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At least 235 records · Page 13

Coupled Bison/Thermochimica Simulation of Hydride Formation in Zircalloy Cladding

The report describes implementation of the thermodynamics models for simulation of zirconium hydride formation in LWR fuel cladding. The current models for ZrH x formation are based on empirical rate models that must calibrated to materials and regimes of interest. The thermodynamic approach enables hydride formation calculations based on fundamental thermodynamics principles and material composition. Thermodynamic formulations and models were implemented in the Idaho National Laboratory fuel performance code Bison using Material, Kernel and AuxKernel methods. The results for the Material and Kernel implementation are shown in this report. The Material method implementation naturally fits the concept of modeling evolution of material state in the material models. Given the relatively high cost of thermodynamic computation, the advantage of the AuxKernel compared to Material implementation is the ability to control the execution and computational cost. The implemented formulations are illustrated on example problems which fit the experimental data and satisfy the computational requirements of conservation equations.

36 MATERIALS SCIENCE↗

Post-Transient Examination Results of RIA Commissioning Teats at the Transient Reactor Test Facility

Six reactivity-initiated-accident (RIA) commissioning tests have been performed at Idaho National Laboratory’s (INL) Transient Reactor Test Facility. Five of these tests were performed using fresh fuel rodlets, and the sixth test was performed using a previously irradiated rodlet from the ATF-2 irradiation experiment. These experiments demonstrate the ability to perform RIA testing including the ability to perform experiments using previously irradiated materials. Post-transient examinations revealed pellet cladding interactions and cladding ballooning depending on the initial boundary conditions of the test. Increased thickness of both zirconium oxide and alpha-zirconium were found in the cladding with tests performed at higher levels of total energy deposition. No oxide or alpha-zirconium was found in the test with the greatest amount of ballooning, which was also the test with the highest initial rod internal pressure. The highest energy deposition test (1110 J/g) resulted in failure of the rodlet. Fragmentation of the fuel pellets occurred in this test with the size of the fragments decreasing inversely to the pellet radius.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Strain-path change tests and physically based polycrystalline modeling of the behavior of recrystallized zirconium alloys - 2016-0092

Zirconium alloys used as cladding tubes for the fuel of pressurized water reactors can undergo high applied stress during power transients. In these conditions, biaxial loading may lead to plastic deformation of the cladding. A comprehensive understanding of the material mechanical response during loading-path change tests is an important step toward predicting the behavior in these specific conditions. Using a nonstandard mechanical testing machine, we conducted loading-path change tests at 623 K on as-received recrystallized Zircaloy-4 tubes. These tests consisted of an axial tensile loading and unloading followed by an internal pressure (or pure hoop tension) loading and unloading. These tests examined the kinematic and isotropic hardening components of the strain hardening behavior of the thin cladding tubes. The isotropic hardening was attributed to dislocation multiplication and dislocation-dislocation interactions. The kinematic hardening was attributed to the interaction of the grains with each other. A polycrystalline model was also used and improved in order to simulate the tests. A good prediction of the isotropic and kinematic hardening was provided by the modeling. (authors)

36 MATERIALS SCIENCE↗

Site-specific nanoscale characterization of zirconium hydrides in the hydride rim structure of hydrogen-charged zircaloy-4 cladding

Zircaloy-4 cladding tubes were hydrogen-charged using a static pressure hydrogen charging method to experimentally produce a zirconium-hydride rim structure similar to that observed in Zircaloy-4 cladding tubes after operation in commercial light-water nuclear reactors. Detailed characterization was performed using electron energy loss spectroscopy (EELS) and four-dimensional scanning transmission electron microscopy (4D-STEM). By mapping the position of the plasmon peak in the low-loss EELS spectrum, it was determined that the zirconium hydrides formed were predominantly solid δ-phase zirconium hydrides. 4D-STEM was used to generate phase and strain maps at nanoscale. The interactions between two zirconium hydride platelets less than 200 nm apart leads to the localized lattice rotations and possible phase change, revealing a potential hydride growth mechanism.

4D-STEM↗

Summary and Assessment of Metal Fuel Capabilities in Bison

The work described in this report was performed under funding from the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. This report is issued in satisfaction of the level 2 milestone M2MS-18IN0201012 on Release Bison with Initial Metallic Fuel Performance Capability. Metallic fuel development began in Bison a few years after light water reactor development began (~ 2009). The initial development was funded by the Advanced Fuels Campaign. With recent interest from industry, the Nuclear Regulatory Commission, and the Versatile Test Reactor program, the Nuclear Energy Modeling Simulation Program has increased funding for metallic fuel development and validation in FY18. The material models with the most mature development are binary uranium-zirconium and ternary uranium-plutonium-zirconium alloys for fuel and HT9 for cladding. The first material models were incorporated from open literature sources, which include mechanical and thermal material properties that are functions of temperature, porosity, and zirconium concentration, swelling, fission gas release, zirconium diffusion, creep, and sodium coolant channel boundary conditions. These material models have been utilized to simulate ERB-II fuel pins and the results compared to measurements. The EBR-II experiment measurements mostly consist of cladding strain and zirconium redistribution. As such, comparisons to EBR-II measurements are currently limited to these two figures of merit. Simulations and comparisons to experiment measurements have also been done for transient testing of EBR- II fuel pins in TREAT, which included temperature measurements. Preliminary comparisons between Bison calculations and experiment measurements are favorable. These comparisons have also highlighted the importance of the fuel swelling and fission gas release models on cladding strain. Also, Bison metallic fuel simulations are currently being run in support of ATR experiments and VTR exploratory design calculations. Further development and evaluation of individual material models and fuel system models are planned for FY19.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Fabrication of FeCrAl cladding for accident tolerant fuel

General Electric (GE) is working with the US Department of Energy (DOE) to develop a high performance nuclear fuel that will have enhanced tolerance to failure under severe accident conditions. The leading FeCrAl alloy candidate for the cladding is APMT, a powder metallurgy alloy developed by Kanthal/Sandvik for high temperature applications. Zirconium alloys have been used for cladding the fuel for over six decades. That is, zirconium alloys can be fabricated as several meters long tubes with an outside diameter of about 10 mm and a wall thickness in the order of 1 mm or less. The GE team is working to demonstrate that APMT can be fabricated as long thin walled tubes for the cladding application. A first production route for 10 x 0.5 mm APMT tubes was established showing that pilgering and drawing are viable processes to fabricate the APMT cladding. The mechanical properties of APMT were evaluated before and after the cold forming. Yield strength, tensile strength and hardness increased and elongation decreased due to cold forming, which was expected. No abnormality has been noted in the microstructures. The 3D image visualization of TIG welded APMT rodlet and the images of welding areas taken at different locations show no welding defects such as distortion, porosity or lack of penetration along/adjacent to the welded areas, confirming overall good welds and no opening or defects across the entire width of the tube thickness. Tests have shown no effect of welding on the hardness or mechanical strength in the area nearing the weld.

36 MATERIALS SCIENCE↗

Quantify Contaminant Partitioning in Advanced Chlorination Process

Chlorination of the zirconium in spent fuel cladding has been identified as a significant means of Zr recovery and purification, enabling a 25% (wt.) or greater reduction in waste requiring geologic disposition. Several chlorinating agents have been identified over the last few years, including gases and liquids. In particular, the interactions of sulfur-chloride solvents with zirconium may allow the separation of zirconium from uranium with the selective formation of ZrCl 4 . This report reviews what is known about reactions between chlorinating agents, uranium metal and oxide, Zircaloy and its alloying elements, and fission plus activation products. Plans for testing the feasibility of applying advanced chlorination methods developed for zirconium recycling to the front end of the recycling of spent nuclear fuel are also included.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

ENDF/B-VIII.1: Thermal Neutron Scattering Sublibrary

The thermal neutron scattering law (TSL) sublibrary aims to describe the interaction of incident neutrons at thermal or sub-thermal energies with different compound materials such fuels, moderators and special-purpose materials. In ENDF/B-VIII.1 there was a large number of new and updated TSL evaluations, including traditional moderators (light water, Beryllium metal, Beryllium Oxide, Calcium Hydride, plastics (Polystyrene and Lucite), graphite (reactor-grade and crystalline), anhydrous Hydrogen Fluoride, and heavy paraffinic oil); exotic moderators (Beryllium Carbide, Zirconium Hydride, Yttrium Hydride, Lithium-7 Hydride and Deuteride), FLiBe molten salt, structural materials and cladding (Silicon Carbide, Silicon Dioxide, Zirconium Carbide), fuels (Plutonium Dioxide, Uranium Carbide, Uranium metal, Uranium Nitride, Uranium Dioxide, Uranium Hydride), and special purpose materials. In ENDF/B-VIII.1 we also distribute alongside the evaluated files, a comma-separated file (CSV), named TSL_MAT_numbers.csv, which lists all evaluated files in the current release and their corresponding unique MAT number.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Modeling of complex wear behavior associated with grid-to-rod fretting in light water nuclear reactors

One significant concern in the operation of light water nuclear reactors is the fretting wear damage to fuel cladding from flow-induced vibrations. For years, research on the grid-to-rod fretting (GTRF) phenomena has been underway in countries where nuclear power production is a significant industry. Under the auspices of the U.S. Department of Energy Consortium for Advanced Simulation of Light Water Reactors, an effort has been underway to develop and test an engineering wear model for zirconium alloy fuel rod cladding against a supporting grid. Furthermore, the multi-stage model accounts for oxide layers and wear rate transitions. Our paper describes the basis for a GTRF engineering wear model, the physical significance of the wear factor it contains, and recent progress toward model validation based on a fretting wear testing apparatus that accounts for coolant temperature, pressure, and the presence of periodic impacts (gaps) in grid/rod contact.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Irradiation of Wrought FeCrAl Tubes in the High Flux Isotope Reactor

The Advanced Fuels Campaign within the Nuclear Technology Research and Development program of the Department of Energy Office of Nuclear Energy is seeking to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are one of the leading candidate materials for fuel cladding to replace traditional zirconium alloys because of the superior oxidation resistance of FeCrAl. However, there are still some unresolved questions regarding irradiation effects on the microstructure and mechanical properties of FeCrAl at end-of-life dose levels. In particular, there are concerns related to irradiation-induced embrittlement of FeCrAl alloys due to secondary phase formation. To address this issue, Oak Ridge National Laboratory has developed a new experimental design to irradiate shortened cladding tube specimens with representative 17×17 array pressurized water reactor diameter and thickness in the High Flux Isotope Reactor (HFIR) under relevant temperatures (300–350°C). Post-irradiation examination will include studies of dimensional change, microstructural changes, and mechanical performance. This report briefly summarizes the capsule design concept and the irradiation test matrix for six rabbit capsules. Each rabbit contains two FeCrAl alloy tube specimens. The specimens include Generation I and Generation II FeCrAl alloys with varying processing conditions, Cr concentrations, and minor alloying elements. The rabbits were successfully assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package. Pictures of the rabbit assembly process and detailed dimensional inspection of select specimens are included in this report. The rabbits were inserted into HFIR starting in cycle 472 (May 2017).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Irradiation performance of U-Mo monolithic fuel

High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Advanced 3D characterisation of iodine induced stress corrosion cracks in zirconium alloys

Highlights: • Macroscopic crack direction is affected by metallurgical state. • High resolution XCT reveals complex crack morphology. • 3D-EBSD used to fully characterise attacked grain boundaries. • Evidence of iodine segregation ahead of a crack tip. - Abstract: In this study, a range of advanced techniques have been used to characterise iodine-induced stress corrosion cracks, and investigate their interaction with microstructure in zirconium alloys used as cladding material in nuclear power reactors. C-ring samples, machined from cladding tubes in a recrystallised and cold worked condition, were subjected to compressive stress while submerged in an iodine-ethanol solution to simulate stress corrosion cracks arising from pellet cladding interaction. The complex morphology of these stress corrosion cracks was imaged in 3D using X-ray computed tomography (XCT), and subsequent analysis using both 2D and 3D electron backscatter diffraction (EBSD) allowed for the interaction of these cracks with microstructure to be investigated. Intergranular cracking modes were observed to dominate in the recrystallised microstructure, resulting in cracking in the radial direction with variations on the order of the grain size. Crack propagation in the cold worked microstructure was observed to be correlated with the direction of the elongated grains, resulting in a more tangential crack path with less local variation and a microstructure potentially more susceptible to iodine-induced stress corrosion cracking. 3D-EBSD using serial sectioning allows for the determination of the full five-parameter grain boundary description, and opens up the possibility of a statistical analysis of susceptible boundaries to SCC. Nanoscale secondary ion mass spectrometry (NanoSIMS) was also used to map the chemical distribution in the crack tip region and provides possible evidence of iodine segregation ahead of the crack tip. The results provide new insight into crack propagation mechanisms in recrystallised and cold worked microstructures, and pave the way for a more mechanistic understanding of crack propagation through advanced characterisation of iodine-induced stress corrosion cracks in zirconium alloys.

36 MATERIALS SCIENCE↗

Degradation and Failure Phenomena of Accident Tolerant Fuel Concepts: FeCrAl Alloy Cladding

The U.S. Nuclear Regulatory Commission (NRC) is anticipating licensing applications and commercial use of accident tolerant fuel (ATF) in United States commercial nuclear power reactors. Pacific Northwest National Laboratory is providing technical assistance to the NRC related to the newly proposed nuclear fuel and cladding designs. This report focuses specifically on the iron-chromium-aluminum (FeCrAl) alloys being investigated to replace zirconium-based alloys for fuel cladding and provides current state-of-the-industry information on material properties and fuel performance considerations under reactor operating conditions and design basis accident conditions. Currently, Global Nuclear Fuels (GNF) is the only U.S.-based fuel vendor considering FeCrAl cladding as a near-term ATF design. GNF has tested several different FeCrAl alloys including Kanthal APMT, C26M, and MA956; fueled C26M rods and unfueled Kanthal APMT, C26M, and MA956 rods have been inserted in lead test assemblies at Plants Hatch and Clinton. This report will focus on those three alloys specifically, with broader information given regarding in-reactor and ex-reactor testing of FeCrAl alloys. To support the NRC’s readiness efforts, this report will identify and discuss degradation and failure modes of FeCrAl cladding concepts, including fuel performance characteristics that may not be addressed within existing regulatory documents.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Development and validation of capabilities to measure thermal properties of layered monolithic U-Mo alloy plate-type fuel

The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world’s highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of thermal conductivity behavior of the fuel system as a function of temperature and expected irradiation conditions. The purpose of this paper is to verify and validate the functionality of equipment methods installed in hot cells for eventual measurements on irradiated uranium-molybdenum (U-Mo) monolithic fuel specimens, procedures to operate the equipment, and models to extract the desired thermal properties. The results presented here demonstrate the adequacy of the equipment, procedures and models that have been developed for this purpose based on measurements conducted on surrogate depleted uranium-molybdenum (DU-Mo) alloy samples containing a zirconium diffusion barrier and clad in aluminum alloy 6061 (AA6061). The results are in excellent agreement with thermal property data reported in the literature for similar U-Mo alloys as a function of temperature.

multi-layer↗

Diffusion Barrier Properties of Nitride-Based Coatings on Fuel Cladding

In this work titanium nitride (TiN) and zirconium nitride (ZrN) coatings are proposed as diffusion barriers between stainless steel nuclear fuel cladding and lanthanide fission products. TiN and ZrN have been coated as barrier materials between pure Fe and Ce, i.e. diffusion couples of Fe/TiN/Ce and Fe/ZrN/Ce, annealed up to a temperature of 600 degrees C, and compared to the diffusion behavior of uncoated Fe/Ce. Backscattered electron images and electron dispersive X-ray spectroscopy measurements confirm that, with a 500 nm TiN or ZrN layer, no obvious diffusion is observed between Fe and Ce. Basic diffusion characteristics of the Fe/Ce couple have also been measured and compared with the TiN and ZrN coated ones. The results strongly advocate that TiN and ZrN coatings provide reliable diffusion barrier characteristics against Ce and possibly other lanthanide fission products.

36 MATERIALS SCIENCE↗

Closed Fuel Cycle Waste Treatment Strategy

This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) (Gombert et al. 2008) as a general strategy and associated Waste Treatment Baseline Study (WTBS) (Gombert et al. 2007). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significant additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to a representative full recycle reprocessing flowsheet. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. These applications are occasionally discussed where the data are more readily available. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include; Tritium in low-water cement in high integrity containers (HICs); Iodine-129: As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form with encapsulated nano-sized AgI crystals; Carbon-14 immobilized as a CaCO3 in a cement waste form; Krypton-85 stored as a compressed gas; An aqueous reprocessing high-level waste (HLW) raffinate waste immobilized by the vitrification process; An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel either included in the borosilicate HLW glass or immobilized in the form of a metal alloy or titanate ceramics; Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware super-compacted for disposal or purified for reuse (or disposal as low-level waste, LLW) of Zr by reactive gas separations; Electrochemical process salt HLW incorporated into a glass bonded Sodalite waste form; and Electrochemical process UDS and SS cladding hulls melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from a representative flowsheet are reported. In addition to the above listed primary waste streams, a range of secondary process wastes are generated by aqueous reprocessing of LWR fuel, metal SFR fuel fabrication, and electrochemical reprocessing of SFR fuel. These secondary wastes have been summarized and volumes estimated by type and classification. The important waste management data gaps and research needs have been summarized for each primary waste stream and selected waste process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗