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At least 667 records · Page 37
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Progress in demonstrating the performance of Engineered Barriers for the disposal of HLW and SNF: lab studies and demonstration experiments - 20293

Disposal of high-level radioactive waste (HLW) and spent nuclear fuels (SNF) in a geological disposal facility requires a fundamental understanding of the degradation processes of materials used in its engineered barrier system (EBS) over very long timescales. Beyond the wasteform itself (i.e. the borosilicate glass for HLW or the fuel matrix and cladding for SNF), materials often employed in a disposal concept include metallic containers (typically copper and/or carbon steel) and clay-based buffer materials (typically bentonite). In this paper we explore key developments on the long-term evolution of materials employed for the immobilisation, isolation and containment of HLW and SNF in a geological disposal facility, including nuclear glasses, candidate container materials and clay buffers. Our discussion is based on the latest findings from both experimental studies and in-situ demonstration experiments. We present relevant findings from the long-term evolution of HLW glasses (UK composition), metallic materials for the containment system (envisaged in Switzerland, Sweden, Canada and other countries), and bentonite clays (envisaged in many national disposal programmes). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Preliminary Assessment of Alternate Chlorination Process

Recent advancements towards low-temperature chlorination with disulfur dichloride and thionyl chloride of Al-clad used nuclear fuel (UNF) have been summarized for work control technical expertise development at the Savannah River National Laboratory (SRNL). This technique aims to provide a more effective method for separating U from cladding or alloying elements. A preliminary assessment determined needs for unit operation developments in regard to reaction kinetics and reagent quality. Furthermore, an analysis of an Uruguay fuel plate (U-Al x fuel meat), with low-enriched uranium (LEU) and located at SRNL, was a potential candidate for future experiments with lightly irradiated (0.08% burnup) fuel to test for chlorination. Future proposed work would involve kinetic studies, reagent quality assessments, and the machining of the fuel plate down to an appropriate bench-scale size and testing the effectiveness of low-temperature chlorination on the fuel plate.

Chlorination↗
DOE OSTI2022

Post-irradiation Examination Status for the BSU-269 Project

High-density uranium inter-metallic uranium silicide (U 3 Si 2 ) was initially considered as a potential advanced nuclear fuel for light water reactor (LWR) systems due to several advantages it demonstrated in thermophysical properties, chemical, and irradiation stability. In addition to understanding the basic thermophysical and mechanical properties of U 3 Si 2 under LWR irradiation conditions, it is important to assess its performance with advanced claddings such as FeCrAl, SiC, or conventional Zr-based claddings. Several aspects of the fuel cycle and fuel performance are being evaluated for U 3 Si 2 coupled with different types of claddings. To thoroughly characterize the U 3 Si 2 -cladding system, their chemical compatibilities and the potential for fuel-cladding chemical interaction (FCCI) must be addressed. Interdiffusion studies have been performed between U 3 Si 2 /Zry-4 and U 3 Si 2 /FeCrAl using diffusion couples. For the U 3 Si 2 /Zry-4 system, ZrSi 2 was the primary interdiffusion product observed at 800°C in addition to other secondary phases from the Zr alloying elements such as Fe and Cr. Low melting point U 6 Fe was also observed at 1000°C. For the U 3 Si 2 /FeCrAl system, at temperatures above 500°C, chemical diffusion Fe-U and Fe-U-Si was observed, increasing with rising temperatures. Although information from these non-irradiated tests is a fundamental first assessment, follow-up with focused irradiation tests is necessary. Irradiation can significantly alter the structure and evolution of the FCCI layers compared to the non-irradiated conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Recent High Burnup LOCA Testing at Oak Ridge National Laboratory

To improve fuel cycle economics, increasing the fuel burnup limit in light-water reactors requires a solid technical foundation. Observations from experiments at Halden and Studsvik highlighted severe fuel fragmentation during loss-of-coolant-accident (LOCA) conditions, indicating a need for further technical considerations. These experiments suggest that the fragmentation threshold for high-burnup fuel might be influenced by pre-transient power levels. Consequently, additional LOCA test data are essential to complement existing findings and to deepen our understanding. Oak Ridge National Laboratory’s Severe Accident Test Station has been instrumental in advancing knowledge about high-burnup fuel fragmentation, relocation, and dispersal. This milestone report details two high-burnup LOCA tests designed to evaluate the effects of terminal temperature and grid spacers (or cladding restraints) on fuel fragmentation and relocation susceptibility. In addition, BISON fuel performance modeling and out-of-cell benchmark testing were conducted to better interpret the in-cell test results. These tests were developed in collaboration with fuel vendors to ensure that the results provide valuable data for topical reports and support the Nuclear Regulatory Commission’s review.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Fatigue Test Result for High Burnup Pressurized Water Reactor Rods

The High Burnup Spent Fuel Data Project, sponsored by DOE-NE, is focused on understanding the effects of long-term storage and transportation on high burnup (HBU) (>45 gigawatt days per metric ton uranium) light water reactor fuel. The experiment includes 32 HBU spent nuclear fuel (SNF) assemblies that will be stored in a typical Independent Spent Fuel Storage Installation (ISFSI) and 25 sister rods - nine SNF rods that were removed from the fuel assemblies prior to insertion to the ISFSI and 16 SNF rods removed from similar HBU assemblies. The sister rods provide a baseline of the condition of the HBU rods prior to loading, drying, and long-term dry storage. Destructive examinations of the sister rods are focused on understanding overall SNF rod strength and durability. Both composite fuel and empty cladding will be tested to derive material properties. While the data generated can be used for multiple purposes, a primary goal of the sister rod PIE data and associated measured mechanical properties is to support existing fuel storage licensing and relicensing activities by addressing identified knowledge gaps, and enhance the technical basis for post-storage transportation, handling, and consolidation activities. Fatigue tests of HBU M5-clad, ZIRLO-clad, Zircaloy-4, and low tin Zircaloy-4 segments have been completed using ORNL’s Cyclic Integral Reversible Fatigue Tester (CIRFT). This is the first time ZIRLO-clad rods have been tested. Some of the segments were heat-treated to understand the effects of thermal transients such as those encountered during dry storage vacuum drying on the fatigue lifetime. This paper presents results of the fatigue tests obtained to date and discusses planned future tests.

Montgomery, Rose↗
DOE OSTI2024

Fuel performance analysis of Cr-coated Zircaloy-4 cladding during a prototypical LOCA event using BISON

Here, deformation and failure of chromium (Cr) coated Zircaloy-4 (Zry-4) were studied in loss-of-coolant accident (LOCA) conditions using the BISON fuel performance code. The BISON validation model simulating Halden research reactor experiments was extended to include Cr coatings and higher rod internal pressures to simulate high-burnup fuel. The transient simulations show Cr coatings help relieve stress in the Zry-4 substrate during the transient, delaying the onset of high-temperature creep, which leads to ballooning and bursting of the cladding. A nominal Cr coating thickness of 30 µm delays clad failure by 26 seconds and increases the clad burst temperature by 40 K. A parametric study showed that time to failure and burst temperature both increase with coating thickness, and Cr-coated cladding offers burst resistance under a wide range of rod internal pressures simulating high-burnup fuel. Results indicate that a thin Cr coating provides resistance against ballooning and bursting of cladding during LOCA events.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

A layered 2D computational framework: Theory and applications to nuclear fuel behavior

Nuclear fuel performance computer codes have been developed over the last 50 years to analyze fuel behavior under various operating conditions. Traditionally, these codes used quasi-two-dimensional (also commonly known as 1.5D) representations of the fuel rod, which model the rod using a set of one-dimensional axisymmetric models that represent the behavior at specific axial positions on the rod. Modern fuel performance codes have the ability to investigate full three-dimensional (3D) effects and couple to other physics-based codes for true multiphysics simulations. However, with increasing complexity comes increasing computational costs. Many phenomena of interest involve azimuthally-varying behavior that cannot be represented using the aforementioned quasi-two-dimensional approach, but do not require the use of a full 3D model. To efficiently address these problems, there is a need for a computational framework that provides a compromise between the quasi-two-dimensional and full 3D models. In this paper, we present a new quasi-three-dimensional approach that represents the fuel as a set of 2D planar models that represent the behavior of the fuel cross-section at various axial positions. Presented here are the theory behind the methodology, test cases to illustrate proper implementation, and practical applications of its use in the BISON fuel performance code for a variety of cases in nuclear fuel analysis, such as fuel fracture, axial fuel relocation, and cladding distension and oxidation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Thermal Modelling of Advanced Test Reactor Fuel in a Generalized Dry Storage System

Star-CCM+, a computational fluid dynamics (CFD) software was used to conduct modeling and simulation of the thermal performance of a dry storage configuration consisting of Department of Energy Standardized Canisters (DOESCs) loaded with aluminum-clad spent nuclear fuel (ASNF.) The configuration includes nine DOESCs loaded with Advanced Test Reactor ASNF contained within a stainless steel overcanister centered in a ventilated, concrete overpack. The simulations were used to estimate the maximum temperatures reached by backfill gases inside the overcanister and DOESCs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Computational Modeling of Multi-Pass Rolling Parameters Effect on Resulting Fuel Foil Shape

A focus of the U.S. Department of Energy is to improve production yield and reduce the cost of Low Enriched Uranium (LEU)-molybdenum alloy (U-10Mo) monolithic fuel plates that will be replacing High Enriched Uranium (HEU) oxide dispersion fuels used currently in the United States High Performance Research Reactors (USHPRR). One area of improvement is lowering the high transverse waviness and longitudinal waviness currently present within rolled foils prior to cladding to produce fuel plates. Traditional rolling manufacturing techniques for other metal foils use winders to pull and straighten the foil as it is rolled back and forth to the final thickness. This approach cannot be used to roll thin U-10Mo foils (0.008-0.025” thick) because only small castings can be rolled due to nuclear criticality safety concerns. As a result, the fuel foils are too short (1 m in length) to use traditional winders. Therefore, it is crucial to identify other rolling parameters (i.e., roller friction, axial tension load, roller diameter, and roll pass reduction percent) that might reduce transverse waviness and longitudinal waviness in the rolled fuel foil and develop a high-yield, low-cost multi-pass rolling manufacturing process. This report documents a systematic finite element modeling study to investigate the effects of numerous rolling parameters to reduce resulting transverse waviness and longitudinal waviness in the fuel foil during multi-pass rolling of U-10Mo foils. The rolling of a U-10Mo plate with initial dimensions of 1”x1”x 0.048” is modeled using Abaqus CAE. This rolling is modeled to undergo eight 20% reduction roll passes to a final fuel foil thickness of 0.01”. The elastic-plastic constitutive model of the U-10Mo alloy was input to the fuel foil rolling model. The rollers were modeled as rigid bodies. A comparison of rolling friction coefficients of 0.3 and 0.7 over a wide range of applied axial tension loads were investigated in order to evaluate the effect of using a lubricant during rolling. The effect of roller diameter on the resulting transverse waviness and longitudinal waviness of the fuel foil over a wide range of axial tension loads were also investigated by modeling rollers 7/8” and 3.75” in diameter. The results of this systematic finite element method study will aid manufacturers in producing low transverse waviness and reduced longitudinal waviness in U-10Mo fuel foils.

U-10Mo, FEA, Rolling, Residual Stress, Fuel Foil↗
DOE OSTI2021

Assessment of the BISON Metallic Fuel Performance Models

The US Department of Energy is leading a project to design and construct a fast spectrum test reactor called the Versatile Test Reactor (VTR). The BISON nuclear fuel performance code will be used to model VTR driver fuel, including looking at the effects of differences between the VTR driver fuel element design and the legacy fuel designs and experiments on which it is based. Simulations will be conducted to help determine whether the design’s behavior and performance are properly understood and to assess the margins to cladding failure and fuel melting relative to those predicted for past metallic fuel experiments. These predictions are expected to streamline VTR design and operation by helping inform the VTR driver fuel element design and by providing supplemental information for the fuel design safety basis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Challenges and opportunities to alloyed and composite fuel architectures to mitigate high uranium density fuel oxidation: uranium silicide

We report the challenges and opportunities to alloyed and composite fuel architectures designed and intended to mitigate oxidation of the fuel during a cladding breech of a water-cooled reactor are discussed in three review manuscripts developed in parallel, with the presented article focused on the oxidation performance of uranium silicide. Several high uranium density fuels are under consideration for deployment as accident tolerant and/or advanced technology nuclear reactor fuels, including UN, U 3 Si 2 , UC and UB 2 . Presented here are the literature for the U3Si2 degradation modes, thermodynamics, and oxidation performance of the pure compound and its reported alloyed and composite architectures. Furthermore, this review covers the materials and techniques for the incorporation of additives, dopants, or composite fuel architectures to improve the oxidation/corrosion behavior for high uranium density fuels for use in LWRs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Challenges and opportunities to alloyed and composite fuel architectures to mitigate high uranium density fuel oxidation: Uranium mononitride

We see the challenges and opportunities to alloyed and composite fuel architectures designed and intended to mitigate oxidation of the fuel during a cladding breech of a water-cooled reactor are discussed in three review manuscripts developed in parallel, with the presented article focused on the oxidation performance of uranium mononitride. Several high uranium density fuels are under consideration for deployment as accident tolerant and/or advanced technology nuclear reactor fuels, including one on each: UN, U 3 Si 2 , UC and UB 2 . Presented here is the research motivation for the incorporation of additives, dopants, or composite fuel architectures to improve the oxidation/corrosion behavior of high uranium density nuclear fuels for use in LWRs. Furthermore, this review covers the literature on the degradation modes, thermodynamics, and oxidation performance of pure UN and UN-compounds as well as reported alloyed and composite architectures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

FFTF HT9 Cladding Microstructure Characterization

The sodium-cooled fast reactor (SFR) is a promising candidate for next generation nuclear reactors, operating at extreme conditions which include high temperatures (>500?C core outlet temperature) and significant neutron damage. High-Cr martensitic HT9 steel is an excellent candidate for SFR cladding and duct material due to its compatibility with liquid sodium, good thermal conductivity, resistance to void swelling, and strong creep rupture strength [1-4].However, the harsh in-core environment of SFRs can cause complex microstructural changes and mechanical property degradation in HT-9. Ensuring the safe use of HT9 cladding for metallic fuel requires both a thorough understanding of its mechanical response to microstructure evolution as well as reliable microstructure-sensitive modeling predictions. Microstructure-sensitive modeling of high temperature creep behavior in HT9 cladding for SFR applications currently lack experimental data to model the phenomena accurately. To fill this need, methods to perform microstructural characterization have been developed and performed on HT9.

36 MATERIALS SCIENCE↗
NASA NTRS2005

SOLID SOLUTION CARBIDES ARE THE KEY FUELS FOR FUTURE NUCLEAR THERMAL PROPULSION

Nuclear thermal propulsion uses nuclear energy to directly heat a propellant (such as liquid hydrogen) to generate thrust for space transportation. In the 1960 s, the early Rover/Nuclear Engine for Rocket Propulsion Application (NERVA) program showed very encouraging test results for space nuclear propulsion but, in recent years, fuel research has been dismal. With NASA s renewed interest in long-term space exploration, fuel researchers are now revisiting the RoverMERVA findings, which indicated several problems with such fuels (such as erosion, chemical reaction of the fuel with propellant, fuel cracking, and cladding issues) that must be addressed. It is also well known that the higher the temperature reached by a propellant, the larger the thrust generated from the same weight of propellant. Better use of fuel and propellant requires development of fuels capable of reaching very high temperatures. Carbides have the highest melting points of any known material. Efforts are underway to develop carbide mixtures and solid solutions that contain uranium carbide, in order to achieve very high fuel temperatures. Binary solid solution carbides (U, Zr)C have proven to be very effective in this regard. Ternary carbides such as (U, Zr, X) carbides (where X represents Nb, Ta, W, and Hf) also hold great promise as fuel material, since the carbide mixtures in solid solution generate a very hard and tough compact material. This paper highlights past experience with early fuel materials and bi-carbides, technical problems associated with consolidation of the ingredients, and current techniques being developed to consolidate ternary carbides as fuel materials.

Panda, Binayak↗
DOE OSTI2022

Challenges and opportunities to alloyed and composite fuel architectures to mitigate high uranium density fuel oxidation: Uranium diboride and uranium carbide

Here, the challenges and opportunities to alloyed and composite fuel architectures designed and intended to mitigate oxidation of the fuel during a cladding breach of a water-cooled reactor are discussed in three review manuscripts developed in parallel, with the presented article focused on the oxidation performance of uranium diboride and uranium monocarbide. Several high uranium density fuels are under consideration for deployment as accident tolerant and/or advanced technology nuclear reactor fuels, including UN, U 3 Si 2 , UB 2 , and UC. Presented here is the literature for UB 2 and UC degradation modes, thermodynamics, and oxidation performance of the pure compounds and reported alloyed and composite architectures. Furthermore, this review covers the materials and techniques for the incorporation of additives, dopants, or composite fuel architectures to improve the oxidation behavior for high uranium density fuels for use in LWRs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Catalytic Effects of Silver in Iodine Reactors for Dissolved Used Nuclear Fuel

The dissolution of used nuclear fuel generates a variety of off-gasses including flammable hydrogen and other species that are a concern for environmental release. The H-Canyon facility at the Savannah River Site is currently dissolving aluminum-clad research reactor fuel from material test reactors and the High Flux Isotope Reactor (HFIR) using a mercury-catalyzed nitric acid flowsheet. Savannah River National Laboratory recently developed and deployed a Raman spectrometer to monitor the off-gas stream from the dissolution process. Results from these measurements indicated a lack of the expected hydrogen, nitrous oxide, and nitric oxide in the off-gas stream. It was proposed that the silver on the silver nitrate–coated berl saddles present in the reactors for iodine capture were acting as a catalytic hydrogen recombiner. Nitric oxide is readily oxidized to nitrogen dioxide under normal conditions, but it was unclear what happened to the nitrous oxide. A laboratory-scale iodine reactor was assembled and filled with silver nitrate–coated berl saddles to help ascertain the fate of nitrous oxide and hydrogen. Testing with this laboratory-scale reactor observed the recombination of hydrogen when a simulated dissolver off-gas was passed through the reactor containing silver nitrate–coated berl saddles at the approximate temperatures seen in H-Canyon. However, the nitrous oxide concentration was unchanged, suggesting a more complex process occurring within the off-gas stream before it reaches the iodine reactors at H-Canyon.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Metallic Coating of Cerium Oxide Microspheres

The ability to remove heat is paramount to nuclear fuel performance and longevity. Retaining fission product and separating fuel from reactor coolant and the environment is also necessary to prevent radiological contamination. Conventional nuclear fuel for commercial light water reactors and radioisotope power systems (RPS) is composed of oxide powders pressed into a pellet (cm-scale) and then sealed into a metal cladding to confine the fuel. What typical fuels lack is a method to surround each particle of nuclear fuel in metal, thus providing a more intimate protection layer for accident tolerance and boosting the thermal extraction from the fuel element. In such a way, metal-coated fuel particles increase heat extraction efficiency over clad-pellet designs while increasing the accident tolerance of the fuel. Metal oxide microspheres have wide-ranging applications, including the realm of fuels for nuclear reactors and RPS. Microspheres of uranium oxide/uranium carbide, mixed uranium/plutonium oxides, transuranics, and thorium fuels have been extensively studied. Pacific Northwest National Laboratory has also demonstrated the production of 238 PuO 2 microspheres for RPS applications. Metal-coated oxide microsphere fuels may also be attractive for other applications such as nuclear thermal rockets, future nuclear reactor designs, and catalysts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Bench-Scale Electrolytic Dissolution of Quarter-Scale FCA Cans

In 2016, the Savannah River National Laboratory (SRNL) led, in support of and under sponsorship of the Department of Energy’s National Nuclear Security Administration (DOE/NNSA) Office of Material Management and Minimization (M3), the removal and transfer of the plutonium based Fast Critical Assembly (FCA) fuel from the Japan Atomic Energy Agency (JAEA) Tokai facility to the Savannah River Site (SRS). The team also included JAEA, multiple organizations in Savannah River Nuclear Solutions (SRNS), International Nuclear Services, and many other entities. The FCA fuel removal project completion was a key deliverable for M3 to the 2016 Nuclear Security Summit and constituted the largest inventory of weapons-usable plutonium removed under the nonproliferation program. The FCA materials consist of thousands of stainless steel (SS) clad plates and hundreds of SS clad rods. The FCA fuel elements were packaged in a carrier can and stored at SRS pending disposition of the fuel. Following an assessment of candidate disposition options, SRNS identified electrolytic dissolution (ED) as the most promising disposition option for the FCA plates and their preferred option was endorsed by DOE. This option entails electrochemically dissolving the entire FCA carrier can with fuel elements and was based on bench-scale laboratory testing and historical work on processing SS-clad and zirconium-clad uranium-based fuel in the H-Canyon electrolytic dissolver (last operated in 1980). The FCA plate consists of a plutonium-aluminum metal alloy core hermitically sealed in SS cladding.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗