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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

TEM Characterization of Nanostructured Grade 91 Neutron Irradiated to 6 DPA at 300 °C and 500 °C

Previously, samples of coarse grained (CG, grain diameter > 1 µm), ultra-fine grained (UFG, 1 µm> grain diameter> 100 nm), and nanocrystalline (NC, 100 nm> grain diameter) grade 91 steel were neutron irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). Grade 91 is a ferritic/martensitic steel alloy with a nominal composition of Fe-9%Cr-1%Mo by weight, and it is being studied for applications in advanced nuclear reactor designs including as a fuel cladding material in gas cooled reactors. To assess the viability of this alloy for nuclear applications, the role of grain boundary size has not been sufficiently explored as a mechanism for developing radiation-tolerant materials. To achieve an UFG microstructure the material was processed using equal channel angular pressing (ECAP), and the NC microstructure was achieved through high pressure torsion (HPT). Such samples of various geometries for different types of characterization were irradiated to a dose of 6 displacements per atom (DPA) at separate temperature conditions of 300 °C and 500 °C. Thus, there were a total of 6 experimental conditions, three grain sizes (CG, UFG, and NC), each irradiated at two different temperature conditions (300 °C and 500 °C).

36 - MATERIALS SCIENCE↗
DOE OSTI2025

Monitoring Methods for Early Detection of Inadvertent Fission Product Release at the Advanced Test Reactor

Isotope effluent data obtained during three instances of experiment failures at the Advanced Test Reactor (ATR) are analyzed to provide an overview of the methods used to detect initial signs of unintended fission product release. The data is contextualized with the operational experience, including means of identification and subsequent mitigation strategies, gained during these events. General trends as well as variations in isotopic behavior between the three failures are explored. Background on the Real Time Monitor, a High Purity Germanium detector, and other fission product monitoring systems utilized at the Advanced Test Reactor is also provided. The presented analysis was used to establish administrative action levels which are currently utilized by ATR for early detection of experiment fission product release. Early identification provides time to make programmatic decisions before approaching safety and environmental limits.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2017

GAIN Fuel Safety Research Workshop

The Gateway for Accelerated Innovation in Nuclear (GAIN) Fuel Safety Research Workshop was held in Idaho Falls, Idaho, at the Idaho National Laboratory (INL) meeting center in the Energy Innovation Laboratory (EIL). The purpose of the workshop was to gather fuel safety research needs from the broad range of industrial nuclear technology users and developers. The needs will be analyzed relative to the national laboratory capabilities to help inform program planning and funding. The workshop provided a forum to learn about transient testing from a research and regulatory perspective and tour participation of the Transient Reactor Test Facility (TREAT), Advanced Test Reactor (ATR), and Materials and Fuels Complex (MFC).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Flux in a 1D-Reactor Cell by Response Matrix

Legacy computational reactor physics software tools and protocols currently used in support of the Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice. Thus, they are difficult, if not impossible, to verify and validate (V and V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of their application and protocols from the 1960's and 1970's is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity transport theory based computational software and models to replace the legacy methods, with appropriate (V and V) protocols according to applicable national standards. The new computational tools include: - The 2-dimensional HELIOS deterministic transport code for integrated ATR core fuel and experiment management and Core Safety Analysis Package (CSAP) physics data generation. The MCNP5 3-dimensional stochastic simulation (Monte Carlo) code for verification and validation applications and for specialized ATR core analysis along with its longstanding traditional role in experiment support - The 2-dimensional deterministic NEWT transport code and the 3-dimensional KENO6 stochastic simulation code, with their SCALE-based support infrastructure for use in V and V applications and as supporting modules for sensitivity-uncertainty analysis via the TSUNAMI system. The ATTILA 3-dimensional discrete-ordinates transport code for specialized experiment support and possibly additional applications. Regardless of application, every neutron transport code, new or legacy, requires verification. The aim of this presentation is to provide a neutron transport benchmark for a simple 1D-reactor cell in plane geometry. The benchmark finds use as a standard of comparison for the codes listed above intended for the ATR upgrade. The method developed here is in contrast to the finite Fourier transform solution found in Ref. 1, which is limited to isotropic scattering and particularly difficult to numerically implement unless the total cross section is constant over the cell.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Evaluation of the Enhanced LEU Fuel (ELF) Design for Conversion of the Advanced Test Reactor to a Low-Enrichment Fuel Cycle

A conceptual low-enrichment uranium (LEU) fuel design has been developed for the Advanced Test Reactor (ATR) at Idaho National Laboratory. The ATR is currently fueled with a high-enrichment fuel but is slated to be converted to LEU under programs led by the National Nuclear Security Administration of the U.S. Department of Energy. A conceptual LEU fuel design, the Enhanced LEU Fuel (ELF), has been developed assuming power peaking control through the use of variable fuel meat thicknesses and no use of burnable poison. In initial work, this design was shown to satisfy performance requirements for ATR operation. Following these design calculations, a safety analysis process was initiated to demonstrate that the ELF design would successfully meet safety limits for postulated accident conditions. Those calculations, performed using RELAP5 and ATR-SINDA, require physics analysis to provide spatial power distributions and kinetics parameters for various core operations configurations. Finally, this article describes the findings of the physics analysis and provides predictions for the behavior of a LEU-fueled version of ATR, and compares these to calculations of the performance of the current high-enrichment uranium fuel.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Preliminary Design Evaluation of a Natural Circulation Molten Salt Irradiation Loop

Many promising Molten Salt Reactor (MSR) concepts under development rely on relatively untested salt compositions. Some level of irradiation data will likely be necessary for the licensing of these concepts. A feasibility study is currently underway at Idaho National Laboratory to establish the possibility of hosting a compact molten salt irradiation loop inside the Advanced Test Reactor (ATR). A case study for a uranium-fueled chloride-based salt will be provided in this paper. Neutronic tools are used to obtain flux levels inside of a specific position inside the ATR. Thermal hydraulic simulations are then conducted to scope the achievable natural circulating velocities and inform design decisions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

MP1 Intermediate Characterization Report Summary

Great effort worldwide has been invested in replacing high enriched uranium (HEU) fuel with low enriched uranium (LEU) fuel in nuclear plants. This effort aims in safeguarding nuclear material and promoting the proliferation resistance of current reactors. The United States Department of Energy (DOE), National Nuclear Security Administration's (NNSA) Materials Management and Minimization Reactor Conversion program is actively working to convert civilian research and test reactors from the use of HEU to LEU fuel. To date, DOE NNSA has converted or verified the shutdown of various domestic reactors and also reactors worldwide. The task to develop and test a new LEU fuel while maintaining current performances has been assigned to the High-Performance Research Reactor (USHPRR) program. U alloyed with Mo achieves the goal of high uranium density necessary for the conversion of the high performance research reactors, and has also demonstrated good performance in pile [1]. The alloying process of U with Mo stabilizes the body-centered cubic (bcc) gamma phase (?-U) at room temperature. Without this alloying property enhancement, this phase is only stable at temperatures only above 770 °C. Retaining this isotropic gamma phase within the U-Mo alloy provides swelling and oxidation resistance and desired mechanical properties [2]. Generally 10% w. has been considered sufficient to achieve stabilization. However, local phenomena within the alloy can affect the local Mo content and lead to gamma phase decomposition. Because of this, it has been suggested that controlling gamma phase decomposition is most critical during processing [2-3]. The work detailed in this document is part of the HPPR program and focuses on the characterization of fresh fuel (a.k.a. as fabricated, before reactor testing) in collaboration with other laboratories including PNNL (Pacific Northwest National Laboratory), LANL (Los Alamos National Laboratory) and BWXT (Babcock and Wilcox Technologies). This testing is part of a 4 way validation for which similar samples from master plates were taken and distributed to each laboratory to independently analyze utilizing a standardized set of measurement and analysis procedures. When complete this work will supply a “toolbox” of information for alloying and fabrication parameters with its direct effects on microstructure quantified. Various parameters have been analyzed in this characterization effort (including grain size, quality of the coating layer, composition of the samples, carbide content etc.), based on the direction of the working group in ref.[4]. The final aim of this work is to increase the understanding of the impact of processing conditions on the final fuel microstructure, using fabrication processes that meet “commercial viability” requirements [4]. The strength of this campaign rests on the extent of the samples studied and the variation of the manufacture parameters which will permit to achieve the required understanding. Thus permitting to use these experiments to down-select test samples used to evaluate irradiation performance [4] to be tested in various locations of the Idaho National Laboratory (INL) Advanced Test Reactor (ATR).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

2021 ATR Strategic Plan

Collaborative development of nuclear energy science and technology by three major sectors—academia, the commercial nuclear power industry, and the federal government—is key to meeting the challenges of nuclear energy. All three share a common need for experimental capabilities, whether for basic science investigations, applied research in nuclear fuels and materials, or validation of data. The Advanced Test Reactor (ATR) provides the unique irradiation capabilities to support the need.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Pu-238 Production Calculator and Estimation Tool

A MS Excel tool was created to help track Pu-238 production, Np-237 feedstock use, estimate production scenarios, and coordinate planning for shipments. As-run estimates were used as a basis for beginning and end of life isotopic concentrations. Multiple target types were created to demonstrate a proof of concept while actual results were not yet available The initial goals of this production tool were: • Track unirradiated, currently irradiated, and irradiated targets at the Advanced Test Reactor (ATR) • Estimate production yields from irradiated targets in various positions for each cycle • Provide past, current, and future year quantities of Np-237 and Pu-238 at ATR, amounts shipped to ATR and amounts shipped from ATR • Estimate production from current and future target designs based on data from numerical modeling (values will be provided by Idaho National Laboratory (INL)) • After proof of principle, additional work is planned that will further enhance the tool with: o Incorporate the High Flux Isotope Reactor (HFIR) reactor at Oak Ridge National Laboratory (ORNL) for irradiation o Incorporate irradiations and target shipments from ORNL to INL and back to track Np-237 transfers across the Department of Energy (DOE) Complex

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

A standard capsule design for structural material testing in the Advanced Test Reactor

Nuclear materials testing is commonly completed in various material test and research reactors throughout the world, including the Advanced Test Reactor (ATR), but the current capsule design, analysis, and fabrication process can take years to complete. To decrease the costs and time associated with materials testing, a standard capsule has been designed which houses various specimen geometries and allows irradiation in virtually any ATR test position. The standard capsule features locking end caps which connect and lock together to form the capsule stack, eliminating the need for a basket and maximizing the quantity of specimens which are contained within the capsule. A customizable internal gas gap provides thermal resistance between the specimens and reactor coolant, making specimen temperatures from approximately 370 to 1000 K achievable. The flexibility of the capsule design allows experimenters to choose irradiation positions based off desired neutron flux, with typical fluences per cycle ranging from 8.8x10 19 to 2.3x10 21 n/cm 2 , depending on experiment position. Here this paper presents and discusses the standard capsule design and analysis.

36 MATERIALS SCIENCE↗
DOE OSTI2014

AGR-2 Daily As-run Thermal Analyses

This report documents the daily as-run thermal analyses performed on Advanced Gas-cooled Reactor (AGR) experiment AGR-2 in the Advanced Test Reactor (ATR) for capsules 2,3,5, and 6. Capsule 1 (French) results are in Appendix F, while Capsule 4 (South African) results are in Appendix G. Appendix F, and G are separate documents Several fuel and material irradiation experiments, which support the development of the Next Generation Nuclear Plant (NGNP), are planned for the Advanced Gas Reactor Fuel Development and Qualification Program. The goals of these experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post-irradiation examination (PIE) and safety testing. AGR-2 was the second in this series of planned experiments to test tristructural-isotropic (TRISO)-coated, low enriched uranium oxycarbide (LEUCO) fuel. The AGR-2 experiment was intended to serve as a follow-on to the AGR-1 experiment. The AGR-2 experiment is a high temperature fuel particle test. The test train planned for AGR-2 is based on the experience gained from previous irradiations in the ATR, using instrumented lead experiments. Instrumented lead experiments are used for irradiations requiring a controlled environment and monitored parameters. The experiment test train positions the fuel within the test location and contains sweep gas lines and thermocouple wiring that is routed through access ports to external support systems. The AGR-2 experiment is comprised of six individual capsules, approximately 1.375-in. in diameter by 6-in. long, stacked on top of each other to form the test train. Each capsule contains 12 fueled compacts that are approximately 0.5-in. in diameter by 1-in. long. The compacts are composed of fuel particles bound together by a carbon matrix. Each compact contains approximately 4,150 fissile particles (35 vol% particle packing fraction). Each capsule will be supplied with a flowing helium/neon gas mixture to control the test temperature and to sweep any fission gases that are released to the fission product monitoring system. Temperature control is accomplished by adjusting the gas mixture ratio of the two gases (helium and neon) with differing thermal conductivities. Many similarities exist between this thermal analysis for AGR-2 with the AGR-1 daily as-run report in [1]. Gas gaps were taken from as built drawings in Reference [2]. The finite element heat transfer code ABAQUS [3] was used to perform the thermal analysis. The AGR-2 experiment was placed in the B-12 position in the ATR core as shown in Figure 1. The AGR-2 experiment capsule assembly consists of six capsules axially stacked. Each capsule contains a graphite holder with three equally spaced fuel compact holder openings as shown in Figure 2. Each holder opening accommodates four axially stacked fuel compacts. Thus, each capsule has three stacks by four fuel compacts per stack for a total of 12 fuel compacts per capsule. So, within the entire AGR-2 experiment capsule assembly, there are six capsules by 12 fuel compacts per capsule for a total of 72 fuel compacts. Six finite element heat transfer models were created (one for each capsule), each with a corresponding gas gap.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

A machine-learning-aided data recovery approach for predicting multi-material thermal behaviors in advanced test reactor capsules

Instrumented experiments conducted at test reactors are essential to the deployment of new advanced reactor systems. Designing new experiments and generating data on specific reactor conditions require significant investments in terms of both time and cost. Finite element analysis software can be used to create high-fidelity models of experiment environments in order to support the actual experiments, but computation time remains a concern in terms of applying outcomes to real-time usage of data (e.g., a digital twin [DT]). Here, the present research proposes a machine-learning (ML) aided approach to making temperature and displacement predictions based on the thickness of the outer gas gap on the experimental capsule used for in-pile demonstration of a novel new thermal conductivity probe in the Advanced Test Reactor (ATR). This capsule consisted of U10Zr fuel, a rodlet, sodium, and inner and outer capsules. Gas gaps existed between the fuel and the rodlet, and between the inner and the outer capsule. The learning data pertained to an experimental capsule's radial distributions of temperature and displacement, as obtained based on Abaqus and the physical features. For the first step of ML sequence, the temperature was predicted using three positional parameters. Next, the displacement was predicted using seven additional parameters. Each physical feature was normalized in order to be both nondimensional and standardized. The temperature and displacement predictions showed good agreement with the simulation results in all cases involving interpolation and extrapolation. Furthermore, data similarity enhancement increased the similarity between the training and the target data, thereby increasing the predictive accuracy of the ML models. In certain extrapolation cases involving limited original ML model accuracy, data similarity enhancement and data recovery was able to somewhat improve this accuracy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Improvement of the $\mathrm{BISON U_3Si_2}$ modeling capabilities based on multiscale developments to modeling fission gas behavior

Uranium silicide (U 3 Si 2 ) is a concept explored as a potential alternative to UO 2 fuel used in light water reactors (LWRs) since it may improve accident tolerance and economics due to its higher thermal conductivity and increased uranium density. U 3 Si 2 has been previously used in research reactors in the form of dispersion fuel, but operated at lower temperatures than commercial LWRs. The research reactor data illustrated that significant gaseous swelling occurs as the fuel burnup increases. Therefore, it is imperative to understand the fission gas behavior of U 3 Si 2 under higher temperature LWR operating conditions. In this work, molecular dynamics and phase-field modeling techniques are used to reduce the uncertainty in select modeling assumptions made in developing the fission gas behavior model for U 3 Si 2 in the BISON fuel performance code. These lower length scale informed models are then utilized in the validation of BISON U 3 Si 2 modeling capabilities to simulate the ATF-1 experiments irradiated in the Advanced Test Reactor (ATR). Sensitivity analysis (SA) and uncertainty quantification (UQ) are included as part of the validation process to identify where further experiments and lower length scale modeling would be beneficial. Here, the multiscale modeling approach utilized in this work can be applied to new fuel concepts being explored for both LWRs and advanced reactors (e.g., uranium nitride, uranium carbide).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

BRR Cask Use for Pu-238 Isotope Production

The Department of Energy (DOE), in partnership with its national laboratories and the National Aeronautics and Space Administration (NASA), is responsible to produce Pu-238 isotope in the United States for use in space exploration. Major activities in the DOE complex are focused at Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL). INL is responsible for storing Np-237 feedstock, irradiation qualification in the Advanced Test Reactor (ATR), and irradiating targets containing Np-237 to produce Pu-238. ORNL is responsible for target design, target fabrication, irradiation qualification in HFIR, and processing of targets to extract Pu-238 heat source material. A key part of the program lifecycle is moving irradiated targets from INL to ORNL. The BEA Research Reactor (BRR) cask was identified as a potential shipping cask for the transport of both unirradiated and irradiated targets between the project sites. This paper will discuss the production at INL and shipment of Pu-238 to ORNL using the BRR cask.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

Vibration Analysis of Advanced Test Reactor Mini-Plate Hydraulic Test in the OSU HMFTF

Currently, flow testing is being performed at Oregon State University (OSU) on the Hydro-Mechanical Fuel Test Facility (HMFTF) to aid in the qualification of low enriched plate-type Uranium fuels for use in high performance research reactors (HPRR) in the US, under the Global Threat Reduction Initiative (GTRI). Under this project, varying fuel geometries proposed for irradiation testing in the Advanced Test Reactor (ATR) are being hydraulically tested in the HMFTF prior to being placed in the ATR; these tests included the use of high-resolution tri-axis accelerometers affixed to the experiment in multiple locations. The premise of this study was to conduct a vibration analysis on the test designated as miniplate1 (MP-1) Large B experiment which is planned for irradiation testing in the ATR's Large B experiment location. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Overcoming challenges to support US resumption of high specific activity cobalt-60

Domestic production of high specific activity 60 Co was halted after a target rupture in 2012 at the Advanced Test Reactor (ATR). The Isotope Program (IP) within the US Department of Energy (DOE) Office of Science tasked a multilaboratory team of researchers and managers from Oak Ridge and Idaho National Laboratories with the redesign the radioisotope capsule. The objective of this effort was to create a more robust and reliable design, compared to the pre-2012 target. The team successfully completed this task to produce the DOE-IP cobalt (Co) production capsule design. Furthermore, 66 capsules were successfully fabricated by Oak Ridge National Laboratory (ORNL) and delivered to Idaho National Laboratory (INL) for irradiation in the ATR between January 2014 and October 2016. This paper describes the efforts of the team to prepare and disposition the two initial DOE-IP Co production capsules that were processed in March 2020. These efforts include performing accurate production predictions, experimentally validating predictions with assay measurements, shipping with the Orano-furnished Battelle Energy Alliance Research Reactor shipping package, and disassembling capsules at the isotope vendor site.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

AGR-1 Depletion Benchmark

The AGR-1 TRISO particle irradiation test was the first Advanced Gas Reactor test irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The TRISO particle fuel was 350 μm diameter uranium oxy-carbide (UCO) kernels enriched to 19.75 wt% U-235. The particles were embedded in cylindrical graphite compacts, placed in test capsules, and irradiated over a period of three years or 13 ATR power cycles. The fuel compacts achieved burnups approaching 20% FIMA and accumulated fast fluence up to 4.25E+25 n/m 2 (E n >0.18 MeV). This report describes the AGR-1 test train, capsules, compacts, and TRISO particle fuel in sufficient detail for a physics analyst to model the AGR-1 experiment as part of a larger ATR neutron transport model and perform a depletion calculation. The intent of the data herein is to define a practical numerical TRISO particle depletion benchmark problem for the AGR 1 irradiation test.

42 ENGINEERING↗
DOE OSTI2021

Yttrium Hydride Post-Irradiation Examination Plan

Micro nuclear reactors are expected to use metal hydride moderators to ensure thier compact design for the flexible operational purposes. Department of Energy’s Microreactor program therefore aims to provide infrasture and qualification bed for the development of micro nuclear reactors. Because micro nuclear reactors must be compact to offer a flexible operational purposes, metal hydrides are expected to be used as neutron moderator materials due to their high hydrogen number density. Los Alamos National Laboratory (LANL) directed by the DOE to provide a qualified fabrication techniques of metals hydride, especially the yttrium hydride (YHx) which provides advantages over other metal hydride candiates. Qualification of YHx involves fabrication, property, and performance pillars, which must be met for the deployment of microreactors. In the support of the qualification process metal hydride, LANL has developed fabrication techniques based in direct hydriding and powder compacting. For both manufacturing techniques, fresh material properties of metal hydrides have been reported in the literature. To establish the performance limits and the qualification parameters, an irradiation campaign has been initated at Idaho National Laboratory (INL). The irradation campaign aims to investigate the critical irradiated YHx properties as a function of neutron fluence and temperaute. The first phase of the irradiation campaign includes the effect of the operating temperature under neutron irradiation which provides initial data on the metal hydrides and informs the manufacturing processes. So that, drop-in irradations in Advanced Test Reactor (ATR) has been designed. This document depicts the post-irradation examination (PIE) plan of the ATR irradiated YHx. Objective of the PIE plan is to serve as a qualification basis for YHx as a neutron moderator material by understanding the impacts irradiation has on the physical and thermophysical properties; to establish a link between fabrication and the irradiated YHx properties; to institute a PIE strategy for other solid-state neutron moderator candidates, and; to inform micro nuclear reactor fuel performance and safety codes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗