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At least 379 records · Page 21
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

MCNP Neutronic Design Optimization of an Accelerator-Driven Subcritical Assembly for Mo-99 Production

Molybdenum-99 (99Mo) has been in short supply in the recent decades. The most common productionmechanism, irradiation in a reactor fueled by highly enriched uranium (HEU), raisesproliferation concerns. Novel efforts are being pursued to meet the demand for 99Mo withoutHEU irradiation, relying on the processing of low-enriched uranium or natural uranium. Onesuch concept is to use an electron linear accelerator to generate bremsstrahlung that, through aphotonuclear reaction, can help multiply neutrons in a subcritical assembly. The goal of this workwas to neutronically optimize the subcritical assembly for keff and 99Mo production. This optimizationis unique for The design was optimized in terms of reflector type, pitch, and LEU andNU mass. Results of the paper show the effects of fuel mass, pitch, and reflector material on keffand the NU fission fraction. keff was found to be mostly driven by LEU mass, and the NU fissionfraction was dependent on several parameters, primarily pitch and mass.

Conant, Andrew↗
DOE OSTI2016

Role of Core Exit Axial Reflectors for NTP Full Submersion Criticality Accident Mitigation

One of the only design basis accidents currently codified into law for space nuclear systems is the full-submersion criticality accident. This accident scenario prescribes that the reactor core must remain sub-critical in the event of a full water submersion with the control system in the shut-down position. This accident scenario is of particular interest and relevance for nuclear thermal propulsion (NTP) reactors because of their unique core periphery and internal design considerations. In previous work it has been shown that this accident is particularly sever for low-enriched uranium (LEU) fueled NTP cores due to their moderated spectrum and the minimization of parasitic neutron absorption. These then add to the existing problems associated with the numerous voids within the core and the extensive neutron leakage due to its compact size. In this abstract we examine the role of a core exit axial reflector in minimizing the submersion worth of a graphite composite LEU-NTP core. The neutronics simulations were done using the ENDF/B-VII cross-section library in conjunction with MCNP6.1. The MCNP calculations were done using 20000 particles per cycle with 50 inactive and 300 active cycles. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Monolithic U-10Mo Fuel Development for U.S. High Performance Research Reactors

To support the U.S Department of Energy (DOE), National Nuclear Security Administration (NNSA) mission to reduce the threat posed by the civilian use of highly enriched uranium (HEU) worldwide, the US High Performance Research Reactor (US HPRR) fuel development program is pursuing development and qualification of a new high-density monolithic U-Mo fuel to facilitate conversion of five higher power research reactors located in the U.S. (ATR, HFIR, NBSR, MIT and MURR) and one critical facility (ATR-C) from HEU to low enriched uranium (LEU). These reactors require higher uranium fuel density than is achievable with dispersion fuel systems used to convert lower power research reactors. The down selected fuel system consists of U-10Mo alloy fuel foils having a thin Zr diffusion barrier interlayer, clad in 6061 Al alloy by hot isostatic pressing. In order to support fabrication development, fuel performance evaluations, generic fuel qualification through the Nuclear Regulatory Commission (NRC), and eventual reactor licensing, a series of irradiation experiments and testing campaigns are being planned. The tests are designed to provide data needed for fabrication process down selection, to establish acceptable fuel performance under normal operating conditions and anticipated transients, and to demonstrate successful scale-up to prototypic reactor-specific fuel element designs. A comprehensive testing program is being executed as part of the US HPRR fuel development program to develop and qualify a new monolithic UMo fuel system to allow conversion of five high power research reactors and one critical facility within the US in support of the NNSA mission to convert research reactors worldwide from HEU to LEU fuels. A significant outcome of this program will be the qualification and eventual licensing of a new plate-type fuel system for use in research reactors through the NRC for the first time in nearly 3 decades. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Identification and Quantification of Carbon Phases in Conversion Fuel for the Transient Reactor Test Facility

As part of an overall effort to convert US research reactors to low-enriched uranium (LEU) fuel, a LEU conversion fuel is being designed for the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory. TREAT fuel compacts are comprised of UO{sub 2} fuel particles in a graphitic matrix material. In order to refine heat transfer modeling, as well as determine other physical and nuclear characteristics of the fuel, the amount and type of graphite and non-graphite phases within the fuel matrix must be known. In this study, we performed a series of complementary analyses, designed to allow detailed characterization of the graphite and phenolic resin based fuel matrix. Methods included Scanning Electron Microscopy, Raman spectroscopy, and X-ray Diffraction. Our results indicate that no single characterization technique will yield all of the desired information; however, through the use of statistical and empirical data analysis, such as curve fitting, partial least squares regression, volume extrapolation and spectra peak ratios, a degree of certainty for the quantity of each phase can potentially be obtained. Although no one type of analysis has provided a complete determination of the quantities and identities of all carbon phases present in the fuel matrix, additional information has been gathered and is being used to refine future analysis methods. From the three types of quantification performed, results have been within 1% (92.8%, 93.02% and 93.6% for BSE image analysis, XRD and Raman spectroscopy, respectively). More work remains to identify any trace crystalline carbonaceous phases that may be present, quantify any traces that are identified, and develop and use further complementary bulk phase quantification techniques. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

The Impact of Furnace-Testing on the Microstructure of an Irradiated U-Mo Fuel Plate

The Material Management and Minimization Program (formerly known as the Reduced Enrichment for Research and Test Reactor Program) is responsible for converting research reactors that use highly enriched uranium (HEU) fuels to ones that use low-enriched uranium (LEU) fuels. As part of the development of LEU fuels, a variety of irradiation experiments are being conducted using the Advanced Test Reactor. In addition to the reactor tests, there is interest in generating information about the transient behavior of irradiated fuel plates. For example, it is of interest to determine the temperature at which a blister will form on the surface of a fuel plate, due to fission gas release from the fuel meat, for a particular fuel plate design. Since in-pile testing to determine the 'blister threshold' (the temperature where a blister forms on the surface of a fuel plate) in plate-type fuel is very difficult and expensive and has historically been deemed impractical for fuel qualification testing, an out-of-pile 'blister test' has been developed and employed. This post-irradiation blister anneal test was developed as a conservative simulation of in-pile behavior and has been routinely used for many research reactor fuel designs currently in use. In this test, the irradiated fuel plate is inserted into a furnace at a given temperature, soaked for a period of time, and removed for inspection. If no visible blisters appear, the furnace temperature is increased, and the plate is soaked and inspected again. This process continues until blisters are observed. Recently, a section from an irradiated U-7Mo dispersion fuel plate with Al-4043 matrix, which has a nominal composition of 4.81Si-0.20Fe-0.14Ti-0.16Cu-0.01Cr 0.01Mn-bal Al, was tested in a furnace located in the Hot Fuel Examination Facility, located at the Idaho National Laboratory, up to 500 deg. C, at which point blisters were observed on the surface of the fuel plate section. A sample near the blistered region of the fuel plate section was then characterized using scanning electron (SEM) and transmission electron microscopy (TEM). This paper discusses the results of this characterization and how they compare to those generated for a similar sample characterized after fabrication but before irradiation and for a sample characterized after irradiation but before furnace testing. One goal of this work is to identify the changes in the microstructure during blister testing that contribute to the eventual release of fission gas that results in the development of blisters. SEM analysis of the furnace-tested sample showed that in some locations of the U-7Mo particles, Al-4043 matrix constituents (primarily Si) had penetrated the microstructure along grain boundaries. This provided potential pathways for increased fission gas migration in the fuel meat microstructure during the furnace heating process. Some attack of these boundaries was also observed in the as-fabricated U-7Mo/Al-4043 sample, but there appeared to be additional attack during the furnace annealing. Besides the changes observed in grain boundary features, the TEM analysis showed that the fission gas bubbles that are typically present in the intra-granular regions of the as-irradiated U-7Mo microstructure as a fission gas bubble superlattice had changed in size, morphology, and distribution. The result of this may be an increase in the mobility of the fission gas in intra-granular regions of the fuel. The complete implications of the changes in fission gas morphology and distribution in the fuel meat microstructure and their impact on blistering continues to be investigated. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Developing a U.S. High Performance Research Reactor Transportation Road-map - 19674

The U.S. National Nuclear Security Administration Office of Reactor Conversion through the U.S. High Performance Research Reactor (USHPRR) Project is developing and qualifying a new Low Enriched Uranium (LEU) fuel to convert USHPRRs from the use of highly enriched uranium fuel to LEU fuel. The USHPRR Project is forecast to complete in 2030's. During the Project's life cycle, it will make hundreds of radiological shipments. These shipments will include material of different sizes and radiological characteristics and involve many facilities. These shipments must be integrated with the Project's detailed schedule to ensure success of the Project. The USHPRR Project has the complex challenge to develop a Packaging and Transportation (P and T) Road-map to coordinate the multitude of packaging and shipment activities. A comprehensive packaging and shipping matrix is being developed to assist in meeting this challenge and ensure the USHPRR shipments can be made on schedule. The vision for the matrix is for it to be a bridge between the schedule and the P and T Road-map, and for it to be the one stop place for USHPRR shipment information. An interactive approach to the P and T Road-map and the comprehensive shipping matrix provides an effective strategy to execute the USHPPR Project today and through its life cycle. This paper provides an overview of the development of the USHPRR P and T Road-map and the comprehensive shipping matrix and their benefits to the USHPRR Project. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Minimization of Nuclear Waste using AHWR Reactor Technology

Nuclear sciences have grown from creating reactors to a greater understanding of how to dispose off the waste exhibited by the nuclear activities. Various methodologies have been proposed and are being used in order to do the same. This poster poses great attention to 'Using a fuel cycle dedicated to the production of the lower quantity of minor actinides, is one of the key targets in reducing nuclear waste in india'. A significant fraction of thorium is utilized as a fertile host in the Advanced Heavy Water Reactor (AHWR300-LEU) fuel. Being lower in the periodic table, it leads to a reduction of minor actinides. It produces only half of the minor actinides as a waste as compared to modern LWRs. As thorium oxide has inert nature, it benefits in its prolonged storage. Therefore, safe disposal of spent fuel is achieved due to the inert matrix, on completion of one fuel cycle mode. AHWR is a 300 MWe, vertical, pressure-tube type, boiling light water-cooled, and heavy water-moderated reactor. Employs natural circulation for removal of heat from the reactor core. In-situ conversion of {sup 232}Th into {sup 233}U. The reactor design permits flexible fuel cycle options. There are two types of AHWR depending upon the fuel cycle options. Uses thorium-based oxide fuels: two fuel cycle options. With the adoption of thorium fuel cycle, AHWR300-LEU produces less minor actinides. As compared to modern LWRSs, it produces almost 37% less amount of actinides for the same amount of energy production, which in turn minimizes the burden on waste disposal requirements. But, it is associated with radionuclides such as {sup 231}Pa (produced through absorption of one neutron and emission of two neutrons). Since, the reaction rate of (n,2n) reactions is very low in thermal neutron reactors, thus making the fuel cycle the favorable one. Thorium oxide (ThO{sub 2}) is chemically stable and is the highest oxidation state of thorium, therefore the inert nature of the matrix is advantageous for safe disposal of spent fuel in case of adoption of open fuel cycle mode and offers prolonged storage of spent fuel. The AHWR is thorium fueled Indian nuclear reactor that assimilates several new features with diminished environmental concern due to reduced waste generation. Since thorium fuel cycle produces less long lived nuclear waste, thus it can possibly bring the world cleaner and more sustainable nuclear energy. Currently, India being the only country with large scale research program of thorium nuclear power serves a platform for the development and demonstration of all technologies and advanced safety systems.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Analysis of Using a Hybrid ZIRCEX Process to Generate High-Assay Low-Enriched Uranium (HALEU) - 20158

Idaho National Lab (INL) is investigating the feasibility of providing High-Assay Low Enriched Uranium (HALEU) by recovering Highly Enriched Uranium (HEU) from spent nuclear fuel (SNF) and downblending the HEU with natural, low-enriched, or depleted uranium (NU, LEU, or DU). That investigation identified that significant quantities of HALEU could be produced with a Hybrid ZIRCEX process to further advanced reactor research while reducing INL's environmental liability. Many new advanced reactor designs are being developed with improved safety, efficiency, and economics. Most require nuclear fuel with U-235 enrichment between 5% and 20%, which is defined as HALEU. Although technically feasible, there is no current domestic capability to make HALEU in the US. While it is anticipated that industry will provide HALEU through commercial enrichment once advanced reactors mature, until a market is established, an interim source of HALEU is needed to enable research and demonstration. An interim source of HALEU could be provided by recovering the HEU in some DOE-managed SNF and downblending it to between 5% and 20% using the Hybrid ZIRCEX Process. ZIRCEX is a dry head-end process to remove cladding (zirconium or aluminum) from SNF. Drying of the SNF is needed prior to introduction to ZIRCEX. After cladding removal, uranium and fission products in the bed material are oxidized and elutriated. In a Hybrid ZIRCEX process, the HEU process stream is sent to a very compact, modular solvent extraction process for uranium purification. The uranyl nitrate from solvent extraction is denitrated and calcined to produce a HALEU product ready for fuel fabrication. The long-lived fission products from solvent extraction are immobilized in glass using a small in-can melt. By removing the cladding before dissolution, the volume of glass waste is reduced by a factor of up to 300 times (when compared to not removing the cladding from the fuel). Other radioactive wastes from the process could be disposed of as Class A or Class B low-level waste. In addition to producing HALEU for advanced reactors, consumption of SNF would reduce DoE's environmental liability. Incorporating low-enriched spent fuel (versus exclusive use of LEU, DU, or NU) into the downblending scheme could reduce the SNF inventory even further. Several downblending and enrichment scenarios using INL fuels were investigated in this study. The results of this analysis show that the Hybrid ZIRCEX process could make a significant contribution to needed HALEU feedstock and reduce the environmental liability of managing SNF at INL. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Initial Shipment of MNSR Reactor HEU Spent Nuclear Fuel from Ghana to China

The repatriation of Russian and US origin fresh and spent, high enriched uranium (HEU) research reactor fuel has been conducted since 1996. In 2006, efforts were initiated to convert Chinese designed Miniature Neutron Source Reactors (MNSRs) from HEU to low enriched uranium (LEU) fuel. The Ghana Research Reactor-1 (GHARR-1) is the first of five such MNSR reactors outside of China eligible for conversion and fuel return to China. Under a project involving China and Ghana as operating country, as well as the U.S.A and the IAEA the HEU core is to be repatriated to China and a new LEU core is to be installed. The present paper discusses the preparation of the HEU core for removal, describing the logistics of the shipment of the HEU core and the lessons learned which can be applied to the remaining MNSR shipments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

A Theoretical Investigation of the Chemical Behaviour of Irradiated U-Mo/Mg Fuel

A significant international effort has been put forth to replace Highly Enriched Uranium (HEU) fuel used in research and test reactors with Low Enriched Uranium (LEU) fuel to reduce proliferation concerns. Furthermore, some fuel processing issues have been encountered with some LEU candidate fuel designs. In order to address both of the foregoing issues, new fuel designs must be developed. One of these designs has U-Mo particles dispersed in a Mg matrix (U-Mo/Mg) that is cladded in an Al-alloy. While characterization analyses and thermophysical measurements have been reported for fresh U-Mo/Mg fuel, the effects of irradiation are not yet well known nor are they understood. Given that earlier international campaigns that investigated U-Mo/Al as a candidate fuel design resulted in consistent failures due to the formation of an interaction layer between the U-Mo fuel particles and Al meat, it seems prudent to theorize and predict any phases that may form as a result of irradiation of U-Mo/Mg fuel. The formation of these phases and any interaction that they may have with the cladding may potentially affect the reliability of this fuel design. This short paper provides a preliminary theoretical framework for the prediction of minor phases that may appear in irradiated U-Mo/Mg research reactor fuel, which should be confirmed via postirradiation examination at a later time.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

University of Wisconsin Nuclear Reactor Modeling Improvements

The University of Wisconsin Nuclear Reactor converted from HEU to LEU fuel in 2009. As part of the conversion process, an MCNP model was developed to predict flux distributions and depletion in the core. This core model included an unforeseen excess reactivity bias, resulting in an overestimate of core reactivity, and requiring the core configuration to be shuffled in 2011. The original model was constructed by hand and in order to account for radial and axial burnup in the fuel, and to include sufficient detail in the geometric features, the input file was over 17,000 lines long. Given the challenges in assessing the quality of such a long and complicated MCNP model, this project was designed to build an MCNP model based on a minimal set of input quantities in a modular and repeatable fashion. A Python code was implemented to write the MCNP input file from fundamental reactor data, both geometric and material. The recent conversion to LEU fuel provides a unique opportunity to characterize fuel composition quite accurately in the core. The goal of this project is to produce an accurate, maintainable model that can be used with future reactor physics experiments.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Approach to nuclear criticality in IAN-R1 research reactor

The Republic of Colombia is operating a small 100 kW(t) research reactor. The reactor was provided to Colombia under the U.S. Atoms for Peace Program, and which was fueled with MTR HEU fuel enriched nominally to 93% U{sup 235}. With the cooperation of the International Atomic Energy Agency IAEA, it was prepared a new safety analysis report for performing an HEU to LEU conversion of the IAN-R1 reactor and manufacture TRIGA type LEU (19.7%, enriched) fuel to replace the original MTR-HEU fuel plate assemblies. This paper describes the activities which were achieved during the approach to nuclear criticality. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

The Case for and Against a Gadolinium Bias in SCALE: Round 2

The “Opening Arguments” for and against a gadolinium bias in SCALE were presented at the American Nuclear Society Annual Meeting in Philadelphia, Pennsylvania, in June, 2018. Some critical experiments included in the Oak Ridge National Laboratory Verified, Archived Library of Inputs and Data (VALID) indicate a significant bias as a function of gadolinium concentration. Other experiments indicate that no significant bias exists. The work presented here develops a larger suite of gadolinium-bearing benchmark models to further examine code, data, and benchmark performance. The new benchmark models have been reviewed for accuracy, but documentation and review for addition to the VALID library have not been completed. The problematic benchmarks included in VALID are HEU-SOL-THERM-014 and -016. These are two evaluations from a series of experiments from the Institute for Physics and Power Engineering (IPPE), Russia, documented in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook. The entire set of evaluations also includes HEU-SOL-THERM-015, -017, -018, -019, and -025. Each evaluation contains a different uranium concentration, and different configurations within each evaluation include different gadolinium concentrations. These 7 evaluations contain a total of 52 configurations and form the largest subset of experiments considered, and they allow for a more complete assessment of the performance of these benchmarks than has historically been possible using just the HEU-SOL-THERM-014 and -016 results. Additional solution experiments are considered, including MIX-SOL-THERM-006 and -007 and PU-SOL-THERM-034. MIX-SOL-THERM-007 is in the VALID library, whereas MIX-SOL-THERM-006 and PU-SOL-THERM-034 are not. The results for MIX-SOL THERM-007 have not shown a gadolinium trend. These mixed- and plutonium-fueled solutions include 28 configurations. Some experiments including solid fuel and solid gadolinium are also included. These experiments include highly enriched uranium (HEU) foils moderated with polyethylene in HEU-MET-THERM-010, -016, and -034, and low enriched uranium (LEU) pin arrays with gadolinia absorber rods in LEU-COMP THERM-036 and -043. A total 32 cases with solid fuel are included. The IPPE solution benchmarks show a fairly high degree of variability, but no clear trend as a function of gadolinium concentration can be observed. The mixed- and plutonium-fueled solutions show less variability than the HEU solutions and also no trend relative to gadolinium concentration. The solid-fueled experiments also show no trend as a function of gadolinium content. The entire set of benchmarks shows no clear trend on gadolinium content or the energy of the average neutron lethargy causing fission.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Completion of Critical Experiments with Molybdenum Sleeves at Sandia

Sandia National Laboratories (SNL) and the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) have collaborated on the design and execution of a set of critical experiments that explore the effects of molybdenum in water moderated fuel-rod arrays. The molybdenum is included as sleeves (tubes) on some of the fuel rods in the arrays. The fuel used in the experiments is known at Sandia as the Seven Percent Critical Experiment (7uPCX) fuel. This fuel has been used is several published benchmark evaluations in including LEU-COMP-THERM-78 and LEU-COMP THERM-080.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

International MNSR Training Center

Plans for the conversion of the Chinese origin miniature neutron source reactors (MNSRs) to low-enriched uranium (LEU), and the return of the high- enriched uranium (HEU) core to China, were initiated in February 2011 when an international MNSR Working Group (WG) was established to coordinate these activities. Based on a Ghana Atomic Energy Commission (GAEC) decision, the MNSR core at the GHARR-1 facility was selected for pilot shipment. Working on the preparation of the HEU MNSR materials’ removal from the GHARR-1 facility as a training tool, a mock-up MNSR vessel was built for operator training. Experiencing the usefulness of the mock-up vessel in operator training in the Ghana case has since been developed further into a full-scale training facility, named the International MNSR Training Center (IMTC), which is now available for training operators from MNSR operating countries. The main objective of this Book is to document the IMTC as a full-scale MNSR training facility in operator training for HEU MNSR core discharge and packaging. The goal is to describe the IMTC in its full complexity — that is: (1) to present the IMTC’s buildings with its internal and outer areas; (2) to provide a full listing of the installed equipment, as well as a functional description, available at the IMTC site; (3) to introduce the training materials, tutorials, and training program; and (4) to present the training sessions and Dry-run demonstration of the recently completed NIRR-1 operator training, with their experiences and lessons-learned included. This Book serves as a kind of user manual for the IMTC, which contains all of the information to launch training courses at the site at any time. The target audience for this Book is the MNSR community, especially MNSR operators and regulators who are facing HEU-LEU core conversion and HEU material transportation, who require operator training to carry out their jobs safely and professionally.

99 - GENERAL AND MISCELLANEOUS↗
DOE OSTI2025

White Paper: Research & Development for the Time at Temperature Approach

Recent advancements in nuclear power research are greatly improving reactor safety and performance through the development of Accident Tolerant Fuel (ATF) and Low-Enriched Uranium Plus (LEU+). These innovations can address Departure from Nucleate Boiling (DNB) margins, which are vital for reactor safety. DNB happens when the coolant switches to film boiling, significantly decreasing heat transfer and posing a risk of fuel cladding failure. The U.S. Nuclear Regulatory Commission (NRC) employs conservative DNB criteria, which can potentially restrict the operational flexibility and efficiency of reactors. The Time at Temperature (TaT) approach could provide a more detailed and adaptable operational guideline by establishing acceptable time-temperature limits, accounting for the duration a material can withstand elevated temperatures without losing its integrity. This method allows reactors to operate more efficiently and safely, offering additional operational margins, faster power adjustments, and improved fuel cycle economics. TaT criteria allow for higher power levels and more flexible responses to operational transients, particularly applicable for anticipated operational occurrences (AOOs) that result in short durations of post-DNB conditions. It enhances plant operational flexibility, allows faster startup times, and enables quicker power level adjustments, optimizing fuel loading patterns and improving fuel cycle economics. Implementing TaT limits reduces core design constraints, lowers fuel usage, and reduces costs, essential for the long-term sustainability of Light Water Reactors (LWRs). TaT maximizes the use of advanced fuel technologies like ATF and LEU+, further enhancing their economic and environmental benefits. To apply the TaT approach in existing LWRs, collaborative research activities among various DOE-sponsored programs are essential. These efforts should incorporate fuel experiments, physics-based high-fidelity modeling, ML-based surrogate modeling, and optimization techniques. This whitepaper proposes four research and development areas: 1) Investigation of the feasibility of new operations of LWR with updated safety limits; 2) Assessment of reactor operation limits through uncertainty reduction; 3) Evaluation of power uprate in virtual environment; and 4) Lattice and reactor core design for power uprate. Each area includes why this research is in need and a suggested scope of work. These comprehensive research areas ensure practical and beneficial advancements for existing reactors, translating innovations in nuclear fuel and cladding technology into improved reactor performance and safety.

42 - ENGINEERING↗
DOE OSTI2019

Current Status of the Russian Research Reactor Fuel Return Program

In 1999, the U.S. Department of Energy (DOE), together with the International Atomic Energy Agency (IAEA) and the Russian Federation, agreed to start a new nuclear non-proliferation initiative to return Russian/Soviet-origin highly enriched uranium (HEU) back to its country of origin. Trilateral discussions among the U.S., the Russian Federation, and the IAEA identified more than 20 research reactors in 16 countries (Belarus, Bulgaria, China, Czech Republic, DPRK, Germany, Hungary, Kazakhstan, Latvia, Libya, Poland, Romania, Serbia, Ukraine, Uzbekistan, and Vietnam) that had Russian/Soviet-origin HEU. In 2000, the IAEA Director General sent a letter to 16 countries asking for their willingness to return this spent HEU fuel to the Russian Federation. Fourteen countries responded positively to the Director General’s letter. The primary goal of the Russian Research Reactor Fuel Return (RRRFR) program, managed by DOE’s National Nuclear Security Administration, Office of Material Management and Minimization, is to advance nuclear nonproliferation objectives by encouraging eligible countries to convert their research reactors from HEU to low-enriched uranium (LEU) fuel and to eliminate stockpiles of HEU. In May 2004, the United States and the Russian Federation signed a Government-to-Government Agreement that established the legal framework necessary for cooperation between the United States and the Russian Federation for the return of research/test reactor fuel from eligible countries. To date, the RRRFR program has completed sixty-three successful operations to repatriate almost 2,300 kg of Russian-origin HEU fresh and spent fuel. All Russian-origin HEU has been completely removed from 12 countries: Bulgaria, the Czech Republic, Hungary, Georgia, Latvia, Libya, Poland, Romania, Serbia, Ukraine, Uzbekistan, and Vietnam. At the present time Russian-origin HEU remains in Kazakhstan, Belarus, Germany, DPRK, and China. This paper describes the current status of the RRRFR program and preparations for upcoming HEU removal and down-blending activities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Thermomechanical analysis and modeling of involute-shaped fuel plates using the Cheverton–Kelley experiments for the High Flux Isotope Reactor

Three research reactors with involute-shaped fuel plates are pursuing conversion from highly enriched uranium to low-enriched uranium fuel. Various core design and safety evaluation studies are essential to assess the feasibility of the conversion. The use of 3D computational multiphysics codes is being explored in these analyses and therefore they must undergo a thorough evaluation and quality assurance process due to their potential impact on nuclear safety. Here, the Cheverton and Kelley physical tests performed in the late 1960s to investigate the deflections of HFIR’s outer plate under uniform pressure and temperature fields are simulated by employing commercially available computational codes, with the goals to (1) verify and validate the models and numerical solvers implemented in the codes for thermomechanical analysis of involute reactor plates and (2) to develop a benchmark computational test to evaluate future versions of existing software or newly developed computational codes. The results of the simulations showed good agreement with each other as well as against the Cheverton–Kelley experimental data. Some minor deviations were observed for a few multiphysics cases and their potential origins and impact on the analysis results is investigated in the paper. The validated models increase the confidence in using multiphysics codes to evaluate existing or new LEU designs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗