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Results for “Advanced Fuels Campaign (AFC)”

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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49 records · Page 3
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DOE OSTI2022

Post-Irradiation Fracture Toughness Characterization of Generation II FeCrAl Alloys

FeCrAl alloys are promising candidate materials for the accident tolerant fuel (ATF) cladding application due to their exceptional resistance to oxidation in elevated temperature steam environments. Currently, limited fracture toughness data are available for the FeCrAl alloys, including the FeCrAl alloys newly developed at Oak Ridge National Laboratory (ORNL) under the U.S. Department of Energy’s Advanced Fuels Campaign (AFC) program. In this study, two Generation II candidate FeCrAl alloys, i.e., C06M (81.8Fe-10Cr-6Al-0.03Y-2Mo-0.2Si) and C36M (78.8Fe-13Cr-6Al-0.03Y-2Mo-0.2Si), were irradiated in the High Flux Isotope Reactor (HFIR) at ORNL to assess the fracture characteristics of these alloys after neutron irradiation. A total of six rabbit capsules were irradiated in HFIR at target temperatures of 200°C, 330°C, and 500°C up to target damage doses of 8 displacements per atom (dpa) and 16 dpa. Post-irradiation fracture toughness testing was performed following the Master Curve method in the ASTM E1921 standard. The main findings of this study are:1) Both the C06M and C36M alloys exhibited a similar response to irradiation concerning irradiation hardening and embrittlement.2) The irradiation temperature played different roles in terms of irradiation hardening and embrittlement for both C06M and C36M: after irradiation between 166°C and 204°C, both materials exhibited significant irradiation hardening and embrittlement; after irradiation between 315°C and 343°C, both materials showed small irradiation hardening without irradiation embrittlement. After irradiation between 501°C and 507°C, however, the irradiation softening without irradiation embrittlement was observed in both materials.3) Comparing the microhardness and Master Curve reference temperature T0q before and after neutron irradiation, we did not observe a linear correlation between the two parameters for both C06M and C36M steels. This should be mainly due to a flat response of the Master Curve reference temperature T0q to the irradiations at 166–204°C and 315–343°C ranges4) C06M showed a lower T0q, meaning better toughness, than C36M at the unirradiated condition, and such trend was kept even after neutron irradiation except for the 166–204°C irradiation after which both materials had similar T0q.5) In terms of hardening and embrittlement, the irradiation effect on both C06M and C36M appeared to saturate after an irradiation dose of 7 dpa.

Chen, Xiang↗
DOE OSTI2022

Comment Response for the Draft Combined TREAT-LOC & SATS Integral LOCA Experiment Plan

This document provides a detailed description of comments and responses, to the Advanced Fuel Campaign (AFC) loss-of-coolant accident (LOCA) experiment plan draft document shared in April 2022. Comments were provided from reviewers from U.S. nuclear fuel vendors, the Electric Power Research Institute (EPRI), the U.S. Nuclear Regulatory Commission (NRC), and the U.S. Department of Energy (DOE) all provided comments after solicitation. All comments have been carefully considered and responded to accordingly. Editorial and clarification comments are excluded from this writeup, but all have been addressed directly in the text of the revised plan document.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Thermal-hydraulic and Fuel Performance Scoping Studies of a Flowing Water Capsule in TREAT

The restart of the Transient Reactor Test (TREAT) facility has provided a much needed capability for integral transient testing of nuclear fuel. This testing is necessary to qualify new fuel concepts such as those developed under the Accident Tolerant Fuel (ATF) program, increasing the burnup limit of light water reactor (LWR) fuels, or a variety of other programs under the Advanced Fuels Campaign (AFC). The ATF campaign has been the main driver behind the development and implementation of a variety of capsules for TREAT experiments. The Separate-Effect Test Holder (SETH) was a dry capsule that enabled testing of ATF concepts under reactivity-initiated accident (RIA) heating conditions. Following SETH, the Static Environment Rodlet Transient Test Apparatus (SERTTA) was developed to enable RIA testing in a static water environment. In efforts to support the need for future Loss-of-Coolant Accident (LOCA) tests, the Transient Water Irradiation System for TREAT (TWIST) capsule has been developed that allows for water to drain from around the fuel rod and lower the pressure to simulate LOCA conditions. All these capsules that have been developed for LWR fuel testing all lack the capability for forced convective cooling which in some applications may limit their ability to test under prototypic conditions. A design and modeling effort has been started to modify the TWIST capsule by adding a flow tube and impeller that can create flowing coolant conditions for the fuel rod. RELAP5-3D and BISON models are being used to study the differences between RIA, LOCA, Anticipated Operational Occurrences (AOO), LWR power cycling, and other transient scenarios under flowing conditions capable in the flowing capsule and the current stagnant water capsules (SERTTA and TWIST). The scoping study will provide guidance on the needed capabilities for the flowing capsule and the limitations of the currently developed capsule for LWR testing in TREAT.

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↗
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↗
DOE OSTI2026

Defining Probe Closure Offset and Scan Stitching Methodology for Element Contact Profilometry in the Hot Fuel Examination Facility

The Element Contact Profilometer (ECP) is a non-destructive post-irradiation examination (PIE) method established in the Materials and Fuels Complex (MFC) Hot Fuel Examination Facility (HFEF). The ECP suspends a fuel rod with a standard end-fitting in a vertical orientation and draws it between two opposed Sony probes with sapphire rod tips, which measure the diameter directly as a function of vertical stage position through contact profilometry. The ECP is equipped with concave rollers, which pinch the rod to maintain a consist position between these sapphire probe tips. A schematic of the analysis head is shown in Figure 1. The gripper that suspends the rod end-fitting is able to rotate to allow measuring of any azimuthal angle around the rod. This system is used to characterize the diameter of full-length light water reactor (LWR) irradiated fuel rods at multiple azimuthal angles in support of the Advanced Fuels Campaign (AFC).

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Projection Analysis for ATR Irradiation of the AFC-FAST Experiment

Analyses of the Advanced Fuels Campaign Fission Accelerated Steady-state Test (AFC-FAST) in the Advanced Test Reactor are presented. A detailed methodology was employed to better account for uncertainties in the planned power and duration of sequential reactor loading cycles. By performing coupled depletion analyses at multiple power levels and durations, the differences in experiment heating outputs can be found. The effects of these uncertainties upon multiple experiment configurations were assessed in an effort to streamline the process of planning for and documenting future irradiations. In conclusion, the data generated from this work have been used to help inform assumptions on subsequent projections to perform only a nominal case depletion.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

EDS Analysis of FCCI in AFC-FAST Fuel Pins

The Advanced Fuel Campaign’s Fission Accelerated Steady-state Testing (FAST) program uses metallic fuel pins with small diameters to reach a desired burnup more quickly. This enables accelerated testing of advanced fuel designs and decreases the time between idea conception and commercial usage. Post-irradiation examination is critical in this process, especially with respect to the fuel-cladding chemical interactions (FCCI). In this work, energy dispersive X-ray spectroscopy (EDS) is used to track how elements from the fuel and fission products have diffused through the cladding. Elemental redistribution along the fuel-cladding interface is mapped and FCCI region thicknesses are measured. The correlations between geometry, temperature, burnup, and FCCI thickness are presented.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

AFC Newsletter (February 2022)

No abstract. This is a monthly newsletter containing activities from the Advanced Fuels Campaign for February 2022.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

AFC Newsletter (January 2022)

No abstract for this document. Document will be shared with internal and external AFC collaborators.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Post-irradiation characterization of a high burnup mixed oxide fuel rod with minor actinides

The Advanced Fuels Campaign performed a series of irradiation tests of minor actinide-bearing mixed oxide fuel (MA-MOX), the so-called AFC-2C&D experiments, to investigate the transmutation of long-lived transuranic actinide isotopes contained in spent nuclear fuel via fast reactor technology at burnups exceeding 10 % fission of initial metallic atoms. This manuscript reports the test results derived from one of the five MA-MOX rodlets taken to higher burnup in the AFC-2D irradiation. This includes both non-destructive investigations, such as gamma and neutron spectrometry, and destructive investigations, such as fission gas release, ceramography, and chemical burnup analysis. In addition, the microstructure of the fuel was investigated using advanced electron microscopy techniques including electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). Here, it was observed with EBSD that the pellet had subdivision of the grains and the TEM observed migration of cladding material into the 5 metal precipitates in the fuel which could have been from the higher than desired oxygen/metal ratio. The TEM also showed an enrichment of Cr in fuel clad chemical interaction (FCCI) layer.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗