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Smith, Rebecca E

Publications and source records attributed to Smith, Rebecca E.

16 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Oxidation Penetration in Nuclear Graphite

Study results are presented for seven grades of graphite where 21 cylindrical samples were oxidized to a nominal level of mass loss. Low mass loss samples exhibited ~3% oxidative mass loss, intermediate ~8%, and high ~11%. Flat sample surfaces were covered during oxidation to minimize oxidation penetration at the top and bottom of each sample. Oxidized samples had their diameters reduced stepwise in 1 mm or 2 mm increments. Residual samples were weighed, geometric dimensions were recorded, and Archimedes measurements were taken at each step. Preliminary comparative graphical analysis is presented to illustrate the resultant density gradients observed. Raw tabular data are also provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Graphite Oxidation Activities

Graphite Oxidation Activities to include introduction, rate determination, strength after oxidation, penetration depth analysis, strategic partnership projects, and summary and continuing work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Oxidation Behavior of Irradiated and Unirradiated NBG-25 Graphite

This presentation discusses The xxidation behavior of irradiated and unirradiated NBG-25 Graphite results of thermogravimetric analysis of same-source specimens from the AGC-1 experiement. Minimal annealing effects, Substantial increase in OR with Irradiation using Conventional Rate Analysis, Apparent Dose Dependency, Examination of Onset (Alternate Analysis) Shows Competing Irradiation Effects, and Similar Test Matrix (without annealing) now under way with NBG-18 Graphite is reviewed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Milestone 1.2.14: Surface Characterization of Irradiated Surrogate Non-Native Corrosion Plumes on Aluminum 6061-T6 Specimens

The United States Department of Energy is seeking to evaluate the feasibility of extended dry storage of aluminum-clad spent nuclear fuel (ASNF) in helium-backfilled cannisters. A significant research effort has been devoted to determining the amount of radiation-induced molecular hydrogen generation from corrosion layers that would be present on the fuel assembly surfaces. However, to date, no evidence of radiation-induced changes in the microstructure of the corroded aluminum surfaces have been reported. This report provides surface characterization, in terms of corrosion layer composition and morphology, for aluminum alloy 6061-T6 “plume” samples subjected to one of three potential ASNF drying techniques—“vacuum only”, “vacuum + 4 hours at 100 oC”, and “vacuum + 4 hours at 220 oC”—and gamma irradiated (= 53 MGy) in the presence of helium gas with ~0% added relative humidity. At these high absorbed gamma doses, for the first time ever, radiation-induced circular defects were found on the sample surface, regardless of the drying regime employed. Additionally, large cracks that penetrated through the corrosion layer to expose bare aluminum metal were observed in irradiated “plume” specimens subjected to drying conditions of “vacuum + 4 hours at 220 oC”. The nature and implications of these defects on the extended dry storage of ASNF is unclear, requiring further study.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Oxidation Activities

Slides for Oxidation Activities, oxidation rate, penetration/lathing, and strength after oxidation work. Also includes recent and current oxidation studies, background for rate behavior, strength behavior, density profile, and penetration measurements, oxidation resistant graphite coating development, and irradiated and unirradiated graphite oxidation rate behavior.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Comparison of Irradiated and Unirradiated Graphite Oxidation Performance

This work examines the oxidation behavior of NBG-25 graphite using irradiated specimens and (not irradiated) companion specimens from the Advanced Graphite Creep (AGC) Experiment. Irradiated and companion specimens were quartered, split into four small samples, to enable four oxidation test runs for each (0.5 inch diameter by 0.25 inch tall) piggyback button obtained from the AGC Experiment inventory. These split samples were oxidized in air in a thermogravimetric analyzer (TGA) and benchmarked against observations with separately sourced same-grade specimens of three geometries oxidized either in the TGA or in a vertical furnace built and operated to satisfy the specifications of ASTM D7542. The study considers both irradiation damage and relief of damage by thermal annealing. A range of isothermal oxidation temperatures were tested for specimens exposed to a similar irradiation environment, nominally 6.5 dpa at 650°C. To assess the dose dependency of observed oxidation behavior, specimens with a range of irradiation exposures (up to 6.8 dpa) were tested at the single oxidation temperature of 650°C. In the conventional analysis (rate determined over the 5-10% mass loss range) any annealing effects appear to be negligible, while irradiation to ~6.5 dpa may double or triple the subsequent oxidation rate. However, examination of rate with extent of reaction (over incremental ranges from initial onset up to 5% mass loss) illustrates that (at least for ~6.5 dpa) irradiation initially inhibits oxidation. Slower rates of oxidation than companion split samples are clearly indicated across all oxidation temperatures tested up to ~0.5% mass loss. Over the range of 0.5-1% mass loss, regardless of normalization strategy, there is no statistically meaningful difference in oxidation rate. However, beyond 0.5% mass loss oxidation of the irradiated sample becomes progressively faster than the companion sample (up to the 10% mass loss level observed).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Performance of Graphite Oxidation with Environment and Specimen Geometry

Introduction -Motive for the study (a modest request) •Issues that must be considered -Pertinent oxidation background information •Three oxidation regimes -ASTM D7542 • The primary parameters for oxidation control -Energy of the System -Atmosphere -Reaction Interface

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Graphite Oxidation Studies Irradiated Graphite, GIF Summary, ASME Task Force, ASTM Activities

Introduction Comparison of Graphite Oxidation Rates for Irradiated and Unirradiated NBG-25 (Rebecca Smith and Austin Matthews, INL) Annealing effect over range of oxidation temperatures Irradiation effect (~6.5 dpa) over range of oxidation temperatures Dose dependency of irradiation effect Generation IV International Forum High Level Deliverable Summary of DOE ART Historic Oxidation Research (Cristian Contescu, ORNL) Publications issued over the last ~10 years Acute and chronic oxidation ASME Graphite Oxidation Task Force (C. Contescu, Paul Ryan, ORNL, Joe Bass, A. Matthews, R. Smith, INL) Collaborative effort headed by ORNL and engaging INL and others Initial review of relevant oxidation discussion in ASME code ASTM Activities (C. Contescu) Virtual Meetings New and Revised Standards

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

ASNF Engineering Scale Drying Experiment System Operability Functional Requirements

This checklist specifies the functional requirements for the drying system installed at Holtec International under the scope of INL Contract # 215725, in collaboration with University of South Carolina under INL Contract Number 207046, consistent with the parameters described in INL/EXT-19-56017, Aluminum-clad Spent Nuclear Fuel Engineering Scale Drying Experiment Design. Completion of the ASNF Engineering Scale Drying Experiment Acceptance Criteria for Fabricated Items, INL/MIS-20-58551, is a prerequisite.

36 MATERIALS SCIENCE↗
DOE OSTI2020

ASNF Engineering Scale Drying Experiment Acceptance Criteria for Fabricated Items

This document specifies the acceptance criteria for items fabricated by Holtec International under the scope of INL Contract # 215725, consistent with the parameters described in INL/EXT-19-56017, Aluminum-clad Spent Nuclear Fuel Engineering Scale Drying Experiment Design. Verification is confirmed with initials (or the need for remediation may be identified) along with the date of inspection for each requirement. Verification may be performed by independent proxy (University of South Carolina participant, under INL Contract Number 207046) or by an INL employee, remotely by camera images.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Completion of Level 3 Milestone M3AT-20IN0305040333

This item is an interoffice memo documenting the completion of of Level 3 Milestone M3AT-20IN0305040333. The status and progress made in the studies on irradiated graphite using thermogravimetric analysis (TGA) for isothermal oxidation of irradiated buttons, non-irradiated companions, and same-grade specimens are detailed.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Evaluation of Aluminum-clad Spent Nuclear Fuel during Drying and Dry Storage

The DOE SNF Working Group ASNF Subgroup and the U.S. Nuclear Waste Technical Review Board have identified knowledge gaps regarding the management of aluminum-clad spent nuclear fuel (ASNF).[1, 2] The objectives of this research are to observe, sample and analyze the nature of aluminum surface corrosion and corrosion layer chemistry relevant to the ASNF performance over the duration of the dry storage segment of the fuel life cycle. The primary goal is to build an understanding of ASNF behavior, from reactor service and wet storage through drying and dry storage, with parallel testing of surrogate materials. The premise is that residual water affiliated with the aluminum oxide from corrosion during reactor service and wet storage, continues to influence ASNF performance during dry storage, and that even after successful drying the surface conditions of ASNF have yet to be satisfactorily characterized. The initial scope collected and analyzed specimens from irradiated non-fuel components of aluminum-clad Advanced Test Reactor (ATR) fuel elements that are routinely removed at the end of reactor service life. Techniques have been developed for remote material sampling from ATR elements from vented dry storage at the Irradiated Fuel Storage Facility (IFSF). These techniques involve 1) nipping off a small corner of the aluminum near where the end box had been removed and 2) scraping the exterior (non-fueled) plate surface to collect an oxide sample. This second effort targets ATR elements expected to have seen the longest residence in storage, both in the CPP-603 pool (with poor water quality) and in dry storage at IFSF, within the fuel handling scheduled to consolidate ATR at IFSF and receive the remaining inventory at CPP-666. Follow on work has been proposed to nip corners and collect scrapings immediately following ATR fuel drying, before placement in dry storage. Sample analysis will use multiple techniques to understand the chemical composition and morphology of the surface oxides. A simultaneous effort is progressing to develop surrogate materials to enable rigorous controlled experiments. Consistently produced, these surrogate materials can then be used to demonstrate limitations inherent to remote sampling and to validate performance models under relevant constraints. The combined efforts will substantiate ASNF performance and fuel management decisions regarding drying and packaging over a more complete range of life cycle conditions. [1] DOE Spent Nuclear Fuel Working Group, Aluminum-Clad Spent Nuclear Fuel: Technical Considerations and Challenges for Extended (>50 Years) Dry Storage, DOE/ID RPT#: 1575, prepared by Aluminum-Clad SNF Sub Working Group, June 2017. [2] U.S. Nuclear Waste Technical Review Board, Management and Disposal of U.S. Department of Energy Spent Nuclear Fuel, A Report to the United States Congress and the Secretary of Energy, December 2017, p. 143.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Comparison of Oxidation Performance of Graphite Samples in TGA versus Vertical Furnace

Oxidation influences the working life of nuclear graphite and can impact function during or after air (or steam) ingress accidents. Temperature, gas flow, oxygen content, and dimensional constraints at the microstructural and macrostructural levels can substantially influence the oxidation rate. The nature of these constraints must be considered when attempting to extrapolate performance from experimental results to service applications, but must also be evaluated in order to compare data among disparate experimental systems. This work compares oxidation test performance from a vertical furnace system (functioning as described by ASTM D7542) with tests using a thermogravimetric analyzer (TGA) adapted to operate in an analogous manner.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

History and Status of DOE’s Standardized Canister – 19657

The U.S. Department of Energy (DOE) is responsible for managing spent nuclear fuel (SNF) that is currently in, or will in the future come into, its possession. The DOE standardized canister will be relied upon to confine radionuclides and preclude moderator in order to ensure safe, long-term management and disposition of DOE SNF. This strategy offers an avenue for avoiding the characterization costs (i.e. funds and exposure) and other technical challenges associated with the traditional approach of relying on fuel-specific properties as the basis for demonstrating compliance with the regulator. The use of the standardized canister offers the following benefits: • The standardized canister design is complete. It has been tested under a variety of drop and accident scenarios and was subsequently licensed under 10 CFR 72 as a storage system as part of the FWENC Facility. It is also part of the license submittal for Yucca Mountain. • There is high confidence that the standardized canister design can be stored within commercially available storage casks and licensed as part of a transportation-system package. • The robustness of the standardized canister simplifies accident analyses and does not rely on characterization of the contained waste. This simplifies safety analysis, provides a greater level of confidence with stake holders and reduces cost and risk. • Significant savings result from a common design for the associated packaging, closure, and loadout facilities. • Design efforts, procurement, and fabrication are simplified • Safety and surety of operations are enhanced by standardizing handling activities and associated training, procedures, equipment, and tooling. The standardized canister is the preferred package for these materials and has withstood scrutiny by the Yucca Mountain Repository License Application, the NRC’s Advisory Committee on Nuclear Waste, the Nuclear Waste Technical Review Board, an American Society of Mechanical Engineers peer review, and the Foster Wheeler Environmental Corporation storage license application. This evaluation summarizes and documents the status of the design and rationale for the DOE standardized canister and outlines the remaining needs to allow deployment. In addition, it discusses the factors that have the potential to influence storage, transportation, and disposal of DOE SNF.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

History and Status of DOE’s Standardized Canister

[Poster] The U.S. Department of Energy (DOE) is responsible for managing spent nuclear fuel (SNF) currently in its possession, as well as SNF that will come into its possession in the future. DOE will rely upon the standardized canister to confine radionuclides and precluding the moderator in order to ensure safe, long-term management and disposition of SNF in its possession. This strategy helps minimize characterization costs (i.e., funds and exposure) and other technical challenges associated with the traditional approach of relying on fuel-specific properties to demonstrate compliance with the regulator. The use of the standardized canister offers the following benefits: • The standardized canister design is complete. It has been tested under a variety of drop and accident scenarios and was subsequently licensed under 10 CFR 72 as a storage system as part of the Foster Wheeler Environmental Corporation (FWENC) Facility. It is also part of the license submittal for Yucca Mountain. • There is high confidence that the standardized canister design can be stored within commercially-available storage casks and licensed as part of a transportation-system package. • A robust standardized canister simplifies accident analyses and does not rely on characterization of the contained waste. This simplifies safety analysis, provides a greater level of confidence with stakeholders and reduces cost and risk. • A common design for the associated packaging, closure, and loadout facilities results in significant savings. • The DOE standardized canister would simplify design efforts, procurement, and fabrication. • Standardizing handling activities and associated training, procedures, equipment, and tooling leads to enhanced safety and surety of operations. The standardized canister is the preferred package for these materials. It withstood scrutiny by the Yucca Mountain Repository License Application, the Nuclear Regulatory Commission’s (NRC) Advisory Committee on Nuclear Waste, the Nuclear Waste Technical Review Board, an American Society of Mechanical Engineers peer review, and the FWENC storage license application. This evaluation summarizes and documents the status of the design and rationale for the DOE standardized canister, and outlines remaining needs before deployment. In addition, it discusses the factors that have the potential to influence storage, transportation, and disposal of DOE SNF.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗