantly changing the quantities of SNM within the control area. Within an MSR facility, this would include fresh fuel receipt and loading, waste streams that may contain SNM, irradiated fuel storage outside of the reactor core, and any irradiated fuel processing that may happen after SNM has been removed from the reactor. All of these process steps could rely on measurements of SNM quantities compared with documented inventories. Any discrepancies from predicted (i.e., book) inventories and measured inventories could be quantified as inventory differences, consistent with traditional MC&A guidance from the NRC (e.g., in NUREG-1065 Revision 2, NUREG-2159 Revision 1, and RG 5.29 Revision 2). Within the reactor system, additions and removals to the book inventory include depletion of the SNM (e.g., fission of 235 U), which complicates the use of physical inventories. SNM control, however, can also likely be effectively applied to detect theft of SNM throughout a liquid-fueled MSR facility. To complement these approaches, prior technical reports have identified that a diversion path analysis may be a useful, risk-informed, and performance-based tool to determine suitable elements of an MC&A approach for the reactor system within a liquid-fueled MSR facility.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗