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At least 667 records · Page 37
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Sister Rod Nondestructive Examination Final Report

This report documents work performed under the Spent Fuel and Waste Disposition for the US Department of Energy (DOE) Office of Nuclear Energy (NE) Spent Fuel and Waste Science and Technology program. This work was performed to fulfill the Level 2 Milestone M2SF-17OR010201021, “Documentation of Non-destructive Tests on Sister Pins,” within work package SF-17OR01020102. The High Burnup Spent Fuel Data Project, sponsored by DOE-NE, is focused on understanding the effects of long-term storage and transportation on high burnup (HBU) (>45 gigawatt days per metric ton uranium) light water reactor fuel. The goals of the project are to “provide confirmatory data for model validation and potential improvement, provide input to future spent nuclear fuel (SNF) dry storage cask design, support license renewals and new licenses for Independent Spent Fuel Storage Installations (ISFSIs), and support transportation licensing for high burnup SNF". In support of project goals, 25 sister rods were removed from fuel assemblies at the North Anna Power Station. Nine rods were removed from the project assemblies, and 16 rods were removed from similar HBU assemblies. The 25 sister rods were shipped to Oak Ridge National Laboratory (ORNL) Irradiated Fuels Examination Laboratory (IFEL) in early 2016 for detailed nondestructive examination (NDE) and destructive examination (DE). The detailed examinations will provide essential information on the physical state of the HBU rods and the fuel contained in the rods prior to loading, drying, and long-term dry storage. Similar tests will be performed at the end of the long-term storage period to identify any changes in the fuel rods’ properties during the dry storage period.

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DOE OSTI2018

Aluminum Clad Spent Nuclear Fuel Task 2: Oxide Layer Radiolytic Gas Generation Resolution Experiment Test Plan

This document presents the test plan for Task 2 of the Aluminum Clad Spent Nuclear Fuel Long Term Dry Storage Technical Issues Action Plan. This task’s overarching goal is the resolution of gas generation resulting from the radiolytic degradation of the oxide layers on aluminum. The uncertainty associated with the build-up of gaseous radiolytic products inside the ASNF storage canisters is one of the dominant technical challenges associated with long term dry storage of aluminum clad fuel. The objectives of the Task 2 test plan are to: 1. Understand how radiation chemistry inside the ASNF dry storage canister impacts extended dry storage performance. 2. Understand the radiation induced mechanisms governing the production of radiolytic gases and alteration of gas composition within a canister. 3. Correlate the observed changes in material properties of ASNF with the radiolytic degradation of aluminum oxide in fill environments of variable moisture and gas composition. 4. Compile a predictive kinetic modeling tool, to provide quantitative mechanistic insight into the in-cask radiation chemistry of ASNF. 5. Interface with the canister modeling activity of Task 3 to deliver degradation rates for package performance of both sealed and vented scenarios.

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DOE OSTI2018

Oxyhydroxides on aluminum spent nuclear fuel: formation studies and removal practices to prevent radiolytic gas production

Aluminum-clad nuclear fuel has been at the forefront in research and commercial reactors. After being removed from the reactor, the spent nuclear fuel is placed in large-volume water basins (wet storage) that are constantly monitored for water quality to minimize corrosion/deterioration. The water provides both radiation shielding and cooling for the freshly removed fuel. It is highly desired to have a pathway for fuel to be transitioned to dry storage in sealed casks, so that spent nuclear fuel can be safely transported and dispositioned in a permanent repository, and basin capacity can be made available for temporary wet storage of fresher spent fuel. A challenge for safe, road-ready dry storage is the hydrated aluminum oxides (oxyhydroxides) that are prevalent on aluminum spent nuclear fuel that has been stored under water for years. If water is released from the oxyhydroxides during dry storage, it can create corrosion within the storage canister, and hydrogen can be generated through radiolysis of the oxyhydroxides, causing pressurization and flammability risks. The focus of this project is to develop knowledge of the oxyhydroxide growth and production of gases through radiolysis and to apply empirical models for predicting these phenomena. The expertise developed will be used to formulate conditions and parameters for oxyhydroxide removal/dehydration from fuel cladding through vacuum or forced-air gas drying, to prevent problems during permanent dry storage.

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DOE OSTI2015

Effects of Lower Drying-Storage Temperatures on the DBTT of High-Burnup PWR Cladding

The purpose of this research effort is to determine the effects of canister and /or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures and hoop stresses. Extensive precipitation of radial hydrides could lower the failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods stored in pools, at temperatures below the ductile- to-brittle transition temperature (DBTT).

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DOE OSTI2018

Validation of k eff Calculations for Extended BWR Burnup Credit

The US Nuclear Regulatory Commission and consensus standards recommend validation of the numerical methods used in criticality safety analyses. This validation requires the comparison of computational results with measurements of physical systems which are neutronically similar to those used in the safety analysis being performed. To this end, this document examines the methods available to generate sensitivity data to help identify systems similar to spent boiling- water reactor (BWR) fuel in a flooded spent nuclear fuel transportation or storage cask. A large number of critical benchmark experiments are surveyed using sensitivity/uncertainty (S/U) techniques to assess their applicability to BWR burnup credit (BUC) beyond the burnup of peak reactivity. Multiple burnups of BWR assemblies are considered herein, as well as the actinide-only (AO) and actinide-and-major-fission-product (AFP) isotope sets. Sample validations are completed for representative application models to demonstrate that appropriate validation is possible and to indicate the bias and bias uncertainty values expected for related applications.

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DOE OSTI2019

Transportation Planning for Irradiated Fuel-Cladding and Integral FeCrAl Experiments

Building the material database for accident tolerant fuel concepts requires movement of experiments to ORNL from other national laboratories, commercial reactors, and international collaborators. This report summarizes the current fleet of available and proposed casks for handling small scale fuel shipments and the immediate challenges to be addressed in order to perform shipments in support of accident tolerant fuels development under the Advanced Fuels Campaign. Shipping recommendations and a description of unresolved issues associated with the shipment are provided for three experiments: ATF-1 LOCA (ATF-18), the Halden irradiated Instrumented Fuel Assembly (IFA) 796 experiment, and the Hatch lead test rods. Each of these experiments represents an important step in continuing the accelerated testing of FeCrAl cladding for commercial applications.

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DOE OSTI2019

Considerations for Canister Design and Drop Testing

This paper discusses some considerations for the proposed design and analysis of canisters used to ship reactor debris from the Fukushima Daiichi nuclear plant. Energy-limited events, such as cask drops and other dynamic impacts, are evaluated much differently than other structures using static analysis. Static analysis of structures makes sense for engineers because static analysis is intuitive and material typically behaves in a linear elastic manner. Dynamic analyses are typically not intuitive because of non-linear material behavior. This is why canister designs need to be evaluated by dynamic finite element analysis and/or laboratory drop tests.

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DOE OSTI2017

Technical Feasibility of Direct Disposal of Electrorefiner Salt Waste

The experimental breeder reactor (EBR-II) used fuel with a layer of sodium surrounding the uranium-zirconium fuel to improve heat transfer. Disposing of this EBR-II used fuel in a geologic repository without treatment is not prudent because of the potentially energetic reaction of the sodium with water. In 2000, the US Department of Energy decided to treat the EBR-II sodium-bonded used fuel in an electrorefiner (ER), which produces a metallic waste, mostly from the cladding. The salt remaining in the ER contains most of the actinides and fission products. Two baseline waste forms were proposed for disposal in a mined repository; the metallic waste, which was to be cast into ingots, and the ER salt waste, which was to be further treated to produce a ceramic waste form. However, alternative disposal pathways for metallic and salt waste streams are being investigated that may reduce the complexity. For example, performance assessments show that both mined repositories in salt and deep boreholes in basement crystalline rock can easily accommodate the ER salt waste without treating it to form a ceramic waste form. Hence the focus of a direct disposal option, as described herein, is now on the feasibility of packaging the ER salt waste in the near term such that it can be transported to a repository in the future without repackaging. A vessel for direct disposal of ER salt waste has been previously proposed, designed, and a prototype manufactured based on desirable features for use in the hot cell. The reported analysis focused on the feasibility of transporting this proposed vessel and whether any issues would suggest that a smaller or larger size is more appropriate. Specifically, three issues are addressed (1) shielding necessary to reduce doses to acceptable levels; (2) the criticality potential and the ease which it can be shown to be inconsequential, and (3) temperatures of the containers in relation to acceptable cask limits. The generally positive results demonstrate that direct disposal of ER in the proposed packaging is feasible without the need to secure funding to modify the facility.

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DOE OSTI2019

High-Burnup Demonstration: Thermal Modeling of TN-32B Vacuum Drying and ISFSI Transients

This report describes thermal modeling done for the High-Burnup Spent Fuel Data Project that is being performed by the U.S. Department of Energy’s Spent Fuel and Waste Science and Technology research and development program. The purpose of this project is to investigate the performance of high-burnup spent nuclear fuel in dry storage. Part of this project is a demonstration test being performed with a storage module in the North Anna Nuclear Power Station’s Independent Spent Fuel Storage Installation (ISFSI). The storage module selected for this demonstration is an Orano TN-32B High-Burnup cask which has been loaded with fuel assemblies from the North Anna spent fuel pool.

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DOE OSTI2019

Fall 2019 Dissertation Update [Slides]

Summary of Presentation: Motivation - Problem Description; Feature Identification. Discussion of Features - Flat Flux Region Near Cask Centerline; Periodic Depressions of Flux in Fuel Region; Abrupt "Level-Off" Region; Non-exponential Reduction in Overpack; Asymmetric Flux. Future Work.

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DOE OSTI2019

Infiltration Data Package for the E-Area Low-Level Waste Facility Performance Assessment

The infiltration data package contains the input parameters, the cap design and material properties assumptions, and the modeling results for the Hydrologic Evaluation of Landfill Performance (HELP) infiltration model simulations performed in support of the E-Area Low-Level Waste Facility Performance Assessment (PA). The infiltration estimates establish the upper boundary condition for the PORFLOW vadose-zone model and GoldSim model simulations for the following E-Area disposal unit types: Slit and Engineered Trenches, Low-Activity Waste Vault (LAWV), Intermediate-Level Vault (ILV), Componentin- Grout (CIG) special waste form trench segments, and the Naval Reactor Component Disposal Areas (NRCDAs). The infiltration data package builds upon relevant, foundational PA technical reports and memoranda from the past 15 years and is supported by three important components: the HELP model input parameter datasheets, HELP model input and output filenames and directory structure, and infiltration rates as a function of time for each scenario for each disposal unit type. For Slit and Engineered Trenches, a single best-estimate Bahia-grass case of two-percent slope and 585- foot slope length represents an upper bound on intact closure-cap infiltration rates over a 10,000-year period. Building upon the single, bounding, and intact case, two different approaches were considered for incorporating the effect of localized cap subsidence on infiltration rates in the PORFLOW flow and transport model simulations. In the first approach, a Monte Carlo probabilistic model was developed to calculate infiltration rates for 0.54%, 2%, 3.6%, and 4.9% subsidence scenariosa reflecting historical and future non-crushable content for the Slit and Engineered Trenches. The resulting blended (spatially and slope-length averaged) infiltration rates become an inflow upper boundary condition for the PORFLOW vadose zone model simulations. In the second and preferred approach, a Monte Carlo rectangle packing algorithm (Danielson, 2019) and a simplified equation for the total mass flux of water draining into a subsided compartment (Equation 1) will be employed to implement a weighted blending of radionuclide fluxes to the water table using the results of deterministic PORFLOW vadose zone simulations representative of specific subsidence cases. Finally, HELP-model results from sensitivity studies of the intact infiltration case were fit to a log-logistic function to generate infiltration profiles over a 10,000-year period for most-optimistic, more-optimistic, best-estimate, more-pessimistic, and most-pessimistic cases, for both the intact and four subsidence scenarios. For the LAWV and IL vaults, infiltration rate profiles for a 10,000-year period were developed for both an on-vault (above the concrete/metal vault roof) and an off-vault (10-foot soil zone adjacent to vault walls) scenario. The purpose of the off-vault simulations was to confirm that subsurface runoff from the concrete vault roof will adequately drain through the lowermost backfill layers adjacent to the vault walls. The LAWV and ILV infiltration profiles assume collapse of the concrete roof at relative Years 2,905 and 7,100, respectively. The identical Bahia-grass case with two-percent slope and 585-foot slope length assumed for the Slit and Engineered Trenches will represent an upper bound on intact infiltration rates for the NRCDAs and CIG special waste form as well. While the NRCDA casks are assumed to remain structurally intact (i.e., non-assumed to collapse 300 years after the end of operations. Because both rows of the existing CIG trench segments in ST23 (formerly CIG01) extend across the planned cap crest, the subsided CIG trench segments are treated conservatively in the HELP model as a 100%-subsidence case with zero subsurface run-on. The CIG special waste form trench segment infiltration profile, therefore, consists of an intact-cap period from relative Year 100 to Year 300, followed by a 100%-subsidence period from relative Year 300 to Year 10,100.

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DOE OSTI2019

Update on Investigations of Viability of Cold Spray and FSW as a Spent Nuclear Fuel Dry Storage Canister Mitigation Tool

The purpose of this work is to investigate the use of cold spray and friction stir welding or processing (FSW/P) for repair and mitigation of chloride-induced stress corrosion cracking (CISCC) in dry cask storage system (DCSS) canisters to ensure integrity far beyond their original license period. This report provides a viability analysis of cold spray and friction stir welding for repair and fabrication of DCSS canisters. This report presents experimental work and a research strategy to optimize and evaluate cold spray and FSW/P for CISCC resistance. The literature reports that the following three conditions must exist for CISCC to occur: tensile stress, corrosive environment, and susceptible material. These conditions must be understood and accounted for while evaluating CISCC susceptibility. Furthermore, non-obvious factors, such as surface conditions and geometric confinement, can dramatically increase the intensity of one or more of the three listed conditions for CISCC. Arc welding should be considered for either temporary or permanent repair when combined with a mitigation technology. It is known that alloys used in the existing inventory of spent nuclear fuel storage canisters become sensitized in areas around fusion welds. Technologies, such as cold spray or FSW, can mitigate the sensitization caused by fusion welds. FSW/P and cold spray, potential alternatives for fabrication and repair, are solid-phase processes. No melting occurs and the energy input can be controlled such that no harm is done to the base metal during these processes. FSW/P should be considered for use in new fabrication of canisters and repair of existing canisters. Low heat input and grain refinement result in material with improved mechanical properties and comparable generalized corrosion performance relative to base metal. FSW/P offers the potential of through-thickness CISCC resistance. Cold spray should be considered for production, mitigation, and repair of canisters. Cold spray has demonstrated the ability to deposit stainless steel, Inconel alloys, and other alloys on stainless steel 304L such that galvanic potential is matched and resistance to pitting is improved. The technical basis for cold spray mitigation and repair within the overpack using remote robotic equipment is established. A non-obvious application to the research community is the sealing of crevices. All reported instances of CISCC in the fielded nuclear applications were associated with crevice corrosion. This can be avoided by cold spraying to seal crevices. The bulk of the work planned for FY 2020 will be generating coupons for CISCC testing by collaborative projects at Pacific Northwest National Laboratory and Sandia National Laboratories. First an exploratory study will be executed that will include screening studies and process optimization. This task will identify the effect of process parameters on microstructure and CISCC resistance. Optimized parameters will be used to produce a smaller number of coupons for full testing and evaluation. Test methods for evaluating CISCC resistance must meet the following criteria: (1) normalized surface geometry and exposed process microstructure, (2) account for residual stresses, (3) be in a controlled and appropriate test environment, and (4) be numerically quantifiable and repeatable. Failure to meet these criteria could lead to results that cannot be used for comparative analysis of processes and materials.

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DOE OSTI2019

Initial DOE SNF Standardized Canister Storage Configuration Alternatives Study

This study looks at storage configuration alternatives for United States Department of Energy (DOE) Spent Nuclear Fuel (SNF) Standardized Canisters loaded with DOE-owned SNF. The focus of the study is for the DOE SNF Standardized Canisters that would be packaged in the Idaho Nuclear Technology and Engineering Center (INTEC) area at Idaho National Laboratory (INL), but its general findings are useful across the DOE Complex. The study considered storage in existing facilities, modified existing facilities, new facilities, and new modular systems. The existing INTEC facilities considered in the study include the CPP-603 Irradiated Fuel Storage Facility (IFSF), the CPP-749 Underground Fuel Storage Facility (UFSF), the CPP-666 SNF Wet Storage Basin Facility, CPP-1774 Three Mile Island (TMI) Independent Spent Fuel Storage Facility (ISFSF), and the CPP-2707 Cask Pad Facility. The new facilities considered in the study include the Foster Wheeler designed Idaho Spent Fuel (ISF) Independent Spent Fuel Storage Installation (ISFSI) and some conceptual designs. The new modular systems considered in the study include Holtec International’s HI-STAR/HI-STORM, NAC International’s MAGNASTOR/UMS/MPC and Orano TN’s MATRIX/NUHOMS. Based on considerations of various factors, including but not limited to capacity, impact to facility operations, capital costs, expenditure rate, licensing issues, agreements with the State of Idaho, and road ready conditions, new modular storage systems from commercial vendors may have clear advantages over other storage alternatives. This is further supported by recent experience with West Valley Demonstration Project (WVDP) storage project for High Level Waste Glass canisters.

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DOE OSTI2019

Shaker Table Test

This report describes the Shaker Table Test conducted on September 12, 2018, at the Dynamic Certification Laboratories (DCL) in Sparks, Nevada. This report satisfies Milestone M3SF-19SN010202021 Shaker Table Test, Sandia National Laboratories (SNL) Work Package (Parent WBS # 1.08.01.02.02; Work Package #SF-19SN01020202). The Shaker Table Test is related to the Multi-Modal Transportation Test (MMTT) conducted in 2017. During the MMTT, accelerations and strains were measured on the transportation platform, ENsa UNiversal (ENUN) 32P dual-purpose rail cask, cradle, basket, and three surrogate 17x17 pressurized water reactor (PWR) assemblies (one from SNL, one from Spain and one from Korea).

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DOE OSTI2017

Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

A preliminary evaluation of removing spent nuclear fuel (SNF) from 14 shutdown nuclear power plant sites was performed. At these shutdown sites the nuclear power reactors have been permanently shut down and the sites have been decommissioned or are undergoing decommissioning. The shutdown sites were Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, San Onofre, Vermont Yankee, and Fort Calhoun. The evaluation was divided into four components: (1) characterization of the SNF and greater-than-Class C low-level radioactive waste (GTCC waste) inventory, (2) a description of the on-site infrastructure at the shutdown sites, (3) an evaluation of the near-site transportation infrastructure and transportation experience at the shutdown sites, and (4) an evaluation of the actions necessary to prepare for and remove SNF and GTCC waste. The primary sources for the inventory of SNF and GTCC waste were the U.S. Department of Energy (DOE) spent nuclear fuel inventory database, industry publications such as StoreFUEL, and government sources such as the U.S. Nuclear Regulatory Commission. The primary sources for information on the conditions of on-site infrastructure and near-site transportation infrastructure and experience included information collected during site visits, information provided by managers at the shutdown sites, Facility Interface Data Sheets compiled for DOE in 2005, Services Planning Documents prepared for DOE in 1993 and 1994, industry publications such as Radwaste Solutions, and Google Earth. State representatives, State Regional Group representatives, Tribal representatives, and a Federal Railroad Administration representatives have participated in eleven of the shutdown site visits. Every shutdown site was found to have at least one off-site transportation mode option for removing its SNF and GTCC waste; some have multiple options. Experience removing large components during reactor decommissioning provided an important source of information used to identify the transportation mode options for the sites. Especially important in conducting the evaluation were site visits, through which information was obtained that would not have been available otherwise. Extensive photographs taken during the site visits proved to be particularly useful in documenting the current conditions at or near the sites. Additional conclusions from this evaluation include: The 14 shutdown sites use designs from 4 different suppliers involving 11 different (horizontal and vertical) dry storage systems that would require the use of 9 different transportation cask designs to remove the SNF and GTCC waste from the shutdown sites. Although some changes to transportation certificates of compliance will be required, the SNF at the initial 9 shutdown sites (Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion) is in dual purpose dry storage canisters that can be transported, including a small amount of high-burnup fuel. Most sites indicated that 2-3 years of advance time would be required for its preparations before shipments could begin. Some sites could be ready in less time. As additional sites such as Palisades, Pilgrim, Three Mile Island, Oyster Creek, Indian Point, and Diablo Canyon shut down, these sites will be included in updates to the evaluation.

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DOE OSTI2020

PORFLOW Implementation of Vadose Zone Conceptual Model for Naval Reactor Component Disposal Areas in the E-Area Low Level Waste Facility Performance Assessment

PORFLOW models have been developed to implement the proposed conceptual models for the two Naval Reactor Component Disposal Areas (NRCDAs), 643-7E and 643-26E, for the purpose of evaluating dose impacts and producing disposal limits for the E-Area Low-Level Facility (LLWF). New NRCDA conceptual models were proposed by Wohlwend and Butcher (2019) for two areas within/adjacent to the E-Area LLWF used as storage/disposal sites for reactor components from the U.S. Navy since 1987. Naval Reactor (NR) waste is comprised of highly radioactive components consisting of activated corrosion-resistant metal alloy contained within welded thick-walled steel casks, and auxiliary equipment primarily contaminated with lower levels of Activated Corrosion Products (ACPs) residing on the metal surfaces and contained within thinner-walled bolted containers.

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