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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 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

ECAR-3051 As-Run Physics Analysis For The AGC-3 Experiment Irradiated In The ATR

This Engineering Calculations Analysis Report (ECAR) documents the results of the Advanced Test Reactor (ATR) detailed physics analyses performed to calculate the displacements per atom (DPA) and the fast neutron fluence (E > 0.1 MeV) of the Advanced Graphite Creep (AGC) experiment, AGC-3, irradiated in the ATR East Flux Trap (EFT) (see Figure 1) during ATR Cycle 152B, Cycle 154B, Cycle 155A, and Cycle 155B. This ECAR also reports the neutron and photon heat rates for the materials of the AGC-3 experiment for ATR Cycle 152B. The results for these evaluations and analysis are reported herein.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

ECAR-1406 AS-RUN NEUTRONIC ANALYSIS OF THE AGC-1 EXPERIMENT IRRADIATED IN THE ATR SOUTH FLUX TRAP

This Engineering Calculations Analysis Report (ECAR) documents the results of the Advanced Test Reactor (ATR) detailed physics analyses performed to calculate the displacements per atom (DPA) and the fast neutron fluence (E > 0.1 MeV) of the Advanced Graphite Creep (AGC) experiment, AGC-1, irradiated in the ATR South Flux Trap (SFT) (see Figure 1) during ATR Cycle 145A, Cycle 145B, Cycle 146A, Cycle 146B, Cycle 147A, Cycle 148A and Cycle 148B. The results for these evaluations and analysis are reported herein.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

ECAR-2291 AS-RUN PHYSICS ANALYSIS FOR THE AGC-2 EXPERIMENT IRRADIATED IN THE ATR

This Engineering Calculations Analysis Report (ECAR) documents the results of the Advanced Test Reactor (ATR) detailed physics analyses performed to calculate the displacements per atom (DPA) and the fast neutron fluence (E > 0.1 MeV) of the Advanced Graphite Creep (AGC) experiment, AGC-2, irradiated in the ATR South Flux Trap (SFT) (see Figure 1) during ATR Cycle 149A, Cycle 149B, Cycle 150B, Cycle 151A, and Cycle 151B. This ECAR also reports the neutron and photon heat rates for the materials of the AGC-2 experiment for ATR Cycle 149B. The results for these evaluations and analysis are reported herein.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Correlation of As-read and As-run Dosimetry Data

The Nuclear Science Users Facilities (NSUF) is responsible for the coordination of experiments conducted in Advanced Test Reactor (ATR) by Universities and other outside organizations. Completion of the experiment and data reporting process is long and expensive. As such, it is expected that targets for temperature and neutron dosage are met. Simulations of expected reactor performance under specific conditions to meet these targets are critical for this process. However, it is not often that these reports are correlated with the data produced by monitors in ATR. This project aims to correlate simulation and read data from experiments in ATR to gauge uncertainty.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

AGC-4 As-Run Uncertainty Analysis

The concept of measuring neutron flux (or fluence rate) with activation foils has existed since the late 1940s and is a well-used methodology. Activation foils are selected for their unique isotopic cross sections that create reaction products with desirable properties: relatively simple radioactive decay scheme, measurable quantities of gamma-rays or betas, and half-lives appropriate for the irradiation and decay durations of the experiment. The activation materials, deemed flux wires, fluence wires, dosimeters, or flux foils, are fabricated to have a mass appropriate for the irradiation as well so the measurement using a radiation detector following the irradiation is sufficient but not overwhelming in terms of total activity. Activation measurement results are expressed in terms of reaction rate, which can then be used to “adjust” the neutron spectrum. The Advanced Graphite Creep (AGC)-4 experiment contained 26 flux wire packages, each containing an iron and a niobium wire inside a vanadium capsule. For all stackup positions, the value of fast to thermal neutrons is higher than expected, but the total neutron fluence is near the predicted value of ~2 x 1022 n/cm2. As compared to the predicted neutron spectrum, there are four takeaways in the results as compared to the predicted ATR irradiation: 1) Additional fast (fission) neutrons present, 2) Spectrum hardening due to temperature, 3) Spectrum hardening due to container attenuation, 4) Total neutron fluence is close to predicted values. Since the neutron energy spectrum is a key parameter when attributing neutron damage in materials, the adjusted energy-dependent spectrum and the total neutron fluence results from the dosimeters can be incorporated for future calculations.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

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

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

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

AGR-5/6/7 Thermal Model with Non-uniform Gas Gaps

The daily as-run thermal analysis for the Advanced Gas Reactor (AGR)-5/6/7 experiment was documented in ECAR-5633 (Hawkes 2021). A finite element model was created for the entire test train, which consisted of five capsules. The fuel compacts, graphite holders, stainless-steel capsule walls, and all other major components were individually modeled, including all thermocouples (TCs). Daily heat rates for each compact and component in the model were input from daily as-run physics analyses. Daily gas compositions for each capsule were also input. The thermal conductivity of the compacts and graphite holders varied with fast neutron fluence. Gas mixture thermal conductivity was implemented using experimentally attained values from the literature. Fluence and temperature-dependent thermal conductivity was used for the graphite components and the fuel compacts. Radiation heat transfer was implemented. The model was tuned in an attempt to match the TC readings during the first cycle by adjusting the Neolube (graphite lubricant) thickness.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2018

James Webb Space Telescope (JWST) Optical Telescope Element and Integrated Science Instrument Module (OTIS) Cryogenic Test Program and Results

In 2017, the James Webb Space Telescope Optical Telescope Element and Integrated Science Instrument Module (OTIS) underwent cryogenic optical testing at the Johnson Space Center. In this paper, we summarize the successful execution and results of this 100-day test, which was a major program milestone. We summarize the as-run test configuration and provide a top-level as-run timeline. We also provide the top-level functional, optical, thermal, and operational results from the test. We summarize the key technical issues encountered and the resolution of those issues. The results of the OTIS test indicate that the payload should be fully capable of delivering on JWST's exciting scientific potential.

Kimble, Randy A.↗
DOE OSTI2021

An uncertainty quantification method relevant to material test reactors

Within material test reactor calculations, energy dependent flux and reaction rate uncertainties are typically not quantified when performing as-run analyses to determine the neutron field experienced by the experiment. When high fidelity Monte-Carlo codes are used in such analyses, straight forward methods to calculate output uncertainties are not available, instead expert opinion is used to postulate computational uncertainties. New methods to propagate uncertainties through these high fidelity simulations are available when sufficient computational power is available. A tool is developed here for sampling any part of an MCNP input from random distributions to determine output uncertainties based on those inputs. Another tool is developed to sample nuclear data cross-section in ACE format using multi-group nuclear data covariances. The Total Monte-Carlo Method and Gesellschaft für Anlagen-und Reaktorsicherheit method (GRS) are implemented and compared to one another as well as MCNP sensitivity and uncertainty calculations. The methods were applied to the Godiva critical sphere k-eigenvalue, the UAM pin-cell benchmark energy dependent flux and reaction rates, and the Advanced Test Reactor energy dependent flux within an experimental location. Furthermore, the two methods agree well, with GRS allowing for an order of magnitude speedup for reaction rate uncertainty calculations and several orders of magnitude for eigenvalue uncertainty calculations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Californium-252 production at the High Flux Isotope Reactor - I: Validation study using campaign data

This paper presents a series of 252 Cf production validation and code-to-code comparison studies performed based on data from the production campaigns at the High Flux Isotope Reactor (HFIR). These studies support efforts to convert HFIR from using highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. HFIR must maintain its world-class performance and missions following this conversion, and because 252 Cf is a vital neutron-emitting radioisotope used for a variety of high-impact applications (e.g., reactor startup, cancer treatment), the ability to efficiently produce 252 Cf must be preserved. In this work, the HFIRCON, Shift, ORIGEN, and TCOMP codes were deployed, and several sets of data libraries were investigated to better understand the calculation codes and the data biases. As-loaded target composition data, as-run irradiation history data, and post-irradiation measurements from recent multi-cycle irradiation campaigns of the HEU core were used to validate and determine methodology biases. Further, the findings demonstrated a good agreement, with results falling within 3 standard deviations of measurements. This paper lays the ground work for the second paper, which evaluates and compares 252 Cf production and safety metrics with the HEU core and a proposed LEU core.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

FAST irradiations and initial post irradiation examinations – Part I

The Advanced Fuels Campaign Fission Accelerated Steady-state Test (FAST) at Idaho National Laboratory (INL) completed its first irradiation cycle within the Advanced Test Reactor (ATR). The test focused on the irradiation of alloy fuel forms for use in sodium fast reactors. Tests included the use of fuel additives (Sn, Sb, and Pd) for reducing the gettering of lanthanide fission products, zirconium liners to mitigate fuel-cladding chemical interaction, and annular fuel geometry for sodium-free fuel designs. The first cycle of FAST testing was completed and a series of rodlets were removed for the initial post irradiation examination (PIE) of FAST rodlets. The rodlet irradiation conditions was evaluated using Monte Carlo N-Particle (MCNP) for as-run power history and COMSOL for temperature analysis. These rodlets include a set of low burnups (~2.5 % heavy metal [%FIMA]) control rodlets and a helium bonded annular rodlet (4.7 %FIMA). Initial non-destructive PIE has been completed and includes, visual inspection, neutron radiography and gamma scanning of the FAST rodlets. Radiography confirmed the integrity of the experiments as well as showing the complete filling of the annulus in the annular fuel at a modest burnup (4.7 %FIMA) and potential slumping of the cooler rodlets at lower burnup. Precision gamma scanning indicated mostly usual fission product behavior with the exception of the ceasium in the He-bonded annular fuel. Future destructive PIE will be necessary in order to fully understand the effects of accelerated irradiation on U-Zr metallic fuel behavior.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1993

GEOTAIL Spacecraft historical data report

The purpose of this GEOTAIL Historical Report is to document ground processing operations information gathered on the GEOTAIL mission during processing activities at the Cape Canaveral Air Force Station (CCAFS). It is hoped that this report may aid management analysis, improve integration processing and forecasting of processing trends, and reduce real-time schedule changes. The GEOTAIL payload is the third Delta 2 Expendable Launch Vehicle (ELV) mission to document historical data. Comparisons of planned versus as-run schedule information are displayed. Information will generally fall into the following categories: (1) payload stay times (payload processing facility/hazardous processing facility/launch complex-17A); (2) payload processing times (planned, actual); (3) schedule delays; (4) integrated test times (experiments/launch vehicle); (5) unique customer support requirements; (6) modifications performed at facilities; (7) other appropriate information (Appendices A & B); and (8) lessons learned (reference Appendix C).

Boersig, George R.↗
DOE OSTI2021

AGR-5/6/7 Irradiation As Run Predictions Using PARFUME

The PARticle FUel ModEl (PARFUME), a fuel performance modeling code used for high-temperature gas-cooled reactors, was used to model the Advanced Gas Reactor (AGR)-5/6/7 irradiation test using as-run physics and thermal data. The AGR-5/6/7 irradiation test consists of the combined fifth, sixth, and seventh planned irradiations of the AGR Fuel Development and Qualification Program. The AGR-5/6/7 test train is a multi-capsule, instrumented experiment that is designed for irradiation in the 133.4-mm diameter northeast flux trap position of the Advanced Test Reactor (ATR) at Idaho National Laboratory. Each capsule contains compacts filled with uranium oxycarbide unaltered fuel particles. This report documents the calculations performed to predict the failure probability of tristructural isotropic (TRISO)-coated fuel particles during the AGR-5/6/7 experiment. In addition, this report documents the calculated fission product release fraction from the fuel. The calculations include modeling of the AGR 5/6/7 irradiation that occurred from February 2018 to July 2020 over nine ATR cycles, including six normal cycles and three power axial locator mechanism cycles, for a total of approximately 376 effective full power days (EFPD). The irradiation conditions and material properties of the AGR-5/6/7 test predicted zero fuel particle failures in Capsules 1, 3, and 4. Fuel particle failures were predicted in two of the compacts in Capsule 2 and one particle failure is predicted in each one of the compacts in Capsule 5. All compacts that exhibited fuel particle failures predicted by PARFUME were caused by localized stress concentrations in the silicon carbide (SiC) layer caused by cracking in the inner pyrolytic (IPyC) layer. In addition, shrinkage of the buffer and IPyC layer during irradiation resulted in formation of a buffer-IPyC gap. Compacts with a lower irradiation temperature and fluence experienced the smallest buffer-IPyC gap formation. Conversely, higher irradiated temperature compacts with a high fluence experienced the largest buffer-IPyC gap formation. Compact 3-6-3 experienced the largest buffer IPyC gap formation of just under 21.7 µm. The release fraction of fission products silver (Ag), cesium (Cs), and strontium (Sr) vary depending on capsule location and irradiation temperature. The maximum release fraction of Ag occurs in Capsule 3, reaching up to 59.5% for the TRISO fuel particles (compact 3-6-3). The release fraction of the other two fission products, Cs and Sr, are much smaller. A maximum Cs release fraction of 1.1% occurred in compact 3-4-3 and 4.4% for Sr in compact 3-6-3.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

ECAR-1944 AGC-1 AS RUN THERMAL RESULTS

Purpose of this ECAR is to provide documentation of as-run analysis results for each specimen temperature in the AGC-1 test train at specific time points during the irradiation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Pu-238 Production Calculator and Estimation Tool in the Advanced Test Reactor

A MS Excel tool was created to help track Pu-238 production, Np-237 feedstock use, estimate production scenarios, and coordinate planning for shipments at the Idaho National Laboratory (INL) and Advanced Test Reactor (ATR). 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.

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 OSTI2024

FUEL PERFORMANCE SIMULATION OF HIGH BURNUP FUELS IN PLANNED INTEGRAL DESIGN BASIS ACCIDENT EXPERIMENTS

High burnup (HBu) fuel rods from the Byron Nuclear Generating Station (BNGS) were recently received at Idaho National Laboratory (INL) to support a variety of planned Nuclear Energy fuel cycle R&D objectives ranging from fuel performance, fuel recycle, and spent fuel research topics. Among these R&D activities, these fuel rods will be the subjects of multiple in-pile experiment programs at the Transient Reactor Test (TREAT) facility as well as detailed characterization and testing in the hot cells at INL and Oak Ridge National Laboratory (ORNL). TREAT RIA experiments are planned for the Nuclear Energy Agency Framework for Irradiation Experiments (FIDES) Joint Experimental Program called High burnup Experiments in Reactivity Initiated Accident (HERA) program. TREAT and ORNL-furnace LOCA experiments are part of the Department of Energy (DOE) Advanced Fuels Campaign (AFC) program U.S. consensus LOCA test plan, and the in-pile experiments have also been proposed in a FIDES project called Loss of Coolant-High Burnup (LOC-HBu). The results of these test programs will provide crucial data about safety performance enabling extended licensable burnup limits for these fuels. The purpose of this paper is to document fuel performance computational simulations of the BNGS fuel using the Bison code. The detailed assessments include (1) the irradiation history of the fuel to provide prediction of as-run fuel conditions and (2) extending the irradiated fuel conditions into the as-designed experiment conditions for the HERA-HBu RIA experiments and for the LOC-HBu LOCA experiments. The results of these assessments will inform post-irradiation examinations (PIE) of the BNGS parent rods and detailed final design of the planned experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

FUEL PERFORMANCE SIMULATION OF HIGH BURNUP FUELS IN PLANNED INTEGRAL DESIGN BASIS ACCIDENT EXPERIMENTS

High burnup (HBu) fuel rods from the Byron Nuclear Generating Station (BNGS) were recently received at Idaho National Laboratory (INL) to support a variety of planned Nuclear Energy fuel cycle R&D objectives ranging from fuel performance, fuel recycle, and spent fuel research topics. Among these R&D activities, these fuel rods will be the subjects of multiple in-pile experiment programs at the Transient Reactor Test (TREAT) facility as well as detailed characterization and testing in the hot cells at INL and Oak Ridge National Laboratory (ORNL). TREAT RIA experiments are planned for the Nuclear Energy Agency Framework for Irradiation Experiments (FIDES) Joint Experimental Program called High burnup Experiments in Reactivity Initiated Accident (HERA) program. TREAT and ORNL-furnace LOCA experiments are part of the Department of Energy (DOE) Advanced Fuels Campaign (AFC) program U.S. consensus LOCA test plan, and the in-pile experiments have also been proposed in a FIDES project called Loss of Coolant-High Burnup (LOC-HBu). The results of these test programs will provide crucial data about safety performance enabling extended licensable burnup limits for these fuels. The purpose of this paper is to document fuel performance computational simulations of the BNGS fuel using the Bison code. The detailed assessments include (1) the irradiation history of the fuel to provide prediction of as-run fuel conditions and (2) extending the irradiated fuel conditions into the as-designed experiment conditions for the HERA-HBu RIA experiments and for the LOC-HBu LOCA experiments. The results of these assessments will inform post-irradiation examinations (PIE) of the BNGS parent rods and detailed final design of the planned experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗