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At least 649 records · Page 36

Consequence Analysis for Used Fuel Extended Storage

Early identification and evaluation of security issues related to the extended storage of used nuclear fuel is critical. A breach in a dry fuel storage container has the possibility of external gas from the atmosphere interacting with the used fuel rods at high temperatures, resulting in rapid oxidation and possibly the ignition of a zirconium fire. In support of this idea, the current work aims to develop a computational model of heat transfer and fluid flow in and through a breached dry fuel storage cask to determine if the resulting flow conditions are likely to result in a fire.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Draft Geologic Disposal Requirements Basis for STAD Specification

This document provides the basis for requirements in the current version of Performance Specification for Standardized Transportation, Aging, and Disposal Canister Systems, (FCRD-NFST-2014-0000579) that are driven by storage and geologic disposal considerations. Performance requirements for the Standardized Transportation, Aging, and Disposal (STAD) canister are given in Section 3.1 of that report. Here, the requirements are reviewed and the rationale for each provided. Note that, while FCRD-NFST-2014-0000579 provides performance specifications for other components of the STAD storage system (e.g. storage overpack, transfer and transportation casks, and others), these have no impact on the canister performance during disposal, and are not discussed here.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Remote Whispering Applying Time Reversal

The purpose of this project was to explore the use of time reversal technologies as a means for communication to a targeted individual or location. The idea is to have the privacy of whispering in one’s ear, but to do this remotely from loudspeakers not located near the target. Applications of this work include communicating with hostages and survivors in rescue operations, communicating imaging and operational conditions in deep drilling operations, monitoring storage of spent nuclear fuel in storage casks without wires, or clandestine activities requiring signaling between specific points. This technology provides a solution in any application where wires and radio communications are not possible or not desired. It also may be configured to self calibrate on a regular basis to adjust for changing conditions. These communications allow two people to converse with one another in real time, converse in an inaudible frequency range or medium (i.e. using ultrasonic frequencies and/or sending vibrations through a structure), or send information for a system to interpret (even allowing remote control of a system using sound). The time reversal process allows one to focus energy to a specific location in space and to send a clean transmission of a selected signal only to that location. In order for the time reversal process to work, a calibration signal must be obtained. This signal may be obtained experimentally using an impulsive sound, a known chirp signal, or other known signals. It may also be determined from a numerical model of a known environment in which the focusing is desired or from passive listening over time to ambient noise.

42 ENGINEERING↗

Performance Specification for Standardized Transportation, Aging, and Disposal Canister Systems (Rev.2)

This performance specification document addresses use of a standardized transportation, aging, and disposal (STAD) canister system in the spent nuclear fuel (SNF) waste management system. This document was developed to support STAD canister system studies and potential research, development, and demonstration activities. Requirements in this specification may evolve with time based on results from analyses, experiments, design studies, system evaluations and demonstrations, as well as other factors. A STAD canister system consists of a canister, together with a storage or aging overpack/module/vault, a shielded transfer cask (STC), a site transporter, a transportation overpack, and a transportation skid. Three sizes of circular cross section STAD canister variants are specified in this document. The small STAD canister capacity is four pressurized water reactor (PWR) or nine boiling water reactor (BWR) SNF assemblies. The medium STAD canister capacity is 12 PWR or 32 BWR SNF assemblies. The large STAD canister capacity is 21 PWR or 44 BWR SNF assemblies. Using standardized canisters for SNF facilitates the interface between storage at an independent spent fuel storage installation (ISFSI) or a DOEowned interim storage facility (ISF), transportation between facilities, and disposal at a repository. Multiple STAD canisters may be stored in a single storage overpack and/or a single transportation overpack. The US Nuclear Regulatory Commission (NRC) requirements currently used for storage and transportation are adopted for the STAD canister system. Approved approaches (guidance) in NRC storage and transportation documents are also adopted as requirements for the STAD canister system to facilitate obtaining NRC Certificates of Compliance (CoCs) for STAD canister systems. Performance specifications in this document are also informed by STAD canister disposability in various geologic media and consideration of ISF (and geologic repository) handling facility design, operations, and licensing requirements.

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Rationale for the Performance Specification for Standardized Transportation, Aging, and Disposal Canister Systems (Rev.1)

This document provides the rationale for the specifications for the standardized transportation, aging, and disposal (STAD) canister systems provided in Performance Specification for Standardized Transportation, Aging, and Disposal Canister Systems, FCRD-NFST-2014-000579, Rev. 2, referred to as STAD Spec herein. The STAD Spec was developed to support STAD canister system studies and potential research, development, and demonstration activities. Requirements in the specification may evolve with time based on results from analyses, experiments, design studies, system evaluations and demonstrations, as well as other factors. The STAD canister system consists of a canister, together with a storage or aging overpack/module/vault, a shielded transfer cask (STC), a site transporter, a transportation overpack, and a transportation skid. Three sizes of circular cross section STAD canister variants are specified. The small STAD canister capacity is four pressurized water reactor (PWR) or nine boiling water reactor (BWR) spent nuclear fuel (SNF) assemblies. The medium STAD canister capacity is 12 PWR or 32 BWR SNF assemblies. The large STAD canister capacity is 21 PWR or 44 BWR SNF assemblies. There are 60 specifications in the STAD Spec that cover the STAD canister general design attributes, canister safety functional requirements, operational considerations, materials, storage and aging system requirements, and transportation system requirements. The 60 specifications are derived from regulatory requirements, regulatory guidance, STAD canister disposability considerations, and interim storage facility (ISF) and geologic repository handling facility considerations.

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HFIR Dissolution Nuclear Criticality Analysis Information

The purpose of this analysis is to demonstrate the safety of High Flux Isotope Reactor (HFIR) dissolution from a criticality safety perspective and to establish any Criticality Safety Limits (CSLs) that may be necessary. HFIR fuel consists of two annular elements that are designed to be nested together with a beryllium reflector in the reactor configuration. HFIR cores are currently stored in L-Area as separate inner elements and outer elements, each on a unique aluminum carrier. HFIR elements will be shipped to H-Canyon in the 70-ton Cask Railcar and loaded into the dissolver with the HFIR insert installed. The HFIR insert consists of a small well for the inner fuel element and a large well for the outer fuel element.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Preliminary Results of Ancillary Safety Analyses Supporting TREAT LEU Conversion Activities

The Transient Reactor Test Facility (TREAT), located at Idaho National Laboratory (INL), is a test facility designed to evaluate the performance of reactor fuels and materials under transient accident conditions. The facility, an air-cooled, graphite-moderated reactor designed to utilize fuel containing high-enriched uranium (HEU), has been in non-operational standby status since 1994. Currently, in support of the missions of the Department of Energy (DOE) National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Reactor Conversion Program, a new core design is being developed for TREAT that will utilize low-enriched uranium (LEU). The primary objective of this conversion effort is to design an LEU core that is capable of meeting the performance characteristics of the existing HEU core. Minimal, if any, changes are anticipated for the supporting systems (e.g. reactor trip system, filtration/cooling system, etc.); therefore, the LEU core must also be able to function with the existing supporting systems, and must also satisfy acceptable safety limits. In support of the LEU conversion effort, a range of ancillary safety analyses are required to evaluate the LEU core operation relative to that of the existing facility. These analyses cover neutronics, shielding, and thermal hydraulic topics that have been identified as having the potential to have reduced safety margins due to conversion to LEU fuel, or are required to support the required safety analyses documentation. The majority of these ancillary tasks have been identified in [1] and [2]. The purpose of this report is to document the ancillary safety analyses that have been performed at Argonne National Laboratory during the early stages of the LEU design effort, and to describe ongoing and anticipated analyses. For all analyses presented in this report, methodologies are utilized that are consistent with, or improved from, those used in analyses for the HEU Final Safety Analysis Report (FSAR) [3]. Depending on the availability of historical data derived from HEU TREAT operation, results calculated for the LEU core are compared to measurements obtained from HEU TREAT operation. While all analyses in this report are largely considered complete and have been reviewed for technical content, it is important to note that all topics will be revisited once the LEU design approaches its final stages of maturity. For most safety significant issues, it is expected that the analyses presented here will be bounding, but additional calculations will be performed as necessary to support safety analyses and safety documentation. It should also be noted that these analyses were completed as the LEU design evolved, and therefore utilized different LEU reference designs. Preliminary shielding, neutronic, and thermal hydraulic analyses have been completed and have generally demonstrated that the various LEU core designs will satisfy existing safety limits and standards also satisfied by the existing HEU core. These analyses include the assessment of the dose rate in the hodoscope room, near a loaded fuel transfer cask, above the fuel storage area, and near the HEPA filters. The potential change in the concentration of tramp uranium and change in neutron flux reaching instrumentation has also been assessed. Safety-significant thermal hydraulic items addressed in this report include thermally-induced mechanical distortion of the grid plate, and heating in the radial reflector.

Brunett, A. J.↗

Report on UQ and PCMM Analysis of Vacuum Drying for UFD S&T Gaps

This report discusses two phenomena that could affect the safety, licensing, transportation, storage, and disposition of the spent fuel storage casks and their contents (radial hydriding during drying and water retention after drying) associated with the drying of canisters for dry spent fuel storage. The report discusses modeling frameworks and evaluations that are, or have been, developed as a means to better understand these phenomena. Where applicable, the report also discusses data needs and procedures for monitoring or evaluating the condition of storage containers during and after drying. A recommendation for the manufacturing of a fully passivated fuel rod, resistant to oxidation and hydriding is outlined.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Physics Flash August 2016

Physics Flash is the newsletter for the Physics Division at Los Alamos National Laboratory. This newsletter is for August 2016. The following topics are covered: "Accomplishments in the Trident Laser Facility", "David Meyerhofer elected as chair-elect APS Nominating Committee", "HAWC searches for gamma rays from dark matter", "Proton Radiography Facility commissions electromagnetic magnifier", and "Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks."

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

Report on Status of Shipment of High Fluence Austenitic Steel Samples for Characterization and Stress Corrosion Crack Testing

The goal of the Mechanisms of Irradiation Assisted Stress Corrosion Cracking (IASCC) task in the LWRS Program is to conduct experimental research into understanding how multiple variables influence the crack initiation and crack growth in materials subjected to stress under corrosive conditions. This includes understanding the influences of alloy composition, radiation condition, water chemistry and metallurgical starting condition (i.e., previous cold work or heat treatments and the resulting microstructure) has on the behavior of materials. Testing involves crack initiation and growth testing on irradiated specimens of single-variable alloys in simulated Light Water Reactor (LWR) environments, tensile testing, hardness testing, microstructural and microchemical analysis, and detailed efforts to characterize localized deformation. Combined, these single-variable experiments will provide mechanistic understanding that can be used to identify key operational variables to mitigate or control IASCC, optimize inspection and maintenance schedules to the most susceptible materials/locations, and, in the long-term, design IASCC-resistant materials. In support of this research, efforts are currently underway to arrange shipment of “free” high fluence austenitic alloys available through Électricité de France (EDF) for post irradiation testing at the Oak Ridge National Laboratory (ORNL) and IASCC testing at the University of Michigan. These high fluence materials range in damage values from 45 to 125 displacements per atom (dpa). The samples identified for transport to the United States, which include nine, no-cost, 304, 308 and 316 tensile bars, were relocated from the Research Institute of Atomic Reactors (RIAR) in Dimitrovgrad, Ulyanovsk Oblast, Russia, and received at the Halden Reactor in Halden, Norway, on August 23, 2016. ORNL has been notified that a significant amount of work is required to prepare the samples for further shipment to Oak Ridge, Tennessee. The preliminary work for sample shipment between Halden and Oak Ridge includes fabrication of an inner cask sample container, decontamination and preparation of a Type A container, preparation of new activity calculations, all necessary paperwork, and handling. ORNL will continue to work to track progress of sample preparation and shipment status, and to work toward an agreement that covers material shipping costs between the Halden Reactor and the Oak Ridge National Laboratory.

36 MATERIALS SCIENCE↗

Concrete Materials with Ultra-High Damage Resistance and Self- Sensing Capacity for Extended Nuclear Fuel Storage Systems

The objective of this project is to develop a new class of multifunctional concrete materials (MSCs) for extended spent nuclear fuel (SNF) storage systems, which combine ultra-high damage resistance through strain-hardening behavior with distributed multi-dimensional damage self-sensing capacity. The beauty of multifunctional concrete materials is two-fold: First, it serves as a major material component for the SNF pool, dry cask shielding and foundation pad with greatly improved resistance to cracking, reinforcement corrosion, and other common deterioration mechanisms under service conditions, and prevention from fracture failure under extreme events (e.g. impact, earthquake). This will be achieved by designing multiple levels of protection mechanisms into the material (i.e., ultrahigh ductility that provides thousands of times greater fracture energy than concrete and normal fiber reinforced concrete; intrinsic cracking control, electrochemical properties modification, reduced chemical and radionuclide transport properties, and crack-healing properties). Second, it offers capacity for distributed and direct sensing of cracking, strain, and corrosion wherever the material is located. This will be achieved by establishing the changes in electrical properties due to mechanical and electrochemical stimulus. The project will combine nano-, micro- and composite technologies, computational mechanics, durability characterization, and structural health monitoring methods, to realize new MSCs for very long-term (greater than 120 years) SNF storage systems.

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Concrete Materials with Ultra-High Damage Resistance and Self- Sensing Capacity for Extended Nuclear Fuel Storage Systems

The objective of this project is to develop a new class of multifunctional concrete materials (MSCs) for extended spent nuclear fuel (SNF) storage systems, which combine ultra-high damage resistance through strain-hardening behavior with distributed multi-dimensional damage self-sensing capacity. The beauty of multifunctional concrete materials is two-fold: First, it serves as a major material component for the SNF pool, dry cask shielding and foundation pad with greatly improved resistance to cracking, reinforcement corrosion, and other common deterioration mechanisms under service conditions, and prevention from fracture failure under extreme events (e.g. impact, earthquake). This will be achieved by designing multiple levels of protection mechanisms into the material (i.e., ultrahigh ductility that provides thousands of times greater fracture energy than concrete and normal fiber reinforced concrete; intrinsic cracking control, electrochemical properties modification, reduced chemical and radionuclide transport properties, and crack-healing properties). Second, it offers capacity for distributed and direct sensing of cracking, strain, and corrosion wherever the material is located. This will be achieved by establishing the changes in electrical properties due to mechanical and electrochemical stimulus. The project will combine nano-, micro- and composite technologies, computational mechanics, durability characterization, and structural health monitoring methods, to realize new MSCs for very long-term (greater than 120 years) SNF storage systems.

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Test report dot 7A type a liquid packaging

This test report documents the performance of Savannah River National Laboratory’s (SRNL’s) U.S. Department of Transportation (DOT) Specification 7A; General Packaging, Type A shielded liquid shipping packaging and compliance with the regulatory requirements of Title 49 of the Code of Federal Regulations (CFR). The primary use of this packaging design is for the transport of radioactive liquids of up to 1.3 liters in an unshielded configuration and up to 113 mL of radioactive liquids in a shielded configuration, with no more than an A2 quantity in either configuration, over public highways and/or commercial aircraft. The contents are liquid radioactive materials sufficiently shielded and within the activity limits specified in173.435 or 173.433 for A2 (normal form) materials, as well as within the analyzed thermal heat limits. Any contents must be compatibly packaged and must be compatible with the packaging. The basic packaging design is based on the U.S. Department of Energy’s (DOE’s) Model 9979 Type A fissile shipping packaging designed and tested by SRNL. The shielded liquid configuration consists of the outer and inner drums of the 9979 package with additional low density polyethylene (LDPE) dunnage nesting a tungsten shielded cask assembly (WSCA) within the 30-gallon inner drum. The packaging model for the DOT Specification 7A, Type A liquids packaging is HVYTAL.

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The Impact of Operating Parameters and Correlated Parameters for Extended BWR Burnup Credit

Applicants for certificates of compliance for spent nuclear fuel (SNF) transportation and dry storage systems perform analyses to demonstrate that these systems are adequately subcritical per the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Parts 71 and 72. For pressurized water reactor (PWR) SNF, these analyses may credit the reduction in assembly reactivity caused by depletion of fissile nuclides and buildup of neutron-absorbing nuclides during power operation. This credit for reactivity reduction during depletion is commonly referred to as burnup credit (BUC). US Nuclear Regulatory Commission (NRC) staff review BUC analyses according to the guidance in the Division of Spent Fuel Storage and Transportation Interim Staff Guidance (ISG) 8, Revision 3, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

A preliminary evaluation of removing spent nuclear fuel (SNF) from 13 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, and Vermont Yankee. The evaluation was divided into four components: Characterization of the SNF and greater-than-Class C low-level radioactive waste (GTCC waste) inventory A description of the on-site infrastructure at the shutdown sites An evaluation of the near-site transportation infrastructure and transportation experience at the shutdown sites 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 staff, State Regional Group representatives, a Tribal representative, and a Federal Railroad Administration representative have participated in nine 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 13 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 Fort Calhoun, Clinton, Quad Cities, Pilgrim, Oyster Creek, and Diablo Canyon shut down, these sites will be included in updates to the evaluation.

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Assessment of Impact of Monoenergetic Photon Sources on Prioritized Nonproliferation Applications: Simulation Study Report

Near-monoenergetic photon sources (MPSs) have the potential to improve sensitivity at greatly reduced dose in existing applications and enable new capabilities in other applications. MPS advantages include the ability to select energy, energy spread, flux, and pulse structures to deliver only the photons needed for the application, while suppressing extraneous dose and background. Some MPSs also offer narrow divergence photon beams which can target dose and/or mitigate scattering contributions to image contrast degradation. Current broad-band, bremsstrahlung photon sources (e.g., linacs and betatrons) deliver unnecessary dose that in some cases also interferes with the signature to be detected and/or restricts operations, and must be collimated (reducing flux) to generate narrow divergence beams. While MPSs can in principle resolve these issues, they are technically challenging to produce. Candidate MPS technologies for nonproliferation applications are now being developed, each of which have different properties (e.g. broad divergence vs. narrow). Within each technology, source parameters trade off against one another (e.g. flux vs. energy spread), representing a large operation space. To guide development, requirements for each application of interest must be defined and simulations conducted to define MPS parameters that deliver benefit relative to current systems. The present project conducted a broad assessment of potential nonproliferation applications where MPSs may provide new capabilities or significant performance enhancement (reported separately), which led to prioritization of several applications for detailed analysis. The applications prioritized were: cargo screening and interdiction of Special Nuclear Materials (SNM), detection of hidden SNM, treaty/dismantlement verification, and spent fuel dry storage cask content verification. High resolution imaging for stockpile stewardship was considered as a sub-area of the treaty topic, as it is also of interest for future treaty use. This report presents higher-fidelity calculations and modeling results to quantitatively evaluate the prioritized applications, and to derive the key MPS properties that drive application benefit. Simulations focused on the conventional signatures of radiography, photofission, and NRF to enable comparison to present methods and evaluation of benefit.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Concepts for Small-Scale Testing of Used Nuclear Fuel

This report documents a concept for a small-scale test involving between one and three Boiling Water Rector (BWR) high burnup (HBU) fuel assemblies. This test would be similar to the DOE funded High Burn-Up (HBU) Confirmatory Data Project to confirm the behavior of used high burn-up fuel under prototypic conditions, only on a smaller scale. The test concept proposed would collect data from fuel stored under prototypic dry storage conditions to mimic, as closely as possible, the conditions HBU UNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to an Independent Spent Fuel Storage Installation (ISFSI) for multi-year storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

UFD Proof-of-Concept Demonstration

The BISON Spent Nuclear Fuel modeling capability has been demonstrated through simulations of the complete nuclear fuel lifecycle including in-reactor irradiation, spent fuel pool cooling, vacuum drying and finally dry cask storage. The importance of hydrogen in the cladding and the formation of hydrides at various stages of the fuel lifecycle have been a focus of the initial development work. In addition, the integration in BISON of a hydride damage model has improved the prediction of the cladding mechanical behavior. The work described in this report was performed under funding from the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. This report is issued in satisfaction of the level 2 milestone M2MS-17IN0201012 on UFD Proof-of-Concept Demonstration for the INL Engineering-Scale Fuel Performance (BISON) Workpackage.

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