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At least 1,369 records · Page 76
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Stress rupture of SiC/SiC composite tubes under high-temperature steam

SiC-fiber–reinforced SiC matrix composite cladding for light water reactor fuel elements must withstand high-temperature steam oxidation in a loss-of-coolant accident scenario (LOCA). Current composite designs include an outer monolithic SiC layer, in part, to increase steam oxidation resistance. However, it is not clear how such a structure would behave under high-temperature steam in the case when the monolithic layer cracks and carbon interphases and SiC fibers are exposed to the environment. To fill this knowledge gap, stress-rupture tests of prototypic SiC composite cladding at 1000°C under steam and inert environments were conducted. The applied stress was ~120 MPa, which was beyond the initial cracking stress. The failure lifetime under steam was 400–1300 s, while 75% of the composite specimens did not fail after 3 h of total exposure under inert gases. Microstructural observations suggest that steam oxidation activated slow crack growth in the fibers, which led to failure of the composite. The results from this study suggest that stress rupture in steam environments could be a limiting factor of the cladding under reactor LOCA conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Assessment of thermodynamic stability of sapphire in eutectic molten chloride environment

The continued development of molten salt reactors requires highly efficient and stable instruments that can efficiently track monitor conditions of the molten salt during long-term application operation in both the fuel and coolant/heat transfer fluid loops. Sapphire (Al 2 O 3 ) fibers have shown tremendous potential by enabling radiation resistance and a broader operational range of temperature . Thermodynamic modeling using the ThermoCalc and the SGTE Molten Salts (SALT1) and Pure Substances (Pure5) database is done performed to understand the compatibility of Al 2 O 3 fibers with NaCl-MgCl 2 eutectic molten salt in the temperature range of 600 to 1000°C. The thermodynamic calculations show that sapphire fibers are not expected to be compatible with long term exposure to molten chloride salts within this temperature range. Furthermore, various pure metallic elements were evaluated as potential cladding materials for part of the Al 2 O 3 fibers. Based on the thermodynamic analysis, molybdenum (Mo) and nickel (Ni) could be effective cladding materials to enhance the stability of Al 2 O 3 in NaCl and MgCl 2 chloride salt molten bath in the desired temperature range. The presence of a metallic cladding leads to the suppression of aluminum dissolution from Al 2 O 3 , thus improving its stability in chloride salts. Additionally, it is also shown that Al 2 O 3 remains stable up to 2300 K in the presence of pre-existing Al 2 MgO 4 and Al 2 NiO 4 in the eutectic molten chloride bath environment.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

THERMAL SAFETY MARGIN CALCULATION OF THE MP-2 EXPERIMENT IN THE ADVANCED TEST REACTOR

The Mini-Plate 2 (MP-2) irradiation test is a fueled experiment designed for irradiation in multiple test locations in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). The experiment is a drop-in test where small aluminum-clad fuel plate samples (mini plates) are cooled directly by the ATR Primary Coolant System (PCS) water. The MP-2 fuel plate experiment will be irradiated in several different irradiation locations of the ATR. This fueled experiment contains aluminum-clad fuel mini plates consisting of monolithic U-Mo. Four different types of fuel plates with fuel meat thickness and cladding are part of the MP-2 test. A thermal analysis has been performed on the MP-2 experiment. A method for calculating Departure from Nucleate Boiling Ratio (DNBR) and Flow Instability Ratio (FIR) during a reactivity transient using the commercial finite element and heat transfer code ABAQUS is discussed. At the start of an ATR cycle the heat generation rate of the fueled experiment is high and the heat rate multiplier from the outer shim control cylinders is low, while the reverse is true at the end of the ATR cycle. Thermal analyses at 10 day increments during the cycle calculate the DNBR and FIR during a reactivity transient. This technique calculates DNBR for the fuel plate surfaces and FIR for all water components for each finite element surface and node at various times during the ATR cycle. Heat rates vary with time during the transient calculation that are provided by a detailed physics analysis. Oxide growth on the fuel plates is also incorporated Results from the transient calculations are displayed with the ABAQUS post processor. By calculating these parameters at each location in the finite element model, conservatism is replaced with accuracy. This allows for a greater margin for the thermal hydraulic safety parameters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Effects of the foil flatness on the stress-strain characteristics of U10Mo alloy based monolithic mini-plates

The effects of the foil flatness on stress-strain behavior of monolithic fuel mini-plates during fabrication and irradiation were studied. Monolithic plate-type fuels are a new fuel form being developed for research and test reactors to achieve higher uranium densities. This concept facilitates the use of low-enriched uranium fuel in the reactor. These fuel elements are comprised of a high density, low enrichment, U–Mo alloy based fuel foil encapsulated in a cladding material made of Aluminum. To evaluate the effects of the foil flatness on the stress-strain behavior of the plates during fabrication, irradiation and shutdown stages, a representative plate from RERTR-12 experiments (Plate L1P756) was considered. Both fabrication and irradiation processes of the plate were simulated by using actual irradiation parameters. The simulations were repeated for various foil curvatures to observe the effects of the foil flatness on the peak stress and strain magnitudes of the fuel elements. Results of fabrication simulations revealed that the flatness of the foil does not have a considerable impact on the post fabrication stress-strain fields. Furthermore, the irradiation simulations indicated that any post-fabrication stresses in the foil would be relieved relatively fast in the reactor. While, the perfectly flat foil provided the slightly better mechanical performance, overall difference between the flat-foil case and curved-foil case was not significant. Even though the peak stresses are less affected, the foil curvature has several implications on the strain magnitudes in the cladding. It was observed that with an increasing foil curvature, there is a slight increase in the cladding strains.

42 ENGINEERING↗
DOE OSTI2015

THE EFFECTS OF FABRICATION INDUCED RESIDUAL STRESS-STRAIN STATES ON THE IRRADIATION PERFORMANCE OF MONOLITHIC MINI-PLATES

This article evaluates the initial stress-strain state of the monolithic fuel plates and their effects on the overall irradiation performance. The monolithic fuel is a proposed fuel form for high-performance research reactors to achieve higher uranium densities within the core to allow the use of low-enriched uranium fuel. These plate-type fuels consist of a high density low enrichment Uranium-Molybdenum alloy based fuel, a diffusion barrier made of Zirconium, and a cladding made of Aluminum. Because the fabrication of monolithic plates involves multiple stages and the foil material is subjected to various thermo-mechanical processes, a presence of significant stress-strain fields is likely. In addition, the choice of the fabrication process may also cause variations in the foil’s microstructure. There are interests if the overall irradiation performance could be affected by the different initial stress-strain state and the microstructural characteristics. To investigate these concerns, a representative plate was simulated for various initial stress-strain states. First, the co-rolling process for thickness reduction stage of the foil was simulated. For this, a scaled version of the process was evaluated to identify distinct stress profiles on the selected paths. These profiles were then used to define an initial state for the foil before Hot Isostatic Pressing. Additional HIP simulations were also performed to study the cases with stress-free foils prior HIP bonding. The purpose of these simulations was to mimic the stress relief treatment after co-rolling, which aims to remove the residual stresses and minimize the material anisotropy. For the HIP process with initially stress-free foil, several bonding temperatures were considered, including the nominal bonding temperature. The final case considered a plate without initial stresses and strains, assuming a hypothetical case in which the plate would be subjected to a post-fabrication heat treatment to relieve the process-induced stresses before irradiation. Finally, the irradiation processes were simulated for all the cases with distinct pre-irradiation stress-strain profiles. For all cases, the same irradiation parameters were used. The resulting distortions, stress-strain fields and temperature profiles were gathered on selected locations. Finally, a comparative evaluation was made to determine the sensitivity to the pre-irradiation stress-strain variations. The irradiation simulations have revealed that fabrication induced stresses in fuel would be relieved quickly due to large irradiation creep strain. The results indicated that the stress-strain behavior of the fuel foil is not affected significantly by the fabrication process. It is important to note that this excludes microstructural characteristics, which may have important implications on overall behavior. In the cladding however, the stresses continue increase and additional plastic strains are generated. This is due to irradiation hardening by fast neutrons. Additional stresses and strains are formed in the cladding as a result of displacements caused by the volumetric swelling of the fuel foil.

42 ENGINEERING↗
DOE OSTI2015

EFFECTS OF THE FOIL CENTERING ON THE IRRADIATION PERFORMANCE OF U10MO ALLOY BASED MONOLITHIC MINI-PLATES

Monolithic plate-type fuel is a fuel form being developed for high performance research and test reactors to minimize the use of enriched material. These plate-type fuels consist of a high Uranium density, LEU foil contained within diffusion barriers and encapsulated within a cladding material. To benchmark this new design, effects of various geometrical and operational variables on irradiation performance have been evaluated. For this work, the effects of fuel foil centering on the thermo-mechanical performance of the mini-plates were studied. To evaluate these effects, a selected plate from RERTR-12 experiments, the Plate L1P756, was considered. The fuel foil was moved within the fuel plate to study the effects of the fuel centering on stress, strain and overall shape of the fuel elements. The thickness of the fuel foil, thickness of the Zr-liners and total thickness of the plate were held constant, except the centerline alignment of the fuel foil. For this, the position of the fuel foil was varied from the center position to a maximum offset corresponding to the minimum allowable aluminum cladding thickness of 0.1524 mm. Results were then compared with each other and with the ideal case to make a comparative assessment. Fabrication simulations indicated that the thermal expansion mismatch results in warping of the fuel plate during fabrication as the fuel plate is cooled from the HIP temperature when the fuel is not centered. Even if the model is constrained during cooling to simulate the rigid HIP can surrounding the fuel plate during cooling, warping is observed when the constraint is removed. Similarly, irradiation simulations revealed that the fuel offset causes virtually all irradiation-induced swelling to occur on the thin-cladding side of the plate. This is observed even for the smallest offset that was considered. The total magnitude of the swelling is approximately same for all offsets values.

42 ENGINEERING↗
DOE OSTI2017

Thermo-mechanical performance assessment of selected plates from MP-1 high power experiments

Monolithic plate-type fuel is a fuel form that is being developed for the conversion of high performance research and test reactors to low-enrichment uranium fuels. These fuel-plates are comprised of a high density, low enrichment, U-Mo alloy based fuel foil encapsulated in an aluminum cladding. To benchmark this new design, number of plates has been irradiated with satisfactory performance. As a part of continuing evaluation efforts, a set of plates covering range of operational parameters is scheduled to be tested during MP-1 irradiation experiments. It is necessary to evaluate the thermo-mechanical performance of plates during irradiation. For this, selected plates with distinct operational histories; covering low power, high power and high fission density were simulated. Fully coupled three-dimensional models of plates with a capability to evolve mechanical and thermal properties of constituent materials with irradiation time and burn-up were developed. The models input used projected parameters, including plate geometry, irradiation history and coolant conditions as input. The model output included temperature, displacement and stresses in the fuel, cladding and diffusion barrier. The fuel behavioral model considered inelastic behavior including volumetric swelling due to solid and gaseous products, irradiation induced creep, thermal expansion, conductivity degradation and plasticity. A visco-plastic behavioral model was used for the cladding that included thermal creep, irradiation hardening, growth due to fast neutrons and Mises plasticity. The plates were then simulated by using projected irradiation parameters. The resulting temperature, displacement and stress-strains were comparatively evaluated on the selected paths. The results were then compared with those of plates from previous RERTR experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

FATIGUE BEHAVIOR OF SPENT NUCLEAR FUEL RODS IN SIMULATED TRANSPORTATION ENVIRONMENT

Nuclear fuel rod is composed of cladding tube and a specified number of fuel pellets contained. In the United States, spent nuclear fuel (SNF) is expected to be transported to at least one storage facility before permanent disposal. The fatigue behavior of spent nuclear fuel (SNF) rods under reversed cyclic bending must be understood in order to evaluate their vibration integrity in a transportation environment. This is especially important for high-burnup SNFs (>45 GWd/MTU). This report presents the experimental results related to Zircaloy (Zry)-4 - based surrogate rods and high-burnup SNFs, based on recent work performed at Oak Ridge National Laboratory (ORNL). The surrogate rod was made of Zry-4 cladding and alumina pellets, and high-burnup fuel rods were discharged from H.B. Robinson pressurized water reactor. The reversed cyclic bending testing was conducted at 5 Hz under loading control. The effect of pre-hydriding and burnup or irradiation on the flexural rigidity and fatigue life of cladding-pellet system were discussed. The fatigue data obtained are extremely useful to the future certification of SNF storage and transportation cask.

Wang, Jy-An John↗
DOE OSTI2019

MODELING SHOCK AND VIBRATION ON USED NUCLEAR FUEL DURING NORMAL CONDITIONS OF TRANSPORTATION

In 2017, the United States Department of Energy (DOE) collaborated with Spanish and Korean organizations to perform a multimodal transportation test to measure shock and vibration loads imparted to used nuclear fuel (UNF) assemblies. This test used real fuel assembly components containing surrogate fuel mass to approximate the response characteristics of real, irradiated used nuclear fuel. Pacific Northwest National Laboratory was part of the test team and used the data collected during this test to validate numerical models needed to predict the response of real used nuclear fuel in other transportation configurations. This paper summarizes the modeling work and identifies lessons learned related to the modeling and analysis methodology. The modeling includes railcar dynamics using the NUCARS software code and explicit dynamic finite element modeling of used nuclear fuel cladding in LS-DYNA. The NUCARS models were validated against railcar dynamics data collected during captive track testing at the Federal Railroad Administration's Transportation Technology Center in Pueblo, CO. The LS-DYNA models of the fuel cladding were validated against strain gage data collected throughout the test campaign. One of the key results of this work was an assessment of fuel cladding fatigue, and the methods used to calculate fatigue are detailed in this paper. The validated models and analysis methodologies described in this paper will be applied to evaluate future UNF transportation systems.

Klymyshyn, Nicholas A.↗
DOE OSTI2023

MODELING A SPENT NUCLEAR FUEL CASK SEISMIC TEST

The US Department of Energy Spent Fuel and Waste Science and Technology (SFWST) program is planning to conduct a series of full-scale shake table tests to simulate hypothetical earthquake conditions and record the response of surrogate spent nuclear fuel (SNF) assemblies in a canister dry storage system mockup. The shake table motions will represent a range of hypothetical earthquake conditions at hypothetical locations in the continental US to generally define the range of mechanical loads that SNF can be expected to experience during extended dry storage periods. The earthquake conditions will represent seismic hazards in the 2,000-to-20,000-year return period range. The test will use instrumented pressurized water reactor fuel assemblies with surrogate mass inside the fuel rods instead of radioactive fuel pellets. Dummy assemblies with similar mass and dimensions to actual fuel assemblies will occupy the rest of the fuel assembly locations in a SNF canister that holds 32 fuel assemblies. The canister will be located inside a mockup vertical concrete cask. Instrumentation will record the motion of the major components of the complex dynamic system, and strain gauges will be used to record the cladding strain at select locations. Preparations for the test require modeling predictions to identify the range of response of the system and to help select specific earthquake cases to be simulated on the shake table from a large set of potential cases. This paper describes the pretest nonlinear finite element modeling efforts that have been completed to date, including cask system level modeling and fuel assembly modeling in LS-DYNA. The cask system level models are critical for anticipating sliding or tipping of an unanchored cask during the test. The fuel assembly model is needed to estimate the range of cladding strain response and fuel assembly structural response to be expected during the test and cladding strain measurements will be key metrics for model validation and the development of modeling best practices after the test is concluded.

Klymyshyn, Nicholas A.↗
DOE OSTI2014

Reduced yield stress for zirconium exposed to iodine: Reactive force field simulation

Iodine-induced stress-corrosion cracking (ISCC), a known failure mode for nuclear fuel cladding, occurs when iodine generated during the irradiation of a nuclear fuel pellet escapes the pellet through diffusion or thermal cracking and chemically interacts with the inner surface of the clad material, inducing a subsequent effect on the cladding’s resistance to mechanical stress. To complement experimental investigations of ISCC, a reactive force field (ReaxFF) compatible with the Zr-I chemical and materials systems has been developed and applied to simulate the impact of iodine exposure on the mechanical strength of the material. The study shows that the material’s resistance to stress (as captured by the yield stress of a high-energy grain boundary) is related to the surface coverage of iodine, with the implication that ISCC is the result of adsorption-enhanced decohesion.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Integrated polarization-free Bragg filters with subwavelength gratings for photonic sensing

We present polarization-free Bragg filters having subwavelength gratings (SWGs) in the lateral cladding region. This Bragg design expands modal fields toward upper cladding, resulting in enhanced light interaction with sensing analytes. Two device configurations are proposed and examined, one with index-matched coupling between transverse electric (TE) and transverse magnetic (TM) modes and the other one with hybrid-mode (HM) coupling. Both configurations introduce a strong coupling between two orthogonal modes (either TE-TM or HM 1 -HM 2 ) and rotate the polarization of the input wave through Bragg reflection. The arrangements of SWGs help to achieve two configurations with different orthogonal modes, while expanding modal profiles toward the upper cladding region. Our proposed SWG-assisted Bragg gratings with polarization independency eliminate the need for a polarization controller and effectively tailor the modal properties, enhancing the potential of integrated photonic sensing applications.

Optics↗
DOE OSTI2014

Accident Tolerant Fuel Analysis

Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for risk-informed margins management by improving economics and reliability, and sustaining safety, of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory is performing a series of case studies that will explore methods- and tools-development issues, in addition to being of current interest in their own right. One such study is a comparative analysis of safety margins of plants using different fuel cladding types: specifically, a comparison between current-technology Zircaloy cladding and a notional “accident-tolerant” (e.g., SiC-based) cladding. The present report begins the process of applying capabilities that are still under development to the problem of assessing new fuel designs. The approach and lessons learned from this case study will be included in future Technical Basis Guides produced by the RISMC Pathway. These guides will be the mechanism for developing the specifications for RISMC tools and for defining how plant decision makers should propose and evaluate margin recovery strategies.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2015

Thermal-Hydraulic Simulations of Single Pin and Assembly Sector for IVG- 1M Reactor

Thermal-hydraulic simulations have been performed using computational fluid dynamics (CFD) for the highly-enriched uranium (HEU) design of the IVG.1M reactor at the Institute of Atomic Energy (IAE) at the National Nuclear Center (NNC) in the Republic of Kazakhstan. Steady-state simulations were performed for both types of fuel assembly (FA), i.e. the FA in rows 1 & 2 and the FA in row 3, as well as for single pins in those FA (600 mm and 800 mm pins). Both single pin calculations and bundle sectors have been simulated for the most conservative operating conditions corresponding to the 10 MW output power, which corresponds to a pin unit cell Reynolds number of only about 7500. Simulations were performed using the commercial code STAR-CCM+ for the actual twisted pin geometry as well as a straight-pin approximation. Various Reynolds-Averaged Navier-Stokes (RANS) turbulence models gave different results, and so some validation runs with a higher-fidelity Large Eddy Simulation (LES) code were performed given the lack of experimental data. These singled out the Realizable Two-Layer k-ε as the most accurate turbulence model for estimating surface temperature. Single-pin results for the twisted case, based on the average flow rate per pin and peak pin power, were conservative for peak clad surface temperature compared to the bundle results. Also the straight-pin calculations were conservative as compared to the twisted pin simulations, as expected, but the single-pin straight case was not always conservative with regard to the straight-pin bundle. This was due to the straight-pin temperature distribution being strongly influenced by the pin orientation, particularly near the outer boundary. The straight-pin case also predicted the peak temperature to be in a different location than the twisted-pin case. This is a limitation of the straight-pin approach. The peak temperature pin was in a different location from the peak power pin in every case simulated, and occurred at an inner pin just before the enrichment change. The 600 mm case demonstrated a peak clad surface temperature of 370.4 K, while the 800 mm case had a temperature of 391.6 K. These temperatures are well below the necessary temperatures for boiling to occur at the rated pressure. Fuel temperatures are also well below the melting point. Future bundle work will include simulations of the proposed low-enriched uranium (LEU) design. Two transient scenarios were also investigated for the single-pin geometries. Both were “model” problems that were focused on pure thermal-hydraulic behavior, and as such were simple power changes that did not incorporate neutron kinetics modeling. The first scenario was a high-power, ramp increase, while the second scenario was a low-power, step increase. A cylindrical RELAP model was also constructed to investigate its accuracy as compared to the higher-fidelity CFD. Comparisons between the two codes showed good agreement for peak temperatures in the fuel and at the cladding surface for both cases. In the step transient, temperatures at four axial levels were also computed. These showed greater but reasonable discrepancy, with RELAP outputting higher temperatures. These results provide some evidence that RELAP can be used with confidence in modeling transients for IVG.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Report on status of execution of SiC step document

Advanced fuel claddings made entirely or mainly of silicon carbide (SiC) ceramics and/or composites are considered very attractive elements of the accident-tolerant fuels for the light water reactors. In order to translate the promise of SiC composite materials into a reliable fuel cladding, a coordinated program of component level design and materials development must be carried out with many key feasibility issues addressed a-priori to inform the process. With the primary objective of developing a draft blueprint of a technical program that addresses the critical feasibility issues; assesses design and performance issues related with manufacturing, operating, and off-normal events; and advances the technological readiness levels in essential technology elements, a draft plan for the Systematic Technology Evaluation Program for SiC/SiC Composite Accident-Tolerant LWR Fuel Cladding and Core Structures was developed in the FY-14 Advanced Fuels Campaign of the U.S. Department of Energy’s Fuel Cycles Research and Development Program. This document summarizes the status of execution of the technical plan within the activities at the Oak Ridge National Laboratory.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Severe Accident Scoping Simulations of Accident Tolerant Fuel Concepts for BWRs

Accident-tolerant fuels (ATFs) are fuels and/or cladding that, in comparison with the standard uranium dioxide Zircaloy system, can tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operations. It is important to note that the currently used uranium dioxide Zircaloy fuel system tolerates design basis accidents (and anticipated operational occurrences and normal operation) as prescribed by the US Nuclear Regulatory Commission. Previously, preliminary simulations of the plant response have been performed under a range of accident scenarios using various ATF cladding concepts and fully ceramic microencapsulated fuel. Design basis loss of coolant accidents (LOCAs) and station blackout (SBO) severe accidents were analyzed at Oak Ridge National Laboratory (ORNL) for boiling water reactors (BWRs). Researchers have investigated the effects of thermal conductivity on design basis accidents, investigated silicon carbide (SiC) cladding, as well as the effects of ATF concepts on the late stage accident progression. These preliminary analyses were performed to provide initial insight into the possible improvements that ATF concepts could provide and to identify issues with respect to modeling ATF concepts. More recently, preliminary analyses for a range of ATF concepts have been evaluated internationally for LOCA and severe accident scenarios for the Chinese CPR1000 and the South Korean OPR-1000 pressurized water reactors (PWRs). In addition to these scoping studies, a common methodology and set of performance metrics were developed to compare and support prioritizing ATF concepts. A proposed ATF concept is based on iron-chromium-aluminum alloys (FeCrAl). With respect to enhancing accident tolerance, FeCrAl alloys have substantially slower oxidation kinetics compared to the zirconium alloys typically employed. During a severe accident, FeCrAl would tend to generate heat and hydrogen from oxidation at a slower rate compared to the zirconium-based alloys in use today. The previous study, [2], of the FeCrAl ATF concept during station blackout (SBO) severe accident scenarios in BWRs was based on simulating short term SBO (STSBO), long term SBO (LTSBO), and modified SBO scenarios occurring in a BWR-4 reactor with MARK-I containment. The analysis indicated that FeCrAl had the potential to delay the onset of fuel failure by a few hours depending on the scenario, and it could delay lower head failure by several hours. The analysis demonstrated reduced in-vessel hydrogen production. However, the work was preliminary and was based on limited knowledge of material properties for FeCrAl. Limitations of the MELCOR code were identified for direct use in modeling ATF concepts. This effort used an older version of MELCOR (1.8.5). Since these analyses, the BWR model has been updated for use in MELCOR 1.8.6, and more representative material properties for FeCrAl have been modeled. Sections 2 4 present updated analyses for the FeCrAl ATF concept response during severe accidents in a BWR. The purpose of the study is to estimate the potential gains afforded by the FeCrAl ATF concept during BWR SBO scenarios.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

CIRFT Testing of High-Burnup Used Nuclear Fuel Rods from the Pressurized Water Reactor and Boiling Water Reactor Environments (FY 2015 Status Report)

High burn-up spent nuclear fuel (SNF) cladding has a significant amount of microcracks and hydrides, which will reduce the stress intensity required for crack growth. The potential linking of microcracks during vibration loading could also reduce the fatigue threshold/incubation period significantly, accelerating fatigue failure. In addition to cladding damage, the microstructure of fuel pellets and the characteristics of the interfaces between fuel pellets and between fuel pellets and cladding are likely to be dramatically changed after high burn-up. These modifications can have a direct impact on the structural integrity and vibration response of SNF rods in transportation. As a result, vibration has been included as one of the mandatory testing conditions in the structural evaluation of packages used by the US Nuclear Regulation Commission (NRC) in transporting SNF (10 CFR §71.71). Prior to this program development, no current system is available to test the SNF and evaluate the performance of fuel rods during spent fuel transportation. The objective of this research is to develop a system that can appropriately test the response of high-burnup SNF rods in a simulated loading condition.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Modeling and Simulation of Used Nuclear Fuel During Transportation with Consideration of Hydride Effects and Cyclic Fatigue

The objective of this work is to understand the integrity of Used Nuclear Fuel (UNF) during transportation. Previous analysis work has been performed to look at the integrity of UNF during transportation but these analyses have neglected to analyze the effect of hydrides and flaws (fracture mechanics models to capture radial cracking in the cladding). In this study, the clad regions of interest are near the pellet-pellet interfaces. These regions can experience more complex stress-states than the rest of the clad during cooling and have a greater possibility to develop radially reoriented hydrides during vacuum drying.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗