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36 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Benchmark Specifications for TREAT Tests M5, M6, and M7

Detailed information describing three TREAT tests performed on metallic fuels has been collected and organized for use as a benchmark in evaluating the performance of metallic fuel models and codes. The tests, designated M5, M6, and M7, subjected EBR-II-irradiated fuel pins to a single type of overpower transient (at prototypical conditions with full coolant flow and an exponential power rise on an 8 second period) in flowing sodium loops. Six fuel pins were tested; five were ternary (U-19Pu-10Zr) alloy fuel clad in D9 with burnup ranging from 0.8 to 9.8 at. %, and one was binary alloy fuel (U-10Zr) clad in HT9 with 2.9 at. % burnup. The information gathered from the test records is expected to be useful for pre-transient characterization of the irradiated fuel pins as well as the transient analysis of the metallic fuel when subjected to severe accident conditions. This report presents benchmark specifications for the M5, M6, and M7 TREAT tests, identifies where primary sources of benchmark-related information can be found, and includes background information to help a user of the data understand their applicability and limitations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Material Characterization Report of EBR II Mark IV Fuel Element

The focus of this report is the characterization of an unirradiated as-built fuel element from the Experimental Breeder Reactor II (EBR-II) Mark IV driver fuel design. The fuel in this element is a metallic uranium-zirconium alloy, and the cladding material is HT9 stainless steel. Both the metallic fuel and cladding materials are characterized, including chemical composition, grain size, morphology, phase composition, precipitate composition, crystallography, density, and hardness.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Experimental Investigation of FCCI Using Diffusion Couple Test Between UZr Fuel with Sb Additive and Cladding

Alloying additions are introduced into U-Zr fuel in order to bind lanthanides (e.g., cerium) and prevent their migration to the fuel-cladding interface. Antimony (Sb) is being investigated as a candidate additive. The present study focuses on the diffusion couple behavior of U-10Zr (wt%) alloy with Sb against cladding (iron or HT9) at 640°C. The diffusion cross sections were analyzed using a scanning electron microscope and X-ray diffraction. Zr-rind was found at the interface of the fuel alloy, Sb was found to be bound in Sb-Zr precipitates or Sb-Ce precipitates, and no reaction was found between Sb precipitates and the cladding materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Experimental assessment of antimony (Sb) in pure uranium for immobilizing fission product lanthanides

The use of fuel additives is one of the concepts to mitigate fuel cladding chemical interaction (FCCI) for metallic fuel because fission product lanthanides are expected to be immobilized by the additive. Antimony (Sb) has been discovered to be a good candidate in UZr fuel. The present study focuses on its mechanism for immobilization in pure uranium, U–4Sb alloy was fabricated to understand the Sb behavior, while U–4Sb–4Ce was fabricated to simulate the case when lanthanides are generated. Both of the as-cast and annealed samples were characterized by scanning electron microscope (SEM) and energy dispersive spectrometer (EDS). U–Sb precipitates are formed in U–4Sb alloy, while U–Sb and Ce–Sb were found in U–4Sb–4Ce alloy, thermal exposure does not change the Sb-precipitation morphologies or chemical composition. Furthermore, diffusion couple tests between those alloys and cladding materials (Fe or HT9) under 650 °C for 500 h were performed and analyzed using SEM/EDS. Diffusion couple tests demonstrate the reaction between cladding and uranium, while no reaction between Sb-precipitations and cladding materials was found.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Benchmark Exercise for the Control Rod Swelling Evaluation

The VTR core has six reactivity control assemblies and three safety assemblies. The control assemblies or primary control rods are adjusted during the normal operation to balance the core reactivity and to control the reactor power. A typical control assembly radial layout is presented in Figure 1. The figure shows the swelled absorber (B 4 C) rod. Initially, helium gas fills the gap between the pin and the cladding before irradiation swelling takes place. For VTR, HT9 steel was selected as the cladding and duct material. The main neutron absorbing material used in the VTR is B 4 C. When residing in the core, the neutronics, thermophysical, and mechanical properties of the materials used in a control assembly will degrade due to accumulated neutron damage. Material degradation limits how long a control assembly can reside in the core. Many phenomena affect the control assembly lifetime, such as the loss of reactivity worth due to B 4 C depletion, the mechanical interaction of the absorber rod and the cladding due to B 4 C swelling, the helium gas buildup in the pin due to B-10 capture, etc. B 4 C swelling, which causes closure of the gap between the absorber rod and the cladding, is usually considered as the main limiting factor from past experience. An initial study was conducted at PNNL to evaluate the irradiation behavior of a VTR control assembly. The evaluation was performed using the CNRD2 code that was initially developed for the FFTF. The study also included an assessment of the VTR control assembly and focused on a 61-pin control assembly design, which is different from that used (37-pin design) in the core design study. The study conducted by PNNL was reviewed independently by ANL. A Python script referred to as the Control Assembly Evaluation Script (CAES) was developed for the independent review and additional assessment of 37-pin control assembly design. The script has focused on the assessment of the absorber rod swelling for its importance in determining the control assembly lifetime. CAES uses geometry, neutronics, materials data as input to predict the swelling of the absorber rod during its residence in the reactor core. The results from CAES showed some non-negligible differences against the PNNL results. Some of the differences can be attributed to the different interpretation of the control rod assembly dimensions. To resolve this issue, a benchmark exercise was proposed. The benchmark specification was developed by PNNL. The benchmark exercise was performed independently at PNNL and ANL using different codes/scripts (CRND2 and CAES). This memo documents the results calculated using the different codes. However, this report is limited to presenting the results obtained. Further investigation of the cause of the observed difference will be performed as part of future activities, pending continuation of the VTR program.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

HT9 swelling in high burnup fast reactor fuel pin components

The irradiation-induced void swelling of HT9 stainless steel has been observed many times, but there are little data available to quantify the amount of swelling expected as a function of neutron exposure and temperature. This is further complicated by potential heat-to-heat chemical variations or heat treatment conditions during fabrication. Herein, this paper summarizes the results of most of the relevant previously completed studies, then focuses on the swelling behavior of HT9 mixed-oxide fuel cladding and wrapper wire irradiated to high neutron exposures. Note that the hexagonal duct used for the ACO-3 experiment (in the Fast Flux Test Facility, or FFTF) was also made from HT9 and has been studied previously by other researchers. The cladding enables a wider range of operating temperatures. Immersion density and transmission electron microscopy were used to identify a peak in the amount of swelling located axially, just below core centerline, for both the cladding and the wire. The peak roughly coincided with a peak seen in the cladding diameter profile. The wrapper wire showed enhanced swelling compared to the cladding, likely due to over-tempering during wire fabrication heat treatment. The grains showed decreased martensitic structure were larger in size. These data do not provide for the means to fully characterize the swelling of HT9, but do provide further insight into the factors controlling it. The onset of breakaway swelling was not indicated in this data.

36 MATERIALS SCIENCE↗
DOE OSTI2022

BISON fuel performance modeling optimization for experiment X447 and X447A using axial swelling and cladding strain measurements

With the recent need to qualify new reactor designs such as the Versatile Test Reactor (VTR), fuel performance calculations need to be performed to determine safety criteria of the proposed designs. In order to validate the fuel performance results obtained by a fuel performance code, BISON, for new reactor designs, legacy fuel from EBR-II and FFTF MFF with Post -Irradiation Examination (PIE) data need to be used as validation cases to benchmark models. Here in this work, BISON has been paired with the Fuels Irradiation & Physics Database (FIPD) and IFR Materials Information System (IMIS) to supply PIE data for comparison with simulations of EBR-II experiments X447/X447A. X447/X447A were assessed by implementing models for Fuel Cladding Chemical Interaction (FCCI) within BISON and optimizing the friction coefficient between the fuel surface and the cladding, the anisotropic swelling factor, and the HT9 first thermal creep scalar (which scales the first term in the HT9 creep equation) to best match the PIE axial fuel swelling height and cladding profilometry for all pins in X447/X447A. The optimal values were found using a generic algorithm developed to select different values for the three parameters until end criteria was met and error couldn’t be reduced further. The BISON-simulated cladding profilometry was evaluated using Standard Error of the Estimate (SEE) to account for the profile shape of the cladding profilometry. Optimal values for the friction coefficient, anisotropic fuel swelling factor, and HT9 first thermal creep scalar were found to best fit the BISON simulation results to the PIE measurements found in IMIS and FIPD. Improvements to current models are suggested to account for the underprediction of fuel swelling at low burnups and the overprediction of fuel swelling at higher burnups observed for the axial fuel swelling height. Although two pins in EBR-II X447/X447A (DP70 and DP75) were known to fail due to FCCI, none of the pins simulated in BISON reached a cumulative damage fraction (CDF) above 0.008 with FCCI correlations coupled in the BISON simulations. The error estimate generated for all pins in X447/X447A using optimal values was 209 µm, which is deemed acceptable.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

The influence of nitrogen and nitrides on the structure and properties of proton irradiated ferritic/martensitic steel

The 12Cr1MoWV (wt%) ferritic/martensitic steel HT9 is a candidate material for fuel cladding in advanced nuclear reactors, such as the Versatile Test Reactor currently under development. As such, understanding the relationship between microstructure and mechanical properties in the context of irradiation environments for these steels is critical. N content, and more specifically interstitial N, has been hypothesized to be detrimental to irradiated properties at lower temperatures (less than 0.3T m ) to a total of 6 dpa; however, in this work at a dose of 1 dpa the irradiated microstructure was improved with added N, leading to less irradiation hardening. Three variants of HT9 were irradiated with 1.5 MeV protons to a dose of 1 dpa at 300°C. The HT9 variants included Low (10 ppm), Mid (190 ppm), and High (440 ppm) N alloys that were otherwise nearly identical. Changing the N content had a variety of effects on the irradiated defect structures. As N content increased, the average dislocation loop diameter decreased, while the number density of loops increased. Additionally, extensive Ni clustering was observed on dislocations and interfaces. The Mid and High N specimens exhibited significantly less hardening (ΔHV≃100) relative to the Low N specimen (ΔHV≃160). The decrease in hardening is attributed to vanadium carbonitride acting as a sink for Ni clusters that would otherwise form on dislocations. Under the irradiation conditions used, these results suggest increasing the N content in HT9 may have a desirable effect on the irradiated structure and properties at the dose studied, as well as the swelling resistance at higher doses. In other words, N content appears to be a powerful tool for tailoring the self-interstitial atom cluster mobility in F/M steels for different temperature and dose applications.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Multi-principal element alloys for fast reactor cladding applications

Given the extensive list of multi-principal element alloys (MPEAs) within literature and the overwhelming number of alloys that can be made from only a handful of elements, this article proposes a methodology for prioritizing alloys for use as cladding within advanced reactors. This paper applies neutronic, mechanical, and chemical assessments to a collection of MPEAs available within literature for use as advanced reactor cladding. The results are compared to a reference design of HT9, a ferritic/martensitic steel, for employed in sodium-cooled fast reactor. Mechanical assessments determined pressure limits for a thin-walled tube pressure boundary, thus relating the material's mechanical properties to a minimum acceptable wall thickness. Neutronic analyses reveal a maximum allowable wall thickness that a given material must meet to provide a level of neutronic economy that is equivalent to than that of HT9. Lastly, each alloying element is compared to typical fission products found in a fast reactor fuels to mitigate deleterious phenomena such as fuel-cladding chemical interactions (FCCI). These analyses indicate that Mo-Nb-Ti-V-based alloys are likely to be advantageous and that the inclusion of elements with a high neutronic penalty (e.g., Hf) could be considered with minimal consequences.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Reduced Diameter and Lined Cladding Production Status

A contract has been placed with Century Tube Inc. to produce cladding for the FAST experiment. Excess EBR II cladding tubes will be cold drawn down to 0.115 in. diameter × 0.009 in. wall and 0.077 in. × 0.006 in. wall. Lined cladding will also be tested in the FAST experiment. Century Tube will produce several samples, varying assembly the assembly process, with solid zirconium or vanadium rods inside of the HT9. Cladding/liner bonding of each will be evaluated using microscopy. The best one will be further processed to form a liner tube within the cladding. A shrink fit approach is also being developed, where a solid zirconium or vanadium bar is shrunk fit into a cladding piece. This process has shown favorable results and will be compared against the co drawn samples when available.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Validation of MFUEL Metal Fuel Performance Models of SAS4A/SASSYS-1

The fuel characterization models of SAS4A/SASSYS-1 (SAS) have recently been extended to include a new U-Pu-Zr metal fuel model, MFUEL. MFUEL is equipped with mechanistic physics based models to predict the pre-transient characterization and transient response of metal fuel, with emphasis on fuel melting, cladding failure, and the metal fuel’s impact on core reactivity. The MFUEL model will be available in the full version of SAS4A/SASSYS-1 5.7, which is scheduled to be released in June 2023. Fast reactor fuel pins that operated in EBR-II and FFTF with low smear density U-Zr and U-Pu-Zr metal fuels and irradiation resistant ferritic-martensitic cladding showed significant advantages in achieving high burnups and assuring inherent safety characteristics during anticipated transients, design basis events, and beyond design basis events. For safety analysis, a fuel performance model must be able to predict (1) Fuel pin mechanics and compositional and dimensional changes, (2) Clad failure, and (3) Fuel pin thermal resistance. Achieving these high level goals accurately is strongly related to the model performance of individual physical processes taking place within a fuel pin during its lifetime. Metal fuels typically operate above the mid-point temperature of melting during normal, as well as off-normal, conditions. At these elevated temperatures, the availability of thermal activation provides a driving force for various diffusional processes leading to complex phase transformations, micro-structure evolution, significant amounts of fuel swelling, interconnected porosity formation, excessive amounts of fission gas release, and fuel clad chemical interactions. Clad failure in fast reactors primarily occurs as a result of creep rupture augmented by clad wastage formation. The reaction is driven by thermal creep induced dislocation motion, grain boundary cavity nucleation, growth and breakup of grain boundaries. The high level complexity and limited available data requires introducing physics-based modeling approaches to gain extrapolation ability and sensitivity with respect to various conditions. The objective of this report is to perform validation of MFUEL using the experimental data for (1) Normal operation EBR-II fuel behavior, (2) Normal operation PHENIX fuel behavior, (3) HT9 Pressure tube ramp-and-hold creep rupture tests, (4) Whole Pin Furnace (WPF) creep strain, creep rupture and eutectic tests, (5) Fuel Behavior Test Apparatus (FBTA) eutectic tests, and (6) TREAT M5-7 OverPower tests up to clad failure. Section-2 includes a brief description of the MFUEL models. A detailed description of the MFUEL physics-based, semi-empirical models will be presented in the SAS V 5.7 theory manual. Section-3, 4, and 5 describes the validation effort for the pre-transient irradiation, furnace transients, and TREAT M-Series transients, respectively.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Evaluation of Reduced Order Model for HT-9 Creep and Modifications to Current HT-9 Creep Model in BISON

In order to leverage existing reduced order models (ROM) for the modeling of LANL-developed HT-9 cladding constitutive behavior, mechanistic-based models have been implemented into BISON. It is posited that the increased fidelity of the ROM will show improvement over engineering-scale models in terms of comparison against experimental measurements. INL will assist LANL in implementing an HT9 ROM in BISON. Once the HT-9 mechanistic constitutive model can be leveraged in nuclear performance simulations, the numerical results of fast reactor models will be analyzed and compared against the Fuels Irradiation & Physics Database (FIPD) and separate effects tests to attempt a validation process. INL and LANL will work together to perform any necessary improvement that is identified during the implementation and use processes. Successful completion of the milestone will enable the general use of validated mechanistic ROMs for HT-9 cladding in BISON. This report provides a brief introduction to the Los Alamos Reduced Order Model Applied to Nonlinear Constitutive Equations (LAROMANCE) code, mechanical testing used to illustrate its creep and plastic deformation predictions for HT-9 cladding, and a comparison with existing models implemented into BISON. Current models have been modified to more appropriately account for primary thermal creep and results for corresponding testing are included in this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Preliminary HEA Solid Phase Processing Development Report

The development of high temperature fuel cladding materials to withstand a variety of extreme environments have received much attention. Several potential materials systems that have been identified for the fuel systems and core structural materials application in advanced reactor systems are ferritic/martensitic steel (e.g., HT9), austenitic stainless steels (e.g., 316 LN), oxide-dispersion strengthened steels (e.g., 12 YWT), Ni-based alloys and ceramic-based composites depending on the type of the reactors. Though these material systems have promising properties conducive for radiation-resistant performance, they suffer beyond the design-limit from one or more damage processes such as void swelling, radiation embrittlement, phase instability, corrosion, and limited creep life.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Non-Destructive Postirradiation Examination of the AFC-4C Capsule/Rodlet 5 Experiment

The AFC-4C irradiation test was irradiated at the Idaho National Laboratory Advanced Test Reactor to investigate several different fast reactor fuels that could be used to facilitate ultra-high burnup applications in sodium fast reactors. Several different alloys, fuel geometries, bonding materials, and coating/barrier were tested in ferritic-martensitic HT-9 cladding. The AFC-4C capsule 5 was removed from the reactor at around 8.7 at %HM burnup to begin postirradiation examination. This report presents and discusses the non-destructive PIE results of the AFC-4C-capsule / rodlet 5, a U-10Zr solid fuel, sodium bonded with HT9 with internal Cr-coating, and its performance is evaluated against the historical fuel performance of previously irradiated fuel from literature.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Performance of a ferritic/martensitic steel for nuclear reactor applications fabricated using additive manufacturing

Ferritic/martensitic (FM) steels are being targeted for use in a range of advanced reactor concepts as cladding and structural components. FM steels for nuclear reactor applications have historically been produced using traditional methods (e.g., casting and forging), but recently, additive manufacturing processes have become of interest for making FM-based components. Here, the laser-blown-powder additive manufacturing process was used to fabricate an FM steel, HT9, followed by microstructural and mechanical performance evaluations to determine the viability of future use of additive manufacturing for FM-based component fabrication. Results showed that the as-built condition formed a layered structure with alternating layers of δ-ferrite and martensite, which resulted in anisotropic engineering and true-stress, true-strain mechanical performance. Post-build normalizing and tempering treatments alerted the prior austenite grain size and precipitate distributions, and drove the mechanical performance to near-isotropic properties that mimic wrought-processed properties. Overall, the resulting microstructures in all conditions were rationalized in the context of multi-pass welding and the synergies are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2026

On the uncertainties in helium generation predictions for fission and fusion alloys

With ongoing advances in fusion and advanced fission reactors, quantifying irradiation effects in materials is critical. Transmutation-induced helium in cladding and structural materials can drive swelling and embrittlement, thereby reducing these components’ lifespans. Yet most studies ignore the considerable uncertainties in predicting helium generation rates. In this work, we created a code wrapper, F-SCATTER, that automatically performs simulations in FISPACT-II. We used this tool to investigate potential variance in helium generation rate, or He/dpa, calculations based on deviations in alloy composition, irradiating neutron flux spectrum, computational methodology, and nuclear data sources. We used 12 wt% Cr HT9 steel as the reference case and observed a 6.5%–98.3% He/dpa spread based on compositional variation within a single chemical specification, a 1.8%–11.5% He/dpa variation upon the incorporation of a 15% artificial uncertainty in flux at each energy, and a He/dpa difference as high as 231% when using ENDF/B-VIII.0 versus TENDL-2021 data libraries. Similar results were found for other prominent iron-based alloys, including Grade 91, castable nano-structured alloy, and 316H—where additional variations exist based on reactor type (e.g. thermal, fast, or fusion) and alloying elements such as carbon, nitrogen, and nickel. Based on the simulated results, we conclude that a significant part of the heat-to-heat variability in swelling responses of Fe-based alloys can be driven by impurity content in alloy compositions, and, therefore, chemical control should be a key element in supply chain design for advanced nuclear energy systems. Furthermore, we provide critical recommendations on best practices for evaluating and reporting helium production and lattice damage rates when computing predictions with multiphysics programs such as FISPACT-II.

FISPACT-II↗
DOE OSTI2023

Effective Lanthanide Diffusivity through U-Zr Metallic Fuel

Fuel-cladding chemical interaction (FCCI) is one of the main concerns for the performance of U-Zr metallic fuels. Although FCCI results in complex phase transformations and morphological changes in both the fuel and cladding sides of the interface, the formation of a brittle layer in the cladding side of the interface (sometimes referred to as wastage) has one of the most significant impacts on the cladding mechanical integrity. This brittle layer is associated with the formation of intermetallic compounds between the cladding constituents Fe, Cr and the lanthanide fission product species. The formation of this cladding wastage layer has been the focus of recent efforts to develop a mechanistic modeling framework, including production of lanthanides in the fuel, transport through the fuel to the fuel-cladding interface, and phase transformation to form intermetallic phases. To calculate the rate of transport of lanthanides through the fuel, a multi-scale computational approach has been used. Atomistic calculations have been performed to determine the diffusivities of Nd, the most prevalent lanthanide species, through the solid fuel matrix and along pore surfaces. These diffusivities have been used as input parameters for a mesoscale model to determine an effective diffusion coefficient that accounts for the formation of porosity, its interconnection, and infiltration with bond sodium from the fuel-cladding gap. The effective diffusivity has been used in engineering-scale simulations using the BISON fuel performance code, and BISON simulations have been compared to representative validation cases from the EBR-II reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗