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All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Simulation of fission gas diffusion in UO{sub 2} by vacancy clusters

The low solubility of fission gas in nuclear fuel leads to inter- and intra-granular bubbles. These bubbles affect performance of the fuel rod and increase the fuel temperature through the increase of porosity in the fuel pellets and decreased thermal conductivity across the pellet-clad gap as gas is released into the plenum. The nucleation, growth, and inter-connectivity of the fission gas bubbles ultimately depend on the diffusivity of gas atoms through the fuel matrix. Using mechanisms and data previously established from density functional theory (DFT) and empirical potential calculations, diffusion and reaction models have been derived for xenon atoms, uranium atoms, and vacancies in the substitutional and interstitial locations. Although similar to previous models, the diffusion and reaction rates have been re-derived in a more robust manner. Previous studies have under-predicted the experimentally observed diffusivity of xenon atoms in UO{sub 2}, however it has been previously shown that by including xenon atoms bound to two vacancies, the calculated diffusivity approaches the measured data. The derived model for species diffusion and inter-species reactions has been implemented in the MARMOT phase-field code in a modular structure so as to capture the enhanced diffusion of xenon due to bound gas + vacancy-cluster movement. The MARMOT code allows for the diffusion, reaction, and source rates to be solved simultaneously, resulting in steady state or transient concentrations of each species, and ultimately the diffusion behavior of xenon atoms. By including the diffusion of clusters of different combinations of xenon and vacancies, the calculated diffusivity more closely captures the measured diffusivity of xenon when using preliminary parameters. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

First-principles investigation of cerium and neodymium diffusion in BCC chromium and vanadium via vacancy-mediated transport

Lanthanide transport plays a crucial role in the performance and longevity of metallic nuclear fuels. This study examines the diffusion behavior of Ce and Nd—two major fission products—in body-centered cubic (BCC) Cr and V, which are potential liner or coating materials for mitigating fuel-cladding chemical interactions (FCCI). Using density functional theory (DFT) calculations and self-consistent mean-field (SCMF) analysis, the vacancy-mediated diffusion coefficients are evaluated. Our findings reveal that Ce and Nd act as oversized solutes and are strongly bound to vacancies in BCC Cr and V, with diffusivities in Cr significantly lower than in V and in hexagonal closed-packed (HCP) Zr, as investigated in our previous work. The activation energies for Ce and Nd diffusion are 3.39 and 3.32 eV, respectively, in BCC Cr, and 2.56 and 2.33 eV, respectively, in BCC V. Analysis of vacancy drag and partial diffusion coefficient ratios indicates a strong tendency for lanthanide enrichment at vacancy sinks in BCC Cr, and to a lesser extent in BCC V, with this effect persisting up to the melting point in Cr and remaining substantial for Nd in V at high temperatures. Under irradiation, the increase in vacancy concentration is expected to enhance lanthanide transport, potentially accelerating interactions at liner-cladding interfaces. Although BCC Cr exhibits relatively low lanthanide diffusivities under equilibrium conditions, the expected segregation tendencies under irradiation suggest that Zr liners may be a more favorable option. Further investigations using rate theory, cluster dynamics, and phase-field modeling are required to quantitatively assess the performance of these materials in reactor environments.

36 - MATERIALS SCIENCE↗
DOE OSTI2013

Neutron Diffraction Measurement of Residual Stresses in “OSU Plate”

Aluminum-clad U10Mo (including zirconium diffusion layer) fuel assembly plates are bound to be in a state of residual stress due to the mismatch of thermal expansion of the constituent phases. Due to the geometry of the plates, i.e. thin beams, it is expected that the stress will predominantly be in-plane. The thermal expansion of the aluminum cladding, zirconium diffusion layer and U10Mo foil are roughly α Al =22x10 -6 /K, α Zr =5x10 -6 /K, and α U10Mo =10x10 -6 /K. Note that the coefficient of thermal expansion of the zirconium is quoted as that in the a direction in the hexagonal close packed crystal structure because of the strong texture in the rolled zirconium foil, in which the a-axis are strongly preferred in the in-plane direction. During cooling from the hot isostatic pressing (HIP’ing) temperature, the zirconium contracts considerably less than the U10Mo, which in turn, contracts less than the aluminum. Because the aluminum dominates in terms of volume fraction, it is expected that the U10Mo foil and zirconium bonding layer will be in a residual stress state of in-plane compression at room temperature after bonding. This has been verified for the U10Mo foil in a mini fuel plate using x-ray diffraction at the Advanced Photon Source, but measurements of the stress state of the aluminum and zirconium were not possible at the APS. Also, the APS measurements were limited to mini size fuel plates Neutron diffraction experiments of residual stress on the “OSU plate” were completed on the Spectrometer for Materials Research at Temperature and Stress (SMARTS) at the Lujan Center at LANSCE between 2/19-21/2013. Longitudinal and transverse residual strain profiles were measured on an Al-clad U10-Mo OSU plate. Residual strains from the aluminum cladding, the zirconium bonding layer and the U10Mo fuel foil were recorded.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Postirradiation Examination of the ATF-1 Experiments - 2018 Status

The Advanced Fuels Campaign in collaboration with industry are working to develop enhanced accident tolerant fuels for light water reactors. The initial irradiations of these concepts were performed in the Idaho National Laboratory Advanced Test Reactor using simple drop-in style irradiations. These irradiations are collectively referred to as ATF 1. The postirradiation examination of ATF 1 has begun and some concepts have now completed initial examination. This report describes the results of examinations on ATF 1 irradiations designed to test the feasibility of enhancing UO 2 thermal conductivity with additives, ATF 1 irradiations of uranium silicide, and the current status of ATF 1 irradiations that are in process at the hot cell. Enhanced thermal conductivity uranium dioxide composites containing silicon carbide (UO 2 SiC) and diamond (UO 2 diamond) have been irradiated to low burnup. The conditions of this irradiation test and subsequent postirradiation examinations are discussed. These irradiations evaluate fuel microstructure and potential fuel cladding interaction under representative light water reactor power conditions. Both non destructive and destructive techniques have been used to evaluate fuel integrity, fission gas release, fission product distribution, burnup, fuel swelling and cladding strain. Examination of the UO 2 SiC pellets revealed enhanced cracking when compared to a UO 2 rodlet also irradiated under similar conditions. Additionally, instability of the SiC whiskers in the uranium dioxide matrix was observed in the pellet central region, where the local temperatures exceeded 1300 °C. The microstructure of the diamond added UO 2 was severely disrupted during irradiation, resulting in local migration of cesium along the fuel stack and increased fission gas release when compared with the expected release from the Vitanza curve at corresponding values of burnup and irradiation temperature. The postirradiation examination results cast doubt on the suitability of these additives to improve UO 2 fuel performance in a way that would lead to enhanced accident tolerance. Additionally, postirradiation examination data of U 3 Si 2 fuels at low burnup (i.e., < 20 GWd/tHM) for application in advanced Light Water Reactor (LWR) is presented. The U 3 Si 2 pellets show limited cracking in comparison to the expected behavior of UO 2 at same power level. In addition, gamma scanning data did not reveal migration of fission products. Minor homogeneous hardening along the pellet radius due to accumulation of fission products and radiation damage was measured by microindentation. The fission gas release and swelling remains very low. Formation of fission gas bubbles resolvable with optical microscopy occurs from the pellets center outward to approximately 60% of the fuel pellet radius. The overall data suggest a good performance of this accident tolerant fuel candidate at low burnup. Finally, the current status of postirradiation examination on other ATF 1 irradiations will be discussed. These irradiations include UO 2 clad in an FeCrAl alloy, an alternative uranium silicide compound U 3 Si 5 also clad in a FeCrAl alloy, and an interaction test between UO 2 and different FeCrAl alloys.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Fatigue Testing and Characterization of Pre-hydrided Zircaloy-4 Cladding Tubes

The solubility of hydrogen in Zircaloy-4 (Zry-4) at the cladding operating temperature is near 100 wt. ppm. Above this solubility limit, excess hydrogen precipitates as δ-hydride platelets in the cladding material. Because of the effect of the thermal gradient across the cladding thickness on the migration and precipitation of hydrogen, hydride rims are often observed at the cladding outer surface; under excessive corrosion conditions hydride blisters are possible. It is well known that the fatigue of metal alloys, especially high cycle fatigue, is sensitive to the status of surface including the microstructure, roughness, and residual stress. Thus, the question considered herein, is whether excessive hydrogen pickup modifies the microstructure such that it has a degradation effect on the fatigue performance of cladding during operation. This report describes the evaluation of fatigue performance of a pre-hydrided Zry-4 cladding. A commercial Zry-4 was polished and pre-hydrided to 800 ppm and 1300 ppm H contents. The fatigue testing was conducted under strain control at 5 Hz with fully-reversed bending by using a cyclic integrated reversible bending fatigue tester (CIRFT). Six specimens with 1300 ppm and one specimen with 800 ppm were tested. Significant variation in test results were observed. While two of the 1300 ppm specimens failed with less than 1 ×10 5 cycles (at 0.32% and 0.41%), the four other samples did not fail over the strain amplitude range of 0.25% to 0.38%. Interestingly, the fracture initiation site of the failed samples was on the outside diameter (OD) surface rather than on the inside diameter (ID) surface as is typical for the as-polished cladding. This suggested a degrading effect of the hydriding process. Subsequently, three of the unfailed specimens were then further tested at ~0.43% to observe where failure initiation occurred. Significant variation was also observed in these three specimens. One specimen failed at ~9000 cycles while the two others failed at 43000 and 1.06 ×10 5 cycles. The performance here correlated to the location of failure initiation; failure initiated on the OD in the sample that failed after 9000 cycles while it initiated on the ID in the sample that failed after 43000 and 1.06 ×10 5 cycles. Flat features at the OD initiation sites suggest a brittle hydride feature on the surface of those samples was the cause of the degradation in fatigue performance, though the overall hydrogen levels in all samples was similar. A summary of all the observations is provided below: • The un-failed specimens with 1300 ppm H were cycled to failure with a higher amplitude near 0.43%. In addition, the fatigue-treated specimen tended to have a longer fatigue at the same induced amplitude. • Fractography revealed a mixed failure mode for the pre-hydrided specimen. Particularly, the specimen with fracture initiation site (FIS) located on the outer diameter surface of tube tended to have a shorter fatigue life than that of FIS on the inner diameter surface. • Etched cross section was shown to have hydride platelets aligned with tube longitudinal axis as expected, and the density of hydrides is at the similar level as in literature data. Meanwhile, a LECO procedure was applied for hydrogen concentration measurement, which showed the measured hydrogen contents are close to the nominal value. • With the polished cladding tube as baseline, O’Donnell-Lager (O-L) analysis showed that a decrease of about 50% in reduction-of-area (RA) would be needed for the O-L fitting to the fatigue data of 1300 ppm cladding. The suggestion for the next steps is provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

A Critical Review on the History of Fabricating Monolithic U-Mo Fuel Plates

Fabrication of uranium-molybdenum alloy fuels has been occurring since the early 2000s in support of the development of a high-uranium-density low-enrichment fuel for use in high-performance research and test reactors which operate at relatively low temperatures. The primary fuel form—a thin foil of uranium, alloyed with 10wt% molybdenum, which is coated in a layer of zirconium and then encapsulated with aluminum 6061 as a cladding material—was developed over a number of years.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1977

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

Clement, J. D.↗
DOE OSTI2019

Probability of Pinching Failure in PWR SNF after Long-Term Storage

A probabilistic assessment was performed on pinching failure of PWR SNF rod cladding under accidental SNF cask transport conditions. a set of 3,000 rod pinching scenarios was analyzed, with different geometrical, material, and loading parameters, using a Latin Hypercube Sampling (LHS). The investigated accidents include train derailments leading to cask collisions or cask drops. Pinching failure was related to cladding offset strain, defined as the plastic deformation interval in radial cladding direction normalized with respect to the outer cladding diameter. Finite element (FE) models were used to compute SNF rod deformations and the offset strain demand OSD in each of the assessed pinching scenarios. The numerical models account for the contribution of the degraded fuel pellets to rod pinching stiffness. The offset strain capacity OSC was computed using regression and interpolation models based on ring compression (RCT) test results, and account for hydride-related effects on cladding ductility. The results show a low probability of pinching failure. Hydride-related cladding embrittlement affects the structural integrity of PWR SNF rods after long-term dry storage of up to 300 years. Radial hydride patterns or large amounts of circumferentially oriented hydrides reduce the cladding ductile deformation capability when pinched at low material temperatures. The probability of hydride-related pinching failure of PWR cladding is estimated after long-term dry storage to address: How do PWR SNF rod conditions like hydrogen content, vacuum drying peak cladding hoop stress σ{sub θ}, and cladding material temperature affect cladding ductility? Is hydride-related cladding embrittlement a material degradation mechanism that affects the structural integrity of PWR SNF rods under pinching loading? What are the controlling parameters for deformation demands in PWR SNF rods under pinching loading? Is the pinching load amplitude an important parameter that controls the probability of pinching failure? How do parameters like pellet-cladding gap or fuel pellet degradation affect the pinching loading capacity of PWR SNF rods? Statistical modelling of PWR SNF rod conditions in U.S. inventory: A pseudo rod database was generated based on EIA database. Analysis of train accident scenarios: a probability density function (pdf) of a pinching load distribution P{sub 9m} of a 9-m cask drop is linearly scaled to generate a more representative: distribution P{sub AS}. The new distribution considers train collision velocities, cask drops for on- and off-bridge accidents, and cask impact orientation. LHS generates 3,000 rod pinching scenarios based on nine independent geometrical, material, and loading parameters. Probabilistic FE Analysis: The LHS realizations are used to generate 3,000 FE models. The FE analysis is conducted for P{sub 9m}, and P{sub AS}. The workflow is automatized. The OSC is computed with regression and interpolation models, as a function of the hydrogen content C{sub H}, vacuum drying peak σ{sub θ}, and SNF temperature T{sub FR}: The regression data was obtained in RCTs on PWR cladding. The OSD is derived from the cladding deformations in the FE models. Cladding failure was identified via a comparison of OSC and OSD. A 300 and a 30 years-of-dry-storage duration is investigated: The dry-storage period affects the expected SNF temperature and consequently, the OSC and the OSD. The material temperature LHS population is adapted to account for a SNF temperature variation. The probability of hydride-related, ductile cladding failure during a severe transportation accident is not significantly affected by the pinching load amplitude: The probability is slightly reduced from a range of less than 0.0333% to 0.1667% for P{sub 9m}, to a range of less than 0.0333% to 0.1333% for P{sub AS}, depending on the vacuum drying peak cladding hoop stress modelling assumptions. Further, the SNF dry-storage duration does not significantly affect the probability of ductile pinching failure. Based on available train accidents data, the pinching failure probability for a 5,000 km long transportation route ranges from less than 1.62 E-7 to a value of 6.47 E-7, depending on the vacuum drying peak cladding hoop stress modelling assumptions. The pinching load amplitude is not a controlling parameter for the probability of ductile cladding pinching failure, and does not control the cladding deformations once pellet-cladding contact is established, The pellet-cladding gap width, however, plays a key role. Fuel pellet degradation only affects the deformation capacity, if the pellet stiffness is significantly reduced (more than 90%). Only highly embrittled cladding is likely to fail. The deformation demands on the cladding under pinching loading are small.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Validation of advanced in-pin metallic fuel models of SAS4A

The SAS4A safety analysis code has been extended to include mechanistic and physics-based models of U-Pu-Zr and U-Zr metallic fuel pins. The predictions of the code for various phenomena such as fission gas behavior, clad wastage formation, thermal expansion, creep, and gas swelling induced axial fuel relocation, in-pin molten fuel ejection, and clad failure models have been significantly improved. The paper presents the validation effort of the new code using the results of eight sets of TREAT MSeries overpower experiments, prior to clad failure. Fuel axial expansion reactivity and clad failure behavior was benchmarked. The results show that the code predictions are satisfactory. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

BISON Validation for LOCA and PCMI Behaviour Using Measurements From the Halden Reactor Project

BISON is a finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 1.5D and 2D axisymmetric, or 3D geometries. Efforts are underway to validate BISON for application to Light Water Reactor fuels using, to a large extent, experimental data from the Halden Reactor Project. Recent validation activities have included both Loss of Coolant Accident (LOCA) and Pellet Clad Mechanical Interaction (PCMI) experiments. Two such validation cases, the IFA-650.10 LOCA test and the IFA-629.4 power ramp test (PCMI), are summarized in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Microstructural characterization of as-fabricated monolithic plates with boron carbide, aluminum boride, and zirconium boride burnable absorbers

The use of burnable absorbers can be beneficial for nuclear reactors by extending the fuel’s operational cycle, providing additional criticality control, and flattening the power profile. In this work, three burnable absorber materials (boron carbide, aluminum boride, and zirconium boride) embedded in aluminum have been fabricated into foils and clad in AA-6061 for potential use in high performance research reactors. The as-fabricated boron-containing phases were determined using transmission electron microscopy to be AlB2, B4C, and ZrB2. TEM also revealed weak bonding at the B4C-matrix interface. SEM showed a relatively uniform spatial distribution of boron-containing phases for all the candidate materials. Higher porosity was observed in the foil containing ZrB2 in its as-rolled condition. The porosity in the ZrB2 foil was reduced by hot isostatic pressing. The size and shape distributions of the boron-containing phases were analyzed on the criteria of cross-sectional area, perimeter, roundness, circularity, and aspect ratio. A method of converting the 2D burnable absorber dispersoids seen in cross-sectional microscopy images into 3D volumes was derived using both spherical and ellipsoidal geometry models. The difference in calculated burnable absorber dispersoid average volume between the two models ranges from 20% to 100%, which could impact burnable absorber burnout rates due to differences in neutron self-shielding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

White Paper Summary of 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding

This white paper recommends that ASTM International develop standards to address the potential impact of hydrides on the long term performance of irradiated zirconium alloys. The need for such standards was apparent during the 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding and Assembly Components, sponsored by ASTM International Committee C26.13 and held on June 10-12, 2014, in Jackson, Wyoming. The potentially adverse impacts of hydrogen and hydrides on the long term performance of irradiated zirconium-alloy cladding on used fuel were shown to depend on multiple factors such as alloy chemistry and processing, irradiation and post irradiation history, residual and applied stresses and stress states, and the service environment. These factors determine the hydrogen content and hydride morphology in the alloy, which, in turn, influence the response of the alloy to the thermo-mechanical conditions imposed (and anticipated) during storage, transport and disposal of used nuclear fuel. Workshop presentations and discussions showed that although hydrogen/hydride induced degradation of zirconium alloys may be of concern, the potential for occurrence and the extent of anticipated degradation vary throughout the nuclear industry because of the variations in hydrogen content, hydride morphology, alloy chemistry and irradiation conditions. The tools and techniques used to characterize hydrides and hydride morphologies and their impacts on material performance also vary. Such variations make site-to-site comparisons of test results and observations difficult. There is no consensus that a single material or system characteristic (e.g., reactor type, burnup, hydrogen content, end-of life stress, alloy type, drying temperature, etc.) is an effective predictor of material response during long term storage or of performance after long term storage. Multi-variable correlations made for one alloy may not represent the behavior of another alloy exposed to identical conditions and the material responses to thermo-mechanical exposures will be different depending on the materials and systems used. The discussions at the workshop showed several gaps in the standardization of processes and techniques necessary to assess the long term performance of irradiated zirconium alloy cladding during dry storage and transport. The development of, and adherence to, standards to help bridge these gaps will strengthen the technical basis for long term storage and post-storage operations, provide consistency across the nuclear industry, maximize the value of most observations, and enhance the understanding of behavioral differences among alloys. The need for, and potential benefits of, developing the recommended standards are illustrated in the various sections of this report.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Irradiation performance of a U-7Mo in Al-Si matrix dispersion full-size fuel plate assembly

The Korea Atomic Energy Research Institute (KAERI) is leading the Ki-Jang Research Reactor (KJRR) project with the intent to develop a new reactor for medical isotope production and other nuclear research purposes. The KJRR core is designed to use high density fuel system where uranium alloyed with 7 wt% molybdenum (U-7Mo) particles are dispersed in a matrix of aluminum alloyed with 5 wt% silicon (Al-5Si) and clad in aluminum alloy 6061 (Al-6061) to form fuel plates. KAERI developed a fabrication facility to construct KJRR fuel assemblies and partnered with the Idaho National Laboratory (INL) to irradiate a full-size fuel assembly, with 21 total fuel plates, in the Advanced Test Reactor (ATR). Irradiation testing and subsequent Post Irradiation Exam (PIE) campaigns were performed successfully over a multi-year project. Monte Carlo neutronic calculations, coupling with a depletion code, were performed based on ATR’s as-run power history which showed that the highest power plate (plate 20) reached 83.1 % end-of-life (EOL) local burnup based on initial 235 U content. Additionally, finite element thermal modeling was performed based as-run power history which showed a beginning-of-life (BOL) peak local heat flux of 184 W/cm 2 . No anomalous fuel performance was observed during the irradiation and target test conditions were achieved. PIE showed favorable performance of the fuel assembly regarding all important phenomena. This paper describes the KJRR fuel assembly irradiation conditions and PIE data to support the conclusion that it performed well, without evidence of unexpected or problematic fuel performance, within an irradiation test designed to bound the KJRR design environment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Pitch Variation Experiments in Water-Moderated Square-Pitched U(6.9)O 2 Fuel Rod Lattices with Fuel to Water Volume Ratios Spanning 0.08 to 0.67

The US Department of Energy (DOE) Nuclear Energy Research Initiative funded the design and construction of the Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratories. The start-up of the experiment facility and the execution of the experiments described here were funded by the DOE Nuclear Criticality Safety Program. The 7uPCX is designed to investigate critical systems with fuel for light water reactors in the enrichment range above 5% 235 U. The 7uPCX assembly is a water-moderated and -reflected array of aluminum-clad square-pitched U(6.90%)O 2 fuel rods. Other critical experiments performed in the 7uPCX assembly are documented in LEU-COMP-THERM-078, LEU-COMP-THERM-080, LEU-COMPTHERM- 096, LEU-COMP-THERM-097, and LEU-COMP-THERM-101. The twenty-seven critical experiments in this series were performed in 2020 in the SCX at the Sandia Pulsed Reactor Facility. The experiments are grouped by fuel rod pitch. Case 1 is a base case with a pitch of 0.8001 cm and no water holes in the array. Cases 2 through 6 have the same pitch as Case 1 but contain various configurations with water holes, providing slight variations in the fuel-to-water ratio. Similarly, Case 7 is a base case with a pitch of 0.854964 cm and no water holes in the array. Cases 8 through 11 have the same pitch as Case 7 but contain various configurations with water holes. Cases 12 through 15 have a pitch of 1.131512 cm and differ according to the number of water holes in the array, with Case 12 having no water holes. Cases 16 through 19 have a pitch of 1.209102 cm and differ according to number of water holes in the array, with Case 16 having no water holes. Cases 20 through 23 have a pitch of 1.6002 cm and differ according to number of water holes in the array, with Case 20 having no water holes. Cases 24 through 27 have a pitch of 1.709928 cm and differ according to number of water holes in the array, with Case 24 having no water holes. As the experiment case number increases, the fuel-to-water volume ratio decreases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Risk-Informed Safety Analysis for Accident Tolerant Fuels

Accident Tolerant Fuels (ATF) are being tested by different nuclear vendors and research organization and their introduction in light water reactors fleet is planned for the second half of the 2020`s. In the framework of the US-DOE Light-Water Reactor Sustainability program, Risk-Informed Safety Analysis pathway (LWRS-RISA), research activities are being conducted at the Idaho National Laboratory (INL) for developing tools and methods that can help the industry in quantifying the ATF introduction benefits. In this paper we describe the developed risk-informed methodology, the codes improvements and we present some results for selected accidental conditions. The developed methodology combines INL state-of-the-art deterministic Best Estimate tools like RELAP5-3D code, and Probabilistic Risk Analysis tools like RAVEN and SAPHIRE codes. The analyses are performed on a three-loops pressurized water reactor (PWR) simulating accidental conditions like Station Blackout and LB-LOCA and considering near-term ATF (FeCrAl and Chromium-coated clads). Finally, we show, through our methodologies, how the delta-Core Damage Frequency can be assessed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

UNCERTAINTIES FOR IN-REACTOR TESTING OF TRITIUM LOADED GETTERS WITH TEMPERATURE CONTROL

The fourth experiment in TPBAR Materials Irradiation Separate-Effects Test (TMIST) series (TMIST-4) is planned to experimentally determine the effect of neutron irradiation on the equilibrium vapor pressure of tritium over a tritium loaded Nickel-Plated Zirconium (NPZ) getter in a simulated TPBAR service environment. This pressure will be used to estimate the tritium permeation rate. In the thermal design in the TMIST-4, two evaluations are required. 1) The cladding temperature is calculated to estimate the feasibility of detecting potential irradiation enhancement of the partial pressure for an NPZ getter. This work is required to obtain a precise cladding temperature for measurement of a low tritium partial pressure. This work has been completed and indicated that the required/target cladding temperatures based on the TMIST-4 test plan can be satisfied under the test condition range by modifying the capsule component shapes and inserting a thermally conductive metal sleeve. 2) The evaluation of uncertainties associated with the test design and operation is performed. This includes fabrication tolerances, and power and nuclear operating conditions. These factors will impact the component temperatures inside capsules and resulting the tritium partial pressure. The tritium permeation rate is a function of temperature and tritium partial pressure. The prediction of tritium permeation rates affects the feasibility of obtaining measurement resolutions required to reach reasonable conclusion for potential irradiation enhanced performance. This work will investigate the uncertainties.

TTP, TPBAR, TMIST-4, TPBAR Materials Irradiation S↗
DOE OSTI2015

Rail Shock and Vibration Pre-Test Modeling of a Used Nuclear Fuel Assembly

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel (UNF) and high-level radioactive waste (HLW). The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel and HLW generated by existing and future nuclear fuel cycles. The Storage and Transportation staff within the UFDC is responsible for addressing issues regarding the long-term or extended storage (ES) of UNF and its subsequent transportation. Available information is not sufficient to determine the ability of ES UNF, including high-burnup fuel, to withstand shock and vibration forces that could occur when the UNF is shipped by rail from nuclear power plant sites to a storage or disposal facility. There are three major gaps in the available information – 1) the forces that UNF assemblies would be subjected to when transported by rail, 2) the mechanical characteristics of fuel rod cladding, which is an essential structure for controlling the geometry of the UNF, a safety related feature, and 3) modeling methodologies to evaluate multiple possible degradation or damage mechanisms over the UNF lifetime. In order to address the first gap, options for tests to determine the physical response of surrogate UNF assemblies subjected to shock and vibration forces that are expected to be experienced during normal conditions of transportation (NCT) by rail must be identified and evaluated. The objective of the rail shock and vibration tests is to obtain data that will help researchers understand the mechanical loads that ES UNF assemblies would be subjected to under normal conditions of transportation and to fortify the computer modeling that will be necessary to evaluate the impact those loads may have on the integrity of the UNF assembly. The shock and vibration testing along with computer modeling is a vital part of research to achieve closure of a gap in information related to the ability of ES UNF to maintain its safety function when subjected to NCT. In support of this effort, preliminary structural dynamics modeling is presented herein. The modeling investigates the rigidity of a hypothetical cask and cradle structure by comparing it to a monolithic concrete mass. The concrete mass represents a practical option for achieving the necessary cask and cradle mass on a flatbed railcar, but this comparative modeling study investigates whether or not the dynamic loads transmitted through a monolithic concrete configuration are adequately representative of a realistic cask and cradle system. This modeling highlights the need for rail testing by reporting the phenomenon of structural transmissibility. As shown herein, this structural transmissibility can cause an amplification of shock and vibration loads through the structure, which could potentially lead to accelerated mechanical degradation of UNF under NCT.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Special nuclear material simulation device

An apparatus for simulating special nuclear material is provided. The apparatus typically contains a small quantity of special nuclear material (SNM) in a configuration that simulates a much larger quantity of SNM. Generally the apparatus includes a spherical shell that is formed from an alloy containing a small quantity of highly enriched uranium. Also typically provided is a core of depleted uranium. A spacer, typically aluminum, may be used to separate the depleted uranium from the shell of uranium alloy. A cladding, typically made of titanium, is provided to seal the source. Methods are provided to simulate SNM for testing radiation monitoring portals. Typically the methods use at least one primary SNM spectral line and exclude at least one secondary SNM spectral line.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗