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Scaglione, John

Publications and source records attributed to Scaglione, John.

12 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Neutron diffraction illustrates residual stress behavior of welded alloys used as radioactive confinement boundary

Corrosion-resistant welded alloys are frequently used as a leak-tight boundary in critical applications that require confinement of hazardous and/or radioactive substances, including an increasing population of spent nuclear fuel (SNF) canisters. The behavior of residual stresses generated as a result of irregular elastic–plastic deformation during processes such as welding is one of today's key issues to a full understanding of the aging mechanisms that may compromise the confinement boundary. Whether such processes and any subsequent weld repairs, not subjected to post-weld heat treatment, would negatively affect the initial material by introducing through-thickness tensile stresses remains an open question. Here we report the first residual stress measurements using neutron diffraction on the welded joints of a SNF canister. We found significant tensile residual stresses in the as welded sample, indicating that initiation and through-thickness growth of cracks may be possible. Following repair, we observed a stress redistribution and introduction of beneficial compressive stresses. We anticipate our results will improve understanding of confinement susceptibility to aging and guide improvements in repair techniques.

36 MATERIALS SCIENCE↗
DOE OSTI2019

A Salt Repository Concept for CSNF in 21-PWR Size Canisters

The most straightforward concept for disposal of large, heavy packages containing commercial spent nuclear fuel (CSNF) in a repository in bedded salt, would be to emplace them directly on the floor in emplacement tunnels. In-tunnel axially aligned horizontal emplacement would minimize excavated volume and avoid drilling of large-diameter emplacement boreholes. A similar concept was proposed in Germany for direct disposal of POLLUX ® canisters. The repository would be constructed at a depth of 500 to 1,000 m for isolation from the surface, and for sufficient overburden stress to ensure creep reconsolidation of repository openings. It could entail modular panels of emplacement tunnels arranged on headings oriented in cardinal directions from a central core, to accommodate the estimated 140,000 MTU total U.S. CSNF inventory. To do so, the overall area of the repository layout would be approximately 20 km 2 . Many layouts are possible, but the approach should be modular, excavation should be deferred for as long as possible to avoid maintenance, and the emplacement areas should share support facilities and shafts. Vertical shafts would be used in accordance with mining practice in sedimentary basins. Large diameter shafts would be needed for ventilation exhaust and waste transport, with smaller shafts for waste salt removal, men & materials, and ventilation intake. The spacing between disposal tunnels as estimated from thermal modeling, seeks to limit the maximum average areal thermal load in the panels to 11 W/m 2 to control long-term heat buildup in the host rock. Peak salt temperature would occur within a few years and would be dominated by each waste package locally, simplifying thermal management. There would be some flexibility to decrease the package spacing or increase the emplacement thermal power limit Backfilling emplaced waste packages immediately with mine-run crushed salt would provide shielding and expedite reconsolidation. This arrangement would isolate adjacent waste packages from one another by the intervening backfill, especially after it reconsolidates and its properties approach those of intact salt. After the repository is fully loaded and the performance confirmation program is complete, activities to permanently close the repository would be initiated. During closure operations all openings in the host salt would be backfilled, then shafts would be sealed, and boreholes plugged. Plans for the Waste Isolation Pilot Plant (WIPP) show how sealing and plugging could be done. A monitoring program could continue for 50 years or longer after repository closure. With an emplacement thermal power limit of 10 kW per waste package, nearly all the CSNF that is projected to be produced by the current fleet of reactors in the U.S. could be emplaced over a period of approximately 50 years starting in calendar 2048. No barriers to implementation in a reasonable timeframe have been identified from this generic analysis. Engineering challenges include: 1) shaft construction; 2) a very-large capacity shaft hoist; 3) overpack design; 4) a transport-emplacement-vehicle (TEV) for transporting waste packages once they are underground and emplacing them remotely; and 5) remotely operated equipment for emplacing backfill. The method of shaft construction would depend on site-specific conditions, and could involve freezing the subsurface. A shaft hoist with payload capacity of 175 MT seems technically and economically feasible based on development work in Germany, and it would be the largest hoist of its kind. The function of disposal overpacks would be to provide reliable containment during repository operations, which could be accomplished using a corrosion allowance material such as a low-carbon steel. Development effort would be needed to determine overpack thickness (e.g., 7 to 20 cm) that can resist corrosion and loading from salt creep, to rovide containment throughout the repository operational period. The transport-emplacement vehicle (TEV) would be similar to previous concepts, particularly one option proposed for a Yucca Mountain repository. It would move over a rough salt floor on independently driven and steered wheels, and carry heavy shielding in addition to a waste package. By analogy to the safety case for the WIPP in New Mexico, human intrusion is likely to be the dominant mode of radionuclide release from the repository. Treatment of human intrusion for a CSNF repository in salt could depend on promulgation of site-specific changes in the regulations. Radionuclide release and migration would be quite limited for undisturbed conditions. There may be opportunities for improved understanding of salt performance with waste heating, based on future in situ testing in an underground salt research laboratory. This report also discusses a developing area of salt rock mechanics that involves low-stress, low strain-rate creep that might cause large, heavy waste packages to slowly sink. Site-specific sampling, testing, and modeling would be used to determine if the mechanism is important enough to merit consideration in design, or inclusion in performance assessment. Part of engineering design and postclosure safety assessment for a CSNF repository in salt would be to implement a methodology to show that the probability of a criticality event in the repository, when waste packages eventually breach and are flooded, is less than the probability screening threshold for performance assessment. In the methodology, a criticality analysis would be performed for waste packages in the repository, incorporating measures that could be introduced as needed to limit reactivity, for example using fuel selection and loading rules, and crediting the absorption of thermal neutrons by natural chlorine in the environment. A similar analysis has been underway for CSNF stored in dual-purpose canisters. Ideally the strategy would be developed prior to actually loading SNF assemblies into canisters used in waste packages for disposal.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Visualization of the High-Burnup Spent Fuel Rod Phase 1 Test Plan: Technical Memo - Spent Fuel and Waste Disposition (Technical Memo)

Twenty-five high-burnup fuel rods were extracted from seven different fuel assemblies used for power production at the North Anna nuclear power plant and shipped to Oak Ridge National Laboratory (ORNL) in 2016 for detailed non-destructive examination (NDE) and destructive examination (DE). The spent fuel rods were from 17×17 lattices and consist of four cladding types—Zirlo ® , M5 ® , Zircaloy-4, and low tin Zircaloy-4 (Zirc-4). These spent fuel rods are being tested to provide: (a) baseline characterization and mechanical property data that can be used as a comparison to fuel that was loaded into a modified TN-32B cask in November 2017, as part of the high-burnup confirmatory data project and (b) data applicable to high-burnup fuel rods (>45 GWd/MTU) currently stored and to be stored in the dry-cask fleet. The TN-32B cask is referred to as the “Demo” cask and is currently expected to be transported to a separate location and the internal contents inspected in approximately ten years. ORNL has completed the NDE of the twenty-five fuel rods. The purpose of this technical memorandum is to present a simplified summary of the first phase of destructive examinations and test conditions that will be used for communicating with various stakeholders. The destructive examinations will leverage the expertise and capabilities from multiple national laboratories for performing independent measurements of relevant data. Close coordination is required to ensure that all examinations follow well documented procedures and are performed so that measured data and characteristics can be readily compared. Pacific Northwest National Laboratory (PNNL) has published a detailed overview of the test program. ORNL and PNNL developed detailed draft test plans for testing to be performed at their facilities. ORNL and PNNL are in the process of refining these test plans to apply specifically to the testing described in this memorandum. Argonne National Laboratory (ANL) contributed to the ORNL test plan by describing tests to be conducted at ANL. Testing will be based on continuous learning. If a test produces results that are inconsistent with expectations or current trends, further testing will be paused until a path forward is established to understand the results and to identify follow-on testing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

EPRI/DOE High-Burnup Fuel Sister Rod Test Plan Simplification and Visualization

The EPRI/DOE High-Burnup Confirmatory Data Project (herein called the “Demo”) is a multi-year, multi-entity test with the purpose of providing quantitative and qualitative data to show if high-burnup fuel mechanical properties change in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of common cladding alloys from the North Anna Nuclear Power Plant, loading them in an NRC-licensed TN-32B cask, drying them according to standard plant procedures, and then storing them on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened and the mechanical properties of the rods will be tested and analyzed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Summary of Investigations on Technical Feasibility of Direct Disposal of Dual-Purpose Canisters

This report is a revision of the Fiscal Year (FY)-15 summary report on the investigations on technical feasibility of direct disposal of dual-purpose canisters (DPCs) in a geological repository. Notable progress has been made in the postclosure criticality area supporting direct disposal of DPCs in FY-16 and FY-17. As such, this revision updates the postclosure criticality related studies to capture the advances made over the last two FYs.

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

EPRI/DOE High Burnup Fuel Sister Pin Test Plan Simplification and Visualization

The EPRI/DOE High Burnup Confirmatory Data Project (herein called the "Demo") is a multi-year, multi-entity confirmation demonstration test with the purpose of providing quantitative and qualitative data to show how high-burnup fuel ages in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of four common cladding alloys from the North Anna Nuclear Power Plant, drying them according to standard plant procedures, and then storing them in an NRC-licensed TN-3 2B cask on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened and the rods will be examined for signs of aging. Twenty-five rods from assemblies of similar claddings, in-reactor placement, and burnup histories (herein called "sister rods") have been shipped from the North Anna Nuclear Power Plant and are currently being nondestructively tested at Oak Ridge National Laboratory. After the non-destructive testing has been completed for each of the twenty-five rods, destructive analysis will be performed at ORNL, PNNL, and ANL to obtain mechanical data. Opinions gathered from the expert interviews, ORNL and PNNL Sister Rod Test Plans, and numerous meetings has resulted in the Simplified Test Plan described in this document. Some of the opinions and discussions leading to the simplified test plan are included here. Detailed descriptions and background are in the ORNL and PNNL plans in the appendices . After the testing described in this simplified test plan h as been completed , the community will review all the collected data and determine if additional testing is needed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Material Protection, Accounting, and Control Technologies (MPACT): Modeling and Simulation Roadmap

The development of sustainable advanced nuclear fuel cycles is a long-term goal of the Office of Nuclear Energy’s (DOE-NE) Fuel Cycle Technologies program. The Material Protection, Accounting, and Control Technologies (MPACT) campaign is supporting research and development (R&D) of advanced instrumentation, analysis tools, and integration methodologies to meet this goal. This advanced R&D is intended to facilitate safeguards and security by design of fuel cycle facilities. The lab-scale demonstration of a virtual facility, distributed test bed, that connects the individual tools being developed at National Laboratories and university research establishments, is a key program milestone for 2020. These tools will consist of instrumentation and devices as well as computer software for modeling. To aid in framing its long-term goal, during FY16, a modeling and simulation roadmap is being developed for three major areas of investigation: (1) radiation transport and sensors, (2) process and chemical models, and (3) shock physics and assessments. For each area, current modeling approaches are described, and gaps and needs are identified.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Simulation of Proposed Heat Treatment Methods for the Sister Rods

The High Burnup Spent Fuel Data Project (the 'project'), sponsored by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), is focused on experimentally evaluating the effects of long-term dry storage and transportation on high burnup (HBU) pressurized water reactor (PWR) fuel (>45 gigawatt days per metric ton uranium). The goals of the project are to provide confirmatory data for analytical model validation and potential improvement; to provide input to future dry storage cask design; to support license renewals and new licenses for Independent Spent Fuel Storage Installation facilities; and to support transportation licensing for high burnup used fuel. In support of these project goals, an instrumented TN-32B bolted-lid cask (the Research Project Cask or RPC) at the North Anna reactor site will be loaded with intact, HBU PWR fuel assemblies (Project Assemblies) having four different kinds of cladding: standard Zircaloy-4, low-tin Zircaloy-4, ZIRLO{sup R}, and M5{sup R}. In approximately ten years, the instrumented cask will be opened for inspection and several fuel rods from the Project Assemblies will be characterized. The project has also removed 25 'sister rods' from assemblies at North Anna: 9 directly from the project assemblies and 16 from similarly constructed and operated HBU assemblies. The 25 sister rods are all high burnup and cover the range of domestically-used cladding alloys. The 25 sister rods were shipped to Oak Ridge National Laboratory (ORNL) in early 2016 for detailed non-destructive examinations (NDE) followed by destructive examination (DE). Examination of the sister rods will provide a baseline for comparison with the cask rods and will allow for separate effects testing. Dry storage canister loading is completed under water in a designated area of the plant's spent fuel pool. In preparation for dry storage, the fuel assemblies and canister cavity must be drained and dried. Typically, the highest temperatures experienced by the fuel during dry storage are encountered either during drying or transfer of the canister to the storage pad; therefore, as a part of the sister rod testing these temperatures will be simulated in the lab to produce samples that are representative of the peak cladding temperature the fuel experiences either during drying, transfer, or storage. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

U.S. Department of Energy, Office of Nuclear Energy, Nuclear Fuels Storage and Transportation Planning Project Modeling Tools

The U.S. Department Energy, Office of Nuclear Energy's (DOE-NE) Nuclear Fuels Storage and Transportation (NFST) is developing foundational capabilities and tools to support and inform the future deployment of consolidated interim storage and large-scale transportation of SNF. These tools include Used Nuclear Fuel Storage, Transportation, and Disposal Analysis Resource and Data System (UNF-Standards), Stakeholder Tool for Assessing Radioactive Transportation (START), Transportation and Storage Logistics (TSL), Next- Generation System Analysis Model (NGSAM), Execution Strategy Analysis (ESA), Multi-Objective Evaluation Framework (MOEF), and the Centralized Used Fuel Resource for Information Exchange (CURIE). (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Used fuel extended storage security and safeguards by design roadmap

In the United States, spent nuclear fuel (SNF) is safely and securely stored in spent fuel pools and dry storage casks. The available capacity in spent fuel pools across the nuclear fleet has nearly reached a steady state value. The excess SNF continues to be loaded in dry storage casks. Fuel is expected to remain in dry storage for periods beyond the initial dry cask certification period of 20 years. Recent licensing renewals have approved an additional 40 years. This report identifies the current requirements and evaluation techniques associated with the safeguards and security of SNF dry cask storage. A set of knowledge gaps is identified in the current approaches. Finally, this roadmap identifies known knowledge gaps and provides a research path to deliver the tools and models needed to close the gaps and allow the optimization of the security and safeguards approaches for an interim spent fuel facility over the lifetime of the storage site.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Groundwork for Universal Canister System Development

The mission of the United States Department of Energy’s Office of Environmental Management is to complete the safe cleanup of the environmental legacy brought about from five decades of nuclear weapons development and government-sponsored nuclear energy research. Some of the wastes that that must be managed have been identified as good candidates for disposal in a deep borehole in crystalline rock (SNL 2014a). In particular, wastes that can be disposed of in a small package are good candidates for this disposal concept. A canister-based system that can be used for handling these wastes during the disposition process (i.e., storage, transfers, transportation, and disposal) could facilitate the eventual disposal of these wastes. This report provides information for a program plan for developing specifications regarding a canister-based system that facilitates small waste form packaging and disposal and that is integrated with the overall efforts of the DOE’s Office of Nuclear Energy Used Fuel Disposition Campaign's Deep Borehole Field Test. Wastes to be considered as candidates for the universal canister system include capsules containing cesium and strontium currently stored in pools at the Hanford Site, cesium to be processed using elutable or nonelutable resins at the Hanford Site, and calcine waste from Idaho National Laboratory. The initial emphasis will be on disposal of the cesium and strontium capsules in a deep borehole that has been drilled into crystalline rock. Specifications for a universal canister system are derived from operational, performance, and regulatory requirements for storage, transfers, transportation, and disposal of radioactive waste. Agreements between the Department of Energy and the States of Washington and Idaho, as well as the Deep Borehole Field Test plan provide schedule requirements for development of the universal canister system. Future work includes collaboration with the Hanford Site to move the cesium and strontium capsules into dry storage, collaboration with the Deep Borehole Field Test to develop surface handling and emplacement techniques and to develop the waste package design requirements, developing universal canister system design options and concepts of operations, and developing system analysis tools. Areas in which further research and development are needed include material properties and structural integrity, in-package sorbents and fillers, waste form tolerance to heat and postweld stress relief, waste package impact limiters, sensors, cesium mobility under downhole conditions, and the impact of high pressure and high temperature environment on seals design.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗