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At least 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Exploration of LIBS as a novel and rapid elemental mapping technique of nuclear fuels in the form of surrogate TRISO particles

Laser-induced breakdown spectroscopy (LIBS) was employed to characterize coatings on surrogate fuel particles. Tri-structural isotropic (TRISO) particles are a proposed nuclear fuel alternative for high temperature reactors. These particles are constructed of a ZrO 2 kernel (as a surrogate to uranium), surrounded by an inner pyrolytic carbon layer and are surrounded by an outer carbide layer (ZrC, presented here) to act as a barrier to fission products generated during nuclear reactions. These particles are embedded within a graphite compact and housed within the reactor core. Simply put, due to their robust nature, performing elemental analysis of these particles poses a challenge. Presented here, LIBS is explored as a method for characterizing elemental constituents of these particles, with the focus being on rapid elemental mapping and depth profiling. Different from traditional elemental analysis techniques (e.g., inductively coupled plasma – based methods), LIBS is advantageous because it can directly analyze the sample surface and can detect light elements such as C and O, making it a viable technique for the analysis of small, multilayered particles as spatial elemental information is warranted in the production of these particles. In the work presented here, LIBS was successfully used for discerning small layers (30–50 μm), detecting the location of carbon and oxygen layers, providing fast 2-D mapping (<5 min per particle) and rapid depth profiling (10 s per particle).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Development and Experimental Validation of a Heat Transfer Model for Spilled Molten Salt Pools

A spill of radionuclide-bearing molten salt is one of the major postulated events that needs to be analyzed for liquid fluorine salt-cooled high-temperature reactor (FHR) or molten salt reactor licensing purposes. In this postulated event, radioactive source term materials (RSTMs) in the molten salt are discharged from the reactor vessel to the reactor building. The release of RSTMs from the spilled salt pool to the gas space in the reactor building is expected to be controlled by the cooling behavior of the spilled salt, including the growth and shrinkage of the solid crust on the surface of the spilled salt pool. This paper presents a simulation model for spilled salt pool heat transfer and validation efforts. The validation data come from two molten salt spill tests that were performed recently: the PELE2 test by the Rapid Experimental Laboratory of Kairos Power LLC (KP) and the Argonne salt cooling test conducted by Argonne National Laboratory. The former was a large-scale test involving kilograms of molten spilled FLiNaK salt, and the latter was a relatively smaller-scale test targeting various processes associated with a salt spill event. Both tests generated valuable data sets that can be used to assess salt cooling and validate evaluation models. This paper provides a new one-dimensional model that can simulate the cooling process of a spilled salt pool as well as the thermal responses of heat structures, such as the stainless steel liner and the concrete below the salt. The model has been implemented as part of KP-SAM code, which is a branch of the systems code SAM specific to KP FHR. In conclusion, the simulation results of the model are compared with the data of the PELE2 and Argonne tests, and reasonable agreements are observed between the model and test data.

heat transfer model↗
DOE OSTI2022

Thermal Aging Effects on the Yield and Tensile Strength of 9Cr-1Mo-V (Grade 91)

The long-term exposure of 9Cr-1Mo-V (Grade 91) steel to elevated temperature can have a significant effect on reducing its yield and tensile strength. The yield and tensile strength changes, in turn, have potential implications to the assurance of the integrity of components constructed with this material over their design or intended lifetime. The ASME Boiler & Pressure Vessel Code (BPVC), Section III, Division 5 (III-5, high temperature reactors) provides tabulated reduction factors for Grade 91 yield and tensile strength as a function of exposure temperature and time up to 300,000 h. ASME BPVC III-5 s intent to extend these factors to an exposure duration of 500,000 h, the lack of available historic information to support the existing factors, and the recent development of a physics-based prediction model for ASME BPVC application are prime motivation for this study. This work describes results of the conventional time-temperature Hollomon–Jaffe parameter, strength reduction ratio prediction method using an updated, extensive Grade 91 unaged and related aged material strength database. The method, previously used by Oak Ridge National Laboratory in its evaluation of Grade 91 and likely used in development of the existing BPVC III-5 reduction factors, provides strength reduction ratio predictions useful for general component fitness-for-service assessments and for computing BPVC III-5 reduction factors as defined. Specific reduction factors to 500,000 h at 650 °C applicable to ASME BPVC III-5 are computed from the strength reduction ratios.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Setup and verification of a SCALE/KENO platform for generic FHR benchmark calculations

The work presented in this article is preliminary to downstream analysis of a generic fluoride salt-cooled high-temperature reactor (gFHR) core performed by the University of Tennessee in collaboration with Kairos Power (KP). A Monte Carlo transport model of the publicly available gFHR equilibrium core is developed in SCALE/KENO with multigroup energy treatment. Several output quantities of interest are used to verify the simulations against a benchmark model developed by KP using the continuous energy Monte Carlo code Serpent 2. Good agreement is seen in flux and fission rate profiles with a maximum relative difference of 1.4% and 2.8% respectively. Furthermore, an effective multiplication factor bias of 44 pcm was observed between the two simulations. The fuel temperature reactivity coefficient calculated with SCALE is within uncertainty to the reference model. This verification acts as a publicly reproducible benchmark for the gFHR in SCALE/KENO. A simplified depletion model is also presented where a single fuel pebble is depleted to discharge burnup through the equilibrium core while the equilibrium core is assumed to be invariant. This method produces results that intercept the equilibrium core concentrations in every case, however, an interesting artifact of this particular depletion model is uncovered. The phenomenon is shown to be a fundamental feature of the differential rate equations and inspires questions about how this system behaves when the time evolution of the equilibrium core is considered. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Comprehensive Margin Assessment of the ASME Section III, Division 5, Class A Primary Load Design Rules

This report provides a comprehensive margin assessment of the ASME Section III, Division 5, Subsection HB, Subpart B rules for the design of high temperature reactor components against load controlled stress limits. These rules, often called the Code primary load design rules, provide protection against creep rupture and plastic collapse under steady conditions. The method adopted here is to compare the deterministic ASME design life, according to the primary load design rules, against the expected, statistical service life of the component, considering creep rupture as the relevant failure mode. The report provides the design margin of the ASME rules in terms of the probability of premature failure -- the probability that the component fails in service before reaching the ASME design life. To complete this assessment, the report describes the development of a statistical creep life assessment procedure accounting for variations in the component loading, material creep rate, and the material rupture time. This includes a novel method for correlating creep rupture data using a Gaussian process accounting for heat-to-heat variation in the Larson-Miller time-temperature parameter. The report applies the complete margin assessment process to quantify the design margin inherent in the ASME primary load design rules as a function of temperature for relevant component geometries and materials. The overall conclusion is that the ASME rules are conservative, providing designs with very small probability of premature failure, but some optimization of the Code design rules is possible to provide a more uniform design margin.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Thermal Neutron Scattering Law Evaluations for Zirconium Carbide and Critical Mass Calculations [Slides]

Zirconium Carbide (ZrC) is a refractory material currently being investigated for high temperature reactor applications. ZrC was studied as a base of (U,Zr)C solid solution and UC 2 composite fuels for NERVA-type cores during the Rover program. Testing of the material was performed at Nuclear Furnace 1 (NF-1) located at Los Alamos. Thermal Scattering Law (TSL) evaluations generated with FLASSH are a new contribution to ENDF/B-VIII.1.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Interim Mechanical Properties Data from FY2022 ORNL Testing of A709 with Precipitation Treatment for ASME Code Case Data Package

A collaborative research and development effort in support of the Alloy 709 Code Case qualification in the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Division 5, High Temperature Reactors is being conducted at the US Department of Energy’s Oak Ridge National Laboratory, Idaho National Laboratory, and Argonne National Laboratory. A recent assessment on the Alloy 709 development and testing effort concluded that the mechanical properties of Alloy 709 with the precipitation treatment continued to outperform those of Type 316 stainless steel. The assessment also affirmed the recommendation for its Code qualification. This report provides the detailed integrated data generated to date on the two commercial heats of Alloy 709 in plate product form with precipitation treatment condition that were used as the basis for this assessment. Oak Ridge National Laboratory has continued to perform a subset of the Code Case testing for tensile properties, creep rupture, fatigue, and creep-fatigue. This report also updates the key Alloy 709 Code Case testing status and results in FY 2022 at Oak Ridge National Laboratory.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Progress Report on Alloy 709 Base Metal Code Case Testing at ORNL in FY 2023

A collaborative research and development effort in support of the Alloy 709 Code Case qualification in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5, High Temperature Reactors is being carried out at Oak Ridge National Laboratory (ORNL), Idaho National Laboratory (INL) and Argonne National Laboratory (Argonne). In FY 2023, ORNL has continued to conduct a subset of the Code Case testing for tensile, creep rupture, fatigue and creep-fatigue on Alloy 709. This report also updates the key Code Case testing status and results on the first two commercial heats and the preliminary results on the third commercial heat of Alloy 709. The three commercial heats of Alloy 709 are all in plate product form.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Molten Salt Reactor Experiment Simulation using Shift/Griffin

The Department of Energy (DOE)’s NEAMS focuses its efforts on the development of advanced modeling and simulation (M&S) tools for light-water reactors (LWRs) and non–LWRs (i.e., molten salt reactors, high-temperature gas reactors, microreactors, and fast reactors). In the previous fiscal year, the Multiphysics Applications Driver Technical Area funded molten salt reactor (MSR) M&S at Oak Ridge National Laboratory (ORNL) to generate multigroup macroscopic cross sections with Shift for a MSRE 2D lattice model in Griffin. In addition, Shift’s capability to calculate gamma dose rates from activated components in the primary exchangers in a molten salt breeder reactor was also demonstrated. In fiscal year 2023, ORNL generated multigroup macroscopic cross sections using Shift for a 3D MSRE core model. MSRE depletion calculations using Griffin were also demonstrated in this fiscal year. For the depletion calculation, one-group microscopic cross sections for the 3D MSRE core were generated using Shift, and the decay transmutation library from ORIGEN was converted to an ISOXML file, which is required as input in Griffin. Several Monte Carlo codes, such as OpenMC and Serpent, were also used to benchmark and supplement multigroup cross sections generated by Shift. Multigroup libraries were generated with 8 and 20 group structures, and the study found the 8-group structure to be more accurate when comparing Griffin results to continuous energy (CE) Monte Carlo results. The average flux from CE Shift calculations is up to ~6% higher than the CE Serpent calculations because of different values applied for the energy released per fission (κ values). The average flux in the fuel salt calculated by Griffin using cross sections generated with Shift agrees well with the reference CE Shift solution; the same is valid for the corresponding Serpent results. The maximum relative error is ~6% and ~2% compared to the CE Shift and Serpent reference solutions, respectively. Meanwhile, the average flux calculated by Griffin in the graphite moderator shows a higher difference in the thermal range when compared to both reference Monte Carlo solutions; this result suggests a need for improvement in cross section generation for the graphite moderator in the thermal range in both Monte Carlo codes. Griffin depletion calculations using cross sections from Shift and Serpent were performed and compared against ORIGEN calculations, and the nuclide densities calculated by Griffin were found to be generally in agreement with those of ORIGEN. Because a different approach was taken to calculate the energy released per fission ( κ values) in Shift and Serpent, a difference in nuclide densities from differences in the average flux was observed between Griffin using Serpent and Shift cross sections. Griffin calculations with Shift cross sections produced higher average flux in the salt than with Serpent cross sections, leading to higher consumption of 235 U and higher production of 135 Xe. For time-dependent depletion calculations, cross sections were generated with Serpent, and Griffin’s results using these cross sections were compared to CE Serpent depletion results, demonstrating good agreement. The average difference in keff between Serpent and Griffin as a function of burnup is about 155 pcm. Similarly, good agreement with small differences up to ~0.5% was also noticed in the nuclide density of 235 U and 135 Xe. More details regarding the methodologies invoked to generate the cross section to make code-to-code comparisons are discussed further in this report. User feedback on Griffin and Shift capabilities that will enhance these calculations is provided in this report for future consideration. The work performed this fiscal year can be extended further for multiphysics coupling of Griffin-Pronghorn/SAM with Mole to study precursor flow and salt chemistry.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Baseline Characterization Database Verification Report - NBG-18 Billet 635-14

The purpose of this Engineering Calculations and Analysis Report is to present the data being collected in the Baseline Graphite Characterization program, which is directly tasked with supporting the Idaho National Laboratory’s (INL’s) research and development efforts on the Next Generation Nuclear Plant (NGNP)/Very High Temperature Reactor (VHTR)/Advanced Reactor Technology (ART). This program is populating a comprehensive database that will reflect the baseline properties of nuclear-grade graphite regarding individual grade, billet, and position within individual billets. The physical and mechanical property information being collected will be transferred to the NGNP Data Management and Analysis System (NDMAS), and from that database will help populate handbook of property data available to member nations of the Generation IV International Forum (GIF).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Summary of FY 2024 A709 Code Case testing at ANL, INL and ORNL

A collaborative research and development effort in support of the Alloy 709 Code Case qualification in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5, High Temperature Reactors is being carried out at Oak Ridge National Laboratory (ORNL), Idaho National Laboratory (INL), and Argonne National Laboratory (ANL). Key testing data for the Alloy 709 100,000-hr near-term Code Case submittal to ASME is expected to be completed by the end of 2024, with design parameters anticipated to be finalized in FY 2025. This report summarizes the testing results for three commercial heats of Alloy 709 conducted across three laboratories, reviews the current testing status, and outlines the remaining data needed to support the first Alloy 709 Code Case submittal to ASME. The Alloy 709 Code Case plan remains on schedule.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Summary of Leach-Burn-Leach Round-Robin Test Results

Tristructural isotropic (TRISO) coated particle fuel is of interest to multiple member states associated with the Generation IV International Forum (GIF). Qualification and acceptance testing of TRISO fuel includes the determination of particle defect fractions and impurities in the as-fabricated TRISO fuel particle and/or composited fuel form, which are typically determined using burn-leach (BL) or leach-burn-leach (LBL) analysis methods. A round-robin test focusing on LBL analysis was undertaken by the GIF Very High Temperature Reactor System Fuel and Fuel Cycle Project Management Board. Participating in this LBL round-robin test were research staff from the Institute of Nuclear and New Energy Technology (INET) in the People’s Republic of China, the Korean Atomic Energy Research Institute (KAERI) in the Republic of Korea, and Oak Ridge National Laboratory (ORNL) in the United States of America. The motivation of this LBL benchmarking exercise was to explore consistency in results generated by LBL and to elucidate the potential influence of variations in the procedures used by the participating members on the accurate determination of the measured fuel properties.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

New ASME Section III, Division 5 Creep-Fatigue Design Rules Based on EPP and SMT Approaches

The integrated elastic-perfectly plastic (EPP) and simplified model test (SMT) creep-fatigue (CF) design methodology, referred to as the EPP-SMT method, is being developed as an alternative for CF evaluation in the design of pressure boundary components for high-temperature reactors. This report reviews the conceptual basis of the EPP-SMT methodology, summarizes the SMT experimental development efforts and results, and provides the technical basis for finalizing the EPP-SMT CF design curves for Alloy 617, based on a combined experimental and numerical approach conducted in FY 2024. This report presents the effect of hold time on the CF design curves for Alloy 617 at elevated temperatures. It includes proposed EPP-SMT CF design curves and tabulated values for continuous cycling, along with the effects of maximum hold time, for the use of this EPP-SMT CF evaluation method.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Advanced Manufacturing Techniques and Compositions of High Entropy Alloys for Nuclear Applications

In line with the objectives of the Department of Energy, Office of Nuclear Energy, Advanced Materials and Manufacturing Technologies (AMMT) program, this work focuses on new materials development and qualification research and development for next-generation, high-temperature nuclear reactors. High entropy alloys (HEAs) have the potential to serve in these extreme environments of next-generation nuclear reactors because of their unique phase transformation pathways and nanoscale and mesoscale microstructures. The current work focuses on understanding such nuclear-energy-relevant HEAs through a detailed literature survey, selected experimental work, and developing a decision matrix with criteria for the identification of HEAs that may have the most impact and value for further examination.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Geochemical and Microbial Dynamics of Hydrogen in a Methane Storage Reservoir

Hydrogen has been identified as a flexible energy carrier with zero or negative emission across multiple energy systems, and existing natural gas infrastructure could be leveraged if hydrogen gas (H2) was blended with methane (CH4). For example, subsurface methane storage reservoirs could be slightly modified to also store hydrogen if a methane/hydrogen blend were injected. However, the compatibility of methane storage reservoirs to include H2 injection has not been fully demonstrated, and this could lead to geochemical and microbiological reactions that alter the reservoir and stored gas content. It is essential that we understand the impact of H2 gas on the biogeochemistry of subsurface storage reservoirs before deploying large-scale H2-CH4 storage, We collected produced fluid from two separate methane storage reservoirs in the Southwestern US. First, we completed a baseline analysis of the biogeochemistry through qPCR, 16S rRNA sequencing, metagenomic sequencing, and geochemical analysis. Each reservoir was found to have unique geochemical conditions and a unique microbial community structure, with Site 1 having a higher TDS and an abundance of Shewanella and Site 2 having a lower TDS and high abundance of Eubacterium and Acetobacterium. Next, we ran a series of high pressure, high temperature reactors under hydrogen storage conditions with the biological sample from one of the storage reservoirs and a 20% H2-80% CH4 gas blend for up to 7 days. Our results show a decrease of hydrogen by 5% in reactors as early as 1-3 days. Previous hydrogen storage work has linked subsurface microorganisms with methanogenesis hydrogen sulfide production, acid production, and microbial corrosion. Our results show minimal change in the fluid chemistry, with the exception of a decrease in dissolved sulfate concentrations. Taxonomic sequencing demonstrated the presence of microorganisms capable of iron redox, acid generation, and hydrogen sulfide production throughout the reactors, suggesting microbial hydrogen consumption may occur through various metabolic pathways. This work demonstrates that site-specific geochemistry and microbiology may impact the efficiency of hydrogen storage in methane storage reservoirs.

environmental microbiology↗
DOE OSTI2025

FY25 Status Update on A709 Code Case Testing at ANL

This report provides an update on the status of the creep rupture testing on the precipitation-treated (PT) Alloy 709 samples fabricated from the first, the second and the third commercial heats, in support of the first, 100,000-hour Alloy 709 Code Case development under the American Society of Mechanical Engineers (ASME) Section III, Division 5 for high-temperature reactor construction. In Fiscal Year (FY) 25, no new tests were initiated, while 5 previously loaded tests were completed. This report presents an overview of the creep-rupture data gathered at Argonne National Laboratory (ANL) since FY21 that totaled 52 completed tests. Metallographic images showing crack morphologies in some representative samples ruptured in FY25 are also included. Key conclusions include: 1) the creep behavior of the three heats of Alloy 709 can be well described by the Larson-Miller relationship; 2) higher temperatures and larger stresses lead to shorter creep life and higher minimum creep rate; 3) Alloy 709 is overall very ductile under creep; 4) the crack morphologies near fracture surfaces are dependent on material’s grain size and test conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Complete Initial Studies in Microstructural Studies of Nuclear Graphite

Nuclear graphite is a critical material for the performance and safety of high-temperature reactors, making its study essential for the successful deployment of next-generation reactors in the United States. During the past year, Oak Ridge National Laboratory (ORNL) has contributed extensively to advancing the understanding of nuclear graphite’s properties, behavior, and applications; particular focus has been given to irradiation effects, oxidation mechanisms, thermal and mechanical performance, and long-term stability under reactor conditions. This report provides an update on the ongoing publications being prepared at ORNL. These studies are collectively intended to build a comprehensive knowledge base on nuclear graphite. The studies focus on key aspects such as microstructural characterization, oxidation behavior, advanced analytical techniques, molten salt infiltration, technique development, and billet-to-billet variability. By summarizing the progress of these publications, this report outlines ORNL’s contributions to the field, identifies knowledge gaps, and proposes future research directions. The findings presented in these studies will help enhance the understanding of nuclear graphite’s behavior, support the qualification of nuclear graphite materials, further the development of new techniques and guidelines for using microscopy techniques, and improve reactor safety and performance.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Experimental Results on SS 304H and 2.25Cr-1Mo in Support of Developing Inelastic Constitutive Models

To address critical data gaps necessary for updating viscoplastic constitutive material models for Class A materials used in inelastic design analysis under the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5, High Temperature Reactors, Oak Ridge National Laboratory (ORNL) conducted experimental studies on stainless steel 304H (SS 304H) and 2¼Cr-1Mo steel (F22), to probe their temperature dependent deformation behaviors. These studies covered a range of temperatures up to the maximum limits specified in Division 5 and are intended to support the development of material models by collaborators at Argonne National Laboratory (ANL).

36 MATERIALS SCIENCE↗