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DOE OSTI2024

High-Burnup BWR LOCA Burst Analysis Framework Development and Demonstration

Nuclear power currently contributes approximately 20% of total electricity generation in the United States and more than 10% globally. Given the increasing reliance on nuclear energy to achieve our nation’s goal of reaching net-zero carbon emissions by 2050, there is significant pressure on the existing nuclear industry to extend plant operational licenses and improve efficiency. This is crucial as the existing nuclear fleet serves as a vital bridge until new light water and advanced reactors can be developed and deployed, bolstering the supply of carbon-free energy to meet domestic demands. Operational costs primarily consist of plant operation and maintenance and fuel costs, influenced by materials and reactor core designs. These factors, coupled with heavily subsidized renewable energy markets, create a challenging economic environment for the existing light water reactor fleet, as well as for new build projects. To address these economic challenges, the nuclear industry has developed a strategic blueprint aimed at enhancing nuclear power’s economic sustainability. Past initiatives, such as efforts to eliminate fuel failures by 2010 and reduce operating costs by 30% before 2020, have laid the groundwork. Optimizing core design parameters, including burnup limits and enrichment levels, can lengthen cycles, reduce outages, reduce batch reload batch fractions and spent fuel storage requirements, and lower maintenance and operating expenses, thereby enhancing economic viability. In the United States, boiling water reactors (BWRs) comprise approximately one-third of the fleet, although much of the research and development focus has traditionally been on pressurized water reactors (PWRs). Advances in modeling and simulation, particularly through the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, are crucial to the long-term viability of BWRs, just as they are for PWRs. A key research area of the high burnup/increased enriched fuel initiative is focused on addressing loss-of-coolant-accident (LOCA)-related issues. NEAMS has dedicated significant effort to enhancing tools to better support BWRs, with a current focus on showcasing the BWR framework for high-burnup LOCA analysis. This high-fidelity work will demonstrate a best estimate pin-by-pin high-burnup BWR LOCA analysis to assess full-core cladding rupture behavior. This modeling capability will help with better understanding and realistic evaluation of fuel fragmentation, relocation, and dispersal (FFRD) phenomena at BWRs, which then could be used to prevent FFRD at BWRs without penalizing operational parameters. In addition, the results of this work will help identify strategies to identify additional margins or to potentially limit cladding rupture through core design optimizations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Modeling and Simulation of Fuel Dispersal During the Loss-of-Coolant Accident

This document is the compilation of the milestone portion to a larger end of project NEUP report. The executive summary of the modeling portion is provided below: In the event of cladding rupture during a postulated LOCA in a pressurized water reactor, fuel particles, along with fission gases, can be expelled into the reactor core from the fractured fuel rod, a phenomenon referred to as fuel dispersal. The initial stage of fuel dispersal is strongly influenced by the high-pressure ejection of fuel fragments, the size and geometry of the ruptured cladding, and the depressurization history of the fuel rod during the postulated LOCA transient. Depending on the location of the burst orifice relative to the quench front, the dispersal event represents an intricate three-phase flow and heat transfer phenomenon, where high-temperature fuel particles carried by the fission gases interact with the coolant within the narrow subchannels of the fuel assemblies, inducing localized phase change. Given the unique multiphysics nature of this phenomena, the current study develops a dedicated computational framework to predict the mass distribution and cooling of dispersing fuel particles, facilitating post-accident assessment and management of the fuel assemblies. Considering the scale of nuclear reactor applications, a continuum three-fluid model is proposed for simulating the transport of solids within the reactor core. With high-temperature fuel fragments within the liquid media, nucleation sites inducing phase changes are dispersed within the flow domain. Coupled with the fact that the transient dispersal event occurs on different time scales than other three-phase flow applications, this study derives a time-averaged three-fluid flow model without losing generality. The assumptions regarding the continuum treatment of the solid phase and the modeling of fuel dispersal behavior are incorporated to simplify the governing equations and derive applicable closure relations. The computational validation of the model was conducted using adiabatic experimental results obtained from ongoing research at Oregon State University, focusing on characterizing fuel dispersal behavior during simulated LOCA conditions. Settlement characteristics of the solids, quantified by the probability distribution of equivalent particles, closely matched the probability density functions reported in experimental studies. The transport of fuel particles within a scaled 5 × 5 lattice of a pressurized-water reactor rod bundle geometry was modeled through a two-fluid Eulerian framework. The required boundary conditions were evaluated from the fuel performance code BISON in a postulated large-break LOCA scenario. The modeling framework considered solid fuel particles as granular matter, interacting with the gaseous dry steam phase and fission gases through the governing interfacial momentum exchange between the participating fluids. The simulation results provided the volume fraction of the solids obtained at the bottom surface of the enclosing tank geometry. Postulated LOCA leading to fuel dispersal phenomena involves the strong coupling between fuel thermomechanics, cladding deformation, thermal-hydraulics, and fuel particle transport. Incorporation of such a strong coupling in numerical simulation is performed by coupling the multiphysics solvers. In the case of fuel dispersal, a strong coupled simulation can be performed by coupling the BISON code for fuel performance, the TRACE code for system-level thermal hydraulics, and fuel particle transport in Multiphysics Object-Oriented Simulation Environment (MOOSE). For such intricate infrastructure, the MOOSE Framework eases the data transfer between codes. The recent version of MOOSE has incorporated the Navier-Stokes module for the fluid flow. An exploratory exercise was done to gain familiarity with finite volume capabilities in the MOOSE framework to incorporate the Spalart-Allmaras (SA) turbulence model. New finite-volume and auxiliary kernels were introduced to assemble the SA transport equation, compute turbulent viscosity, and evaluate wall distance and diagnostic turbulence terms, fully integrated with existing Navier-Stokes modules. A turbulent lid-driven cavity at a Reynolds number of approximately 10,000 is used for verification. MOOSE shows the robust solver convergence and produces the turbulent features. But it underpredicts the velocity profile and turbulent quantities, emphasizing the need to develop improved SA near-wall treatments (e.g., low-Re corrections or wall functions) as a key direction for future work.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

The Role of Normal Stress and Shear Stress Heterogeneity in the Inferred Depth-Independence of Stress Drop

Earthquake stress drops are inferred to be independent of their source depth,contradicting standard linear scaling predictions for frictional stick-slip modelsof earthquakes, assuming increasing fault normal stress due to rockoverburden. Here, we examine the scaling between averaged stress drops andincreasing normal stress for simulated earthquakes sequences in continuumrate-and-state fault models. Our models exhibit a power-law-like scaling that isweaker than the linearity predicted by traditional friction models. This resultoccurs as the fault dimension becomes increasingly larger than the earthquakenucleation scale with increasing normal stress. Consequently, the averagedbehavior of ruptures becomes increasingly dominated by conditions for rupturepropagation, reflecting more heterogeneous shear stress conditions. As naturalfaults can be considerably larger than the smallest earthquakes they host, suchweaker scaling between averaged rupture conditions and normal stress maypartially explain the lack of an inferred depth dependence of earthquake stressdrops.

Yang, Minghan↗

DOE OSTI2025

Development of Large Bore Rabbit Capsules in Support of BWR Cladding Irradiations in HFIR

The High Flux Isotope Reactor (HFIR) is an ideal tool for materials irradiation testing because of its intense steady-state neutron flux. Many programs take advantage of HFIR’s central flux trap for irradiation experiments using capsules, also known as rabbits, to support advanced materials development and reactor design. The facility that makes up the HFIR flux trap has recently undergone a design change that increases the HFIR primary coolant volumetric flow rate by removing restrictions in the system. As a result, the usable cross-sectional area within the facility increased, opening the door to increase the cross-sectional area of the rabbit capsules that fill the facility. This report documents a new large-diameter rabbit housing that increases the usable volume within the rabbit capsule by 22.6%. However, challenges arise with increasing the capsule size, such as establishing a new maximum capsule operating pressure and determining the thermal-hydraulic characteristics. This report addresses those challenges with previously adopted HFIR safety methods. The rupture pressure of the rabbit housings is demonstrated while verifying that capsule swelling during and after rupture will not block coolant flow. Then, a safety factor is applied to ascertain an administrative operating pressure. Additionally, the thermal-hydraulic performance of the HFIR facility filled with large-diameter rabbit capsules is shown to not violate previously determined safety criteria. Next, heat transfer coefficients are determined for use in design calculations. Furthermore, this report gives an example of internal configurations for the new, larger rabbit capsules that use relevant boiling water reactor (BWR) cladding geometry. Finally, this report documents an example thermal design performance for a rabbit capsule containing six gauge-curved tensile tube specimens. The thermal performance gives predicted temperature distributions within the capsule and shows the expected temperature of the passive thermometers for post-irradiation temperature comparisons.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2025

Development of Large Bore Rabbit Capsules in Support of BWR Cladding Irradiations in HFIR

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Evaluation of the Mechanical Properties of Cast and Wrought CF8C-Plus Relevant to ASME Code Case Qualification

The mechanical behavior of a cast form of an advanced austenitic stainless steel, CF8C-Plus, is compared with that of its wrought equivalent in terms of both tensile and creep-rupture properties and estimated allowable stress values for pressurized service at temperatures up to about 850°C. A traditional Larson-Miller parametric model is used to analyze the creep-rupture data and to predict long-term lifetimes for comparison of the two alloy types. The cast CF8C-Plus exhibited lower yield and tensile strengths, but higher creep strength compared to its wrought counterpart. Two welding methods, shielded-metal-arc welding (SMAW) and gas-metal-arc welding, met the weld qualification acceptance criteria in ASME BPVC Section IX for the cast CF8C-Plus. However, for the wrought CF8C-Plus, while SMAW and gas-tungsten-arc welding passed the tensile acceptance criteria, they failed the side bend tests due to lack of fusion or weld metal discontinuities.

Chen, Xiang↗

DOE OSTI2024

Assessing the Impact of Molten Halide Salts on Creep of Structural Alloys at 650°-750°C

There is a critical lack of data on the mechanical behavior of candidate structural materials for advanced nuclear reactors under molten halide salt environments. Limited legacy data from the molten salt reactor experiment (MSRE) program showed a significant reduction in creep rupture strength of a Ni-base alloy in molten fluoride salt. With ongoing efforts to commercialize different molten salt reactor concepts, the industry can considerably benefit from quantitative information on the impact of molten halide salts on the engineering properties such as creep and fatigue strength of materials of interest. The present work aims to assess the role of molten salt corrosion on the creep behavior of three alloys 316H, 617 and 282 at 650-816 °C. Creep tests were conducted in fluoride (FLiNaK) and chloride (NaCl-MgCl2) salts. Initial results from the ongoing testing will be presented which suggest that the molten salt environment caused a 25-50% reduction in creep rupture lifetime compared to air exposures. Physics-based corrosion and creep models were employed to gain some insights into the potential degradation mechanisms.

Pillai, Rishi↗

DOE OSTI2024

Development of PWHT-Free, Reduced Activation Creep-Strength Enhanced Bainitic Ferritic Steel for Large-Scale Fusion Reactor Components

A compositional modification has been proposed to validate an alloy design which potentially eliminates the requirement of post-weld heat treatment (PWHT) while preserving the advantage of mechanical properties in a reduced activation bainitic ferritic steel based on Fe-3Cr-3W-0.2V- 0.1Ta-Mn-Si-C, in weight percent, developed at Oak Ridge National Laboratory in 2007. The alloy design includes reducing the hardness in the as-welded condition for improving toughness, while increasing the hardenability for preserving the high-temperature mechanical performance such as creep-rupture resistance in the original steel. To achieve such a design, a composition range with a reduced C content combining with an increased Mn content has been proposed and investigated. Newly proposed “modified” steel successfully achieved an improved impact toughness in the as- welded condition, while the creep-rupture performance across the weldments without PWHT demonstrated ~50% improvement of the creep strength compared to that of the original steel weldment after PWHT. The obtained results strongly support the validity of the proposed alloy design.

Yamamoto, Yukinori↗

DOE OSTI2022

Single-cell RNA-Seq reveals changes in immune landscape in post-traumatic osteoarthritis

Osteoarthritis (OA) is the most common joint disease, affecting over 300 million people world-wide. Accumulating evidence attests to the important roles of the immune system in OA pathogenesis. Understanding the role of various immune cells in joint degeneration or joint repair after injury is vital for improving therapeutic strategies for treating OA. Post-traumatic osteoarthritis (PTOA) develops in ~50% of individuals who have experienced an articular trauma like an anterior cruciate ligament (ACL) rupture. Here, using the high resolution of single-cell RNA sequencing, we delineated the temporal dynamics of immune cell accumulation in the mouse knee joint after ACL rupture. Our study identified multiple immune cell types in the joint including neutrophils, monocytes, macrophages, B cells, T cells, NK cells and dendritic cells. Monocytes and macrophage populations showed the most dramatic changes after injury. Further characterization of monocytes and macrophages reveled 9 major subtypes with unique transcriptomics signatures, including a tissue resident Lyve1 hi Folr2 hi macrophage population and Trem2 hi Fcrls + recruited macrophages, both showing enrichment for phagocytic genes and growth factors such as Igf1 , Pdgfa and Pdgfc. We also identified several genes induced or repressed after ACL injury in a cell type-specific manner. This study provides new insight into PTOA-associated changes in the immune microenvironment and highlights macrophage subtypes that may play a role in joint repair after injury.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2022

Role of Cr Content in Microstructure, Creep, and Oxidation Resistance of Alumina-Forming Austenitic Alloys at 850–900 °C

Creep-rupture properties and oxidation behavior of a series of alumina-forming austenitic (AFA) alloys with variations of Cr contents, based on Fe-(13.5-18)Cr-25Ni-4Al-1.5Nb-0.1C in weight percent, have been evaluated at 850–900 °C. The study investigates material responses in the properties and microstructure through compositional modifications in AFA alloys, targeting performance optimization of alloys under high-temperature, corrosive industrial environments. The creep-rupture life of the alloys at 850 °C and 30MPa monotonically decreased with increasing Cr content, which was correlated with changes in secondary phase volume fractions, such as the reduction in B2-NiAl + Laves-Fe2Nb and increase in Sigma-FeCr with Cr content. The oxidation test at 900 °C in a water-vapor containing environment revealed a range of Cr content from 13.9 to 15.7 wt.%, enabling the formation of stable, protective external alumina scale as well as preventing internal oxidation/nitridation for up to total 7000 h exposure. On the other hand, the alloys with >16.7 wt.% Cr formed Sigma precipitates, which caused a reduction in not only Cr but also Al in the austenite matrix, resulting in less oxidation resistance than other alloys. The findings will guide the further optimization of material performance in the AFA alloy series.

36 MATERIALS SCIENCE↗

DOE OSTI2024

A strainmeter array as the fulcrum of novel observatory sites along the Alto Tiberina Near Fault Observatory

Fault slip is a complex natural phenomenon involving multiple spatiotemporal scales from seconds to days to weeks. To understand the physical and chemical processes responsible for the full fault slip spectrum, a multidisciplinary approach is highly recommended. The Near Fault Observatories (NFOs) aim at providing high-precision and spatiotemporally dense multidisciplinary near-fault data, enabling the generation of new original observations and innovative scientific products. The Alto Tiberina Near Fault Observatory is a permanent monitoring infrastructure established around the Alto Tiberina fault (ATF), a 60 km long low-angle normal fault (mean dip 20°), located along a sector of the Northern Apennines (central Italy) undergoing an extension at a rate of about 3 mm yr –1 . The presence of repeating earthquakes on the ATF and a steep gradient in crustal velocities measured across the ATF by GNSS stations suggest large and deep (5–12 km) portions of the ATF undergoing aseismic creep. Both laboratory and theoretical studies indicate that any given patch of a fault can creep, nucleate slow earthquakes, and host large earthquakes, as also documented in nature for certain ruptures (e.g., Iquique in 2014, Tōhoku in 2011, and Parkfield in 2004). Nonetheless, how a fault patch switches from one mode of slip to another, as well as the interaction between creep, slow slip, and regular earthquakes, is still poorly documented by near-field observation. With the strainmeter array along the Alto Tiberina fault system (STAR) project, we build a series of six geophysical observatory sites consisting of 80–160 m deep vertical boreholes instrumented with strainmeters and seismometers as well as meteorological and GNSS antennas and additional seismometers at the surface. By covering the portions of the ATF that exhibits repeated earthquakes at shallow depth (above 4 km) with these new observatory sites, we aim to collect unique open-access data to answer fundamental questions about the relationship between creep, slow slip, dynamic earthquake rupture, and tectonic faulting.

58 GEOSCIENCES↗

NASA NTRS1966

Histopathology of the Chimpanzee Ear

Four female and two male chimpanzees were secured in various positions on a Daisy Decelerator and subjected to forces ranging from 54 to 180 G. It was found that forces in excess of 54 G may rupture the tympanic membranes and cause subepithelial hemorrhages in the middle ear. The majority of cases showed proteinaceous material, with or without cells in the petrous air spaces. When exposed to forces above 119 G, there was engorgement and often rupture of the pericarotid venous plexus. When supine, distortion of both superior and posterior semicircular canals was found. With forces in excess of 54 G, the cupulae of the cristae ampullaris were either elevated or destroyed. The hair processes were also often broken off. The otolithic membranes, especially of the maculae utriculi were also elevated or otherwise distorted and the saccule was often partially collapsed. In several instances, there was an overabundance of a proteinaceous substance in the lumina of the vestibular apparatus and in the cochlear ducts with their associated scalae. In half the cases, the cochlear duct was narrowed by the depression of the vestibular membrane. Although there seems to be considerable individual variation in ability to withstand these forces, neither age, sex nor weight appear to directly influence the results. The possible sources for the materials found both in the air cells and labyrinth are discussed.

ACCELERATION TOLERANCE↗

NASA NTRS1971

The development and evaluation of a cobalt base oxidation resistant dispersion strengthened alloy

The Co-18Cr-20Ni-4 Vol % ThO2 powders were prepared by a flash drying selective reduction process starting with an aqueous solution of metal salts and colloidal thoria. Powders were consolidated and extruded into rods with a minimum density of 99% of theoretical. Swaging and annealing studies were conducted to determine the conditions that would lead to a product with high stress-rupture strength. The best process yielded a stress-rupture life of 7.2 hours at 10 KSI (69 MN sq m and 2000 F (1094 C). The alloy recrystallized to a duplex (coarse-fine) structure and thus did not exhibit the desired strength of 3000 hours at 15 KSI (103.5MN/sq m and 2000 F (1094 C).

Irani, K. K.↗

NASA NTRS1972

High-temperature strength of refractory-metal wires and consideration for composite applications

Tensile and stress-rupture tests were conducted on wires of W-Hf-C, W-Re-Hf-C, ASTAR 811C (a tantalum alloy), and B-88 (a columbium alloy) at room temperature, 1093 C (2000 F), and 1204 C (2200 F). Metallographic examinations were also made of the wire microstructure after testing. Ultimate tensile strength values up to 2170 meganewtons per square meter (314000 psi) at 1093 C (2000 F) and 1940 meganewtons per square meter (281 000 psi) at 1204 C (2200 F) were obtained for W-Re-Hf-C wire. The best strength values obtained for a 100-hour rupture life were, 1410 meganewtons per square meter (205 000 psi) at 1093 C (2000 F) and 910 meganewtons per square meter (132 000 psi) at 1240 C (2200 F) for W-Re-Hf-C wire. The properties obtained suggested that the wires studied showed promise as potential fiber reinforcement in the 1093 to 1204 C (2000 to 2200 F) temperature range.

Petrasek, D. W.↗

NASA NTRS1972

Metal-metal laminar composites for high temperature applications

A study was conducted to obtain indications of the potentialities of laminar metal-metal composites for elevated temperature use. Most of the composites consisted of multiple layers or laminae of tungsten alternated with laminae of Nichrome V, a ductile, weaker but oxidation-resistant alloy. Composites with 50 volume percent of each phase were tested in tension and stress rupture at temperatures of 871 and 1093 C (1600 and 2000 F) and in impact at 23 and 524 C (73 and 975 F). A tension and a short time stress-rupture test was conducted on specimens of 77 v/o W-Re-Hf-C/23 v/o Inconel alloy 600 at 1093 C (2000 F).

Hoffman, C. A.↗

NASA NTRS1971

Velocities of fragments from bursting gas reservoirs.

A solution is obtained from a simplified approach for the problem of the motion of two fragments driven into a vacuum after rupture of a container filled with gas. Following rupture, the two container fragments are driven in opposite directions. From the separation developed between the moving fragments, the originally contained gas escapes to the vacuum, perpendicular to the direction of motion of the fragments, and with locally sonic velocity. The present solution removes two restrictions of an earlier similar approach: (1) the speed of the gas within the original volume (and consequently, by inference, that of the fragments) is small relative to the sonic escape velocity, and (2) the volume between the separating fragments while they are accelerating undergoes negligible change from the original volume.

Taylor, D. E.↗

NASA NTRS1973

Solid-state and fusion resistance spot welding of TD-NiCr sheet

By using specially processed TD-NiCr sheet in both 0.4-mm (0.015-in.) and 1.6-mm (0.062-in.) thicknesses and carefully selected welding procedures, solid state resistance spot welds were produced which, after postheating at 1200 C, were indistinguishable from the parent material. Stress-rupture shear tests of single-spot lap joints in 0.4-mm (0.015-in.) thick sheet showed that these welds were as strong as the parent material. Similar results were obtained in tensile-shear tests at room temperature and 1100 C and in fatigue tests. Conventional fusion spot welds in commercial sheet were unsatisfactory because of poor stress-rupture shear properties resulting from metallurgical damage to the parent material.

Moore, T. J.↗

NASA NTRS1973

Methods for structural design at elevated temperatures

A procedure which can be used to design elevated temperature structures is discussed. The desired goal is to have the same confidence in the structural integrity at elevated temperature as the factor of safety gives on mechanical loads at room temperature. Methods of design and analysis for creep, creep rupture, and creep buckling are presented. Example problems are included to illustrate the analytical methods. Creep data for some common structural materials are presented. Appendix B is description, user's manual, and listing for the creep analysis program. The program predicts time to a given creep or to creep rupture for a material subjected to a specified stress-temperature-time spectrum. Fatigue at elevated temperature is discussed. Methods of analysis for high stress-low cycle fatigue, fatigue below the creep range, and fatigue in the creep range are included. The interaction of thermal fatigue and mechanical loads is considered, and a detailed approach to fatigue analysis is given for structures operating below the creep range.

Ellison, A. M.↗