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At least 559 records · Page 31

Can Collaboration Succeed in Siting a Spent Nuclear Fuel Facility in the United States?—A Challenge in Political Sustainability

We examine the U.S. Department of Energy (DOE)’s collaborative process to locate, build, and operate one or more federal consolidated interim storage facilities (FCISFs) for commercial U.S. spent nuclear fuel—instead of continuing to store the material at over 70 nuclear reactor sites. Technocratic siting of nuclear facilities in the U.S., most of which did not involve meaningful public participation, was not successful. We consider increasing pressure to find at least one FCISF site, as well as the critical role of trust in engaging communities and reaching agreement—leading some observers to assert that DOE is in the “trust building business”, not the siting business. We present case studies with the following: (1) illustrating community engagement that led to a more satisfactory outcome than had been anticipated (Fernald); (2) a planned voluntary process that failed to produce an operating CISF (Office of the Nuclear Waste Negotiator); and (3) a site that demonstrates the ongoing need for negotiations to keep a site open and operational (Waste Isolation Pilot Plant). The essay concludes with the observation that a collaboration-based siting effort can succeed in the U.S., but that five main challenges—related to trust and requiring patience—will need to be addressed.

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Development of a Reliable Fuel Depletion Methodology for the HTR-10 Spent Fuel Analysis

A technical working group formed in 2007 between NNSA and CAEA to develop a reliable fuel depletion method for HTR-10 based on MCNPX and to analyze the isotopic inventory and radiation source terms of the HTR-10 spent fuel. Conclusions of this presentation are: (1) Established a fuel depletion methodology and demonstrated its safeguards application; (2) Proliferation resistant at high discharge burnup ({approx}80 GWD/MtHM) - Unfavorable isotopics, high number of pebbles needed, harder to reprocess pebbles; (3) SF should remain under safeguards comparable to that of LWR; and (4) Diversion scenarios not considered, but can be performed.

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Determination of total Pu content in a Spent Fuel Assembly by Measuring Passive Neutron Count rate and Multiplication with the Differential Die-Away Instrument

A key objective of the Next Generation Safeguards Initiative (NGSI) is to evaluate and develop non-destructive assay (NDA) techniques to determine the elemental plutonium content in a commercial-grade nuclear spent fuel assembly (SFA) [1]. Within this framework, we investigate by simulation a novel analytical approach based on combined information from passive measurement of the total neutron count rate of a SFA and its multiplication determined by the active interrogation using an instrument based on a Differential Die-Away technique (DDA). We use detailed MCNPX simulations across an extensive set of SFA characteristics to establish the approach and demonstrate its robustness. It is predicted that Pu content can be determined by the proposed method to a few %.

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Monitoring, Controlling and Safeguarding Radiochemical Streams at Spent Fuel Reprocessing Facilities with Optical and Gamma-Ray Spectroscopic Methods

The International Atomic Energy Agency (IAEA) has established international safeguards standards for fissionable material at spent fuel reprocessing plants to ensure that significant quantities of weapons-useable nuclear material are not diverted from these facilities. For large throughput nuclear facilities, it is difficult to satisfy the IAEA safeguards accountancy goal for detection of abrupt diversion. Currently, methods to verify material control and accountancy (MC&A) at these facilities require time-consuming and resourceintensive destructive assay (DA). Leveraging new on-line non-destructive assay (NDA) process monitoring techniques in conjunction with the traditional and highly precise DA methods may provide an additional measure to nuclear material accountancy which would potentially result in a more timely, cost-effective and resource efficient means for safeguards verification at such facilities. By monitoring process control measurements (e.g. flowrates, temperatures, or concentrations of reagents, products or wastes), abnormal plant operations can be detected. Pacific Northwest National Laboratory (PNNL) is developing on-line NDA process monitoring technologies based upon gamma-ray and optical spectroscopic measurements to potentially reduce the time and resource burden associated with current techniques. The Multi-Isotope Process (MIP) Monitor uses gamma spectroscopy and multivariate analysis to identify offnormal conditions in process streams. The spectroscopic monitor continuously measures chemical compositions of the process streams including actinide metal ions (U, Pu, Np), selected fission products, and major stable flowsheet reagents using UV-Vis, Near IR and Raman spectroscopy. Multi-variate analysis is also applied to the optical measurements in order to quantify concentrations of analytes of interest within a complex array of radiochemical streams. This paper will provide an overview of these methods and reports on-going efforts to develop and demonstrate the technologies.

controlling↗

Technical Cross-Cutting Issues for the Next Generation Safeguards Initiative's Spent Fuel Nondestructive Assay Project

Ever since there has been spent fuel (SF), researchers have made nondestructive assay (NDA) measurements of that fuel to learn about its content. In general these measurements have focused on the simplest signatures (passive photon and total neutron emission) and the analysis has often focused on diversion detection and on determining properties such as burnup (BU) and cooling time (CT). Because of shortcomings in current analysis methods, inspectorates and policy makers are interested in improving the state-of-the-art in SF NDA. For this reason the U.S. Department of Energy, through the Next Generation Safeguards Initiative (NGSI), targeted the determination of elemental Pu mass in SF as a technical goal. As part of this research effort, 14 nondestructive assay techniques were studied . This wide range of techniques was selected to allow flexibility for the various needs of the safeguards inspectorates and to prepare for the likely integration of one or more techniques having complementary features. In the course of researching this broad range of NDA techniques, several cross-cutting issues were. This paper will describe some common issues and insights. In particular we will describe the following: (1) the role of neutron absorbers with emphasis on how these absorbers vary in SF as a function of initial enrichment, BU and CT; (2) the need to partition the measured signal among different isotopic sources; and (3) the importance of the “first generation” concept which indicates the spatial location from which the signal originates as well as the isotopic origins.

Next Generation Safeguards Initiative↗

Potential Impact of Interfacial Bonding Efficiency on High-Burnup Spent Nuclear Fuel Vibration Integrity during Normal Transportation

Finite element analysis (FEA) was used to investigate the impacts of interfacial bonding efficiency at pellet pellet and pellet clad interfaces on spent nuclear fuel (SNF) vibration integrity. The FEA simulation results were also validated and benchmarked with reverse bending fatigue test results on surrogate rods consisting of stainless steel (SS) tubes with alumina-pellet inserts. Bending moments (M) are applied to the FEA models to evaluate the system responses of the surrogate rods. From the induced curvature, , the flexural rigidity EI can be estimated as EI=M/ . The impacts of interfacial bonding efficiency on SNF vibration integrity include the moment carrying capacity distribution between pellets and clad and the impact of cohesion on the flexural rigidity of the surrogate rod system. The result also indicates that the immediate consequences of interfacial de-bonding are a load carrying capacity shift from the fuel pellets to the clad and a reduction of the composite rod flexural rigidity. Therefore, the flexural rigidity of the surrogate rod and the bending moment bearing capacity between the clad and fuel pellets are strongly dependent on the efficiency of interfacial bonding at the pellet pellet and pellet clad interfaces. The above-noted phenomenon was calibrated and validated by reverse bending fatigue testing using a surrogate rod system.

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Surrogate Spent Nuclear Fuel Vibration Integrity Investigation

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During transportation, SNF experiences unique conditions and challenges to cladding integrity due to the vibrational and impact loading encountered during road or rail shipment. ORNL has been developing testing capabilities that can be used to improve our understanding of the impacts of vibration loading on SNF integrity, especially for high burn-up SNF in normal transportation operation conditions. This information can be used to meet nuclear industry and U.S. Nuclear Regulatory Commission needs in the area of safety of SNF storage and transportation operations.

bending fatigue↗

Development of flaw acceptance criteria for aging management of spent nuclear fuel multi-purpose canisters

A typical multipurpose canister (MPC) is made of austenitic stainless steel and is loaded with spent nuclear fuel assemblies. The canister may be subject to service-induced degradation when it is exposed to aggressive atmospheric environments during a possibly long-term storage period if the permanent repository is yet to be identified and readied. Because heat treatment for stress relief is not required for the construction of an MPC, stress corrosion cracking may be initiated on the canister surface in the welds or in the heat affected zone. An acceptance criteria methodology is being developed for flaw disposition should the crack-like defects be detected by periodic in-service Inspection. The first-order instability flaw sizes has been determined with bounding flaw configurations, that is, through-wall axial or circumferential cracks, and part-through-wall long axial flaw or 360° circumferential crack. The procedure recommended by the American Petroleum Institute (API) 579 Fitness-for-Service code (Second Edition) is used to estimate the instability crack length or depth by implementing the failure assessment diagram (FAD) methodology. The welding residual stresses are mostly unknown and are therefore estimated with the API 579 procedure. It is demonstrated in this paper that the residual stress has significant impact on the instability length or depth of the crack. The findings will limit the applicability of the flaw tolerance obtained from limit load approach where residual stress is ignored and only ligament yielding is considered.

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Passive Tomography for Spent Fuel Verification: Analysis Framework and Instrument Design Study

The potential for gamma emission tomography (GET) to detect partial defects within a spent nuclear fuel assembly is being assessed through a collaboration of Support Programs to the International Atomic Energy Agency (IAEA). In the first phase of this study, two safeguards verification objectives have been identified. The first is the independent determination of the number of active pins that are present in the assembly, in the absence of a priori information. The second objective is to provide quantitative measures of pin-by-pin properties, e.g. activity of key isotopes or pin attributes such as cooling time and relative burnup, for the detection of anomalies and/or verification of operator-declared data. The efficacy of GET to meet these two verification objectives will be evaluated across a range of fuel types, burnups, and cooling times, and with a target interrogation time of less than 60 minutes. The evaluation of GET viability for safeguards applications is founded on a modelling and analysis framework applied to existing and emerging GET instrument designs. Monte Carlo models of different fuel types are used to produce simulated tomographer responses to large populations of “virtual” fuel assemblies. Instrument response data are processed by a variety of tomographic-reconstruction and image-processing methods, and scoring metrics specific to each of the verification objectives are defined and used to evaluate the performance of the methods. This paper will provide a description of the analysis framework and evaluation metrics, example performance-prediction results, and describe the design of a “universal” GET instrument intended to support the full range of verification scenarios envisioned by the IAEA.

spent nuclear fuel assemblies, partial defect veri↗

Examining Supply Chain Resilience for the Intermodal Shipment of Spent Nuclear Fuel and High Level Radioactive Materials

The U.S. Department of Energy (DOE) has a significant programmatic interest in the safe and secure routing and transportation of Spent Nuclear Fuel (SNF) and High Level Waste (HLW) in the United States, including shipments entering the country from locations outside U.S borders. In any shipment of SNF/HLW, there are multiple chains; a jurisdictional chain as the material moves between jurisdictions (state, federal, tribal, administrative), a physical supply chain (which mode), as well as a custody chain (which stakeholder is in charge/possession) of the materials being transported. Given these interconnected networks, there lies vulnerabilities, whether in lack of communication between interested stakeholders or physical vulnerabilities such as interdiction. By identifying key links and nodes as well as administrative weaknesses, decisions can be made to harden the physical network and improve communication between stakeholders. This paper examines the parallel chains of oversight and custody as well as the chain of stakeholder interests for the shipments of SNF/HLW and the potential impacts on systemic resiliency. Using the Crystal River shutdown location as well as a hypothetical international shipment brought into the United States, this paper illustrates the parallel chains and maps them out visually.

law↗

Understanding the Risk of Chloride Induced Stress Corrosion Cracking of Interim Storage Containers for the Dry Storage of Spent Nuclear Fuel: Evolution of Brine Chemistry on the Container Surface.

Although the susceptibility of austenitic stainless steels to chloride-induced stress corrosion cracking is well known, uncertainties exist in terms of the environmental conditions that exist on the surface of the storage containers. While a diversity of salts is present in atmospheric aerosols, many of these are not stable when placed onto a heated surface. Given that the surface temperature of any container storing spent nuclear fuel will be well above ambient, it is likely that salts deposited on its surface may decompose or degas. To characterize this effect, relevant single and multi-salt mixtures are being evaluated as a function of temperature and relative humidity to establish the rates of degassing, as well as the likely final salt and brine chemistries that will remain on the canister surface.

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Examining supply chain resilience for the intermodal shipment of spent nuclear fuel and high level radioactive materials-16309

The U.S. Department of Energy (DOE) has a significant programmatic interest in the safe and secure routing and transportation of Spent Nuclear Fuel (SNF) and High Level Waste (HLW) in the United States, including shipments entering the country from locations outside U.S borders. In any shipment of SNF/HLW, there are multiple chains; a jurisdictional chain as the material moves between jurisdictions (state, federal, tribal, administrative), a physical supply chain (which mode), as well as a custody chain (which stakeholder is in charge/possession) of the materials being transported. Given these interconnected networks, there lies vulnerabilities, whether in lack of communication between interested stakeholders or physical vulnerabilities such as interdiction. By identifying key links and nodes as well as administrative weaknesses, decisions can be made to harden the physical network and improve communication between stakeholders. This paper examines the parallel chains of oversight and custody as well as the chain of stakeholder interests for the shipments of SNF/HLW and the potential impacts on systemic resiliency. Here, by using the Crystal River shutdown location as well as a hypothetical international shipment brought into the United States, this paper illustrates the parallel chains and maps them out visually.

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Development of Transient Coupled Chemical-Thermal-Fluids Simulation for Unsealed Aluminum-Clad Spent Nuclear Fuel Storage Canister

This study examines a preliminary three-dimensional (3D) multi-physics computational fluid dynamics (CFD) model of the unsealed canisters containing aluminum-clad spent nuclear fuel from INL’s ATR in the irradiated fuel storage facility. This 3D CFD model considers fully resolved fuel plates in an unsealed canister with small mass exchange with surrounding air as time evolves in the simulation and contains a network of chemical reactions to consider due to several of the side reactions occurring in the presence of water vapor, nitrogen, oxygen and carbon dioxide rather than in helium backfilled DOE standard canisters. Prior to the CFD modeling, the full set of chemical reactions under these conditions was analyzed, reduced, and refit to a more manageable number of equations for the computational efficiency of the 3D CFD simulation. To deal with the long time-scale of interest for tracking the possible buildup of trace species such as nitric acid, NOx, and hydrogen, a segregated time stepping algorithm is used, wherein the velocity field is only updated periodically, and assume to be in a pseudo steady state. This study first examines the theory of the model, workflow to establish radiolysis reaction network, and initial simulations of the evolutions of thermal fields, hydrogen gas and nitric acid concentrations within unsealed canisters in conditions relevant to the INL INTEC CPP-603 facility over a long period of time.

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Development of Transient Coupled Chemical-Thermal-Fluids Simulation for Unsealed Aluminum-Clad Spent Nuclear Fuel Storage Canister

This study examines a preliminary three-dimensional (3D) multi-physics computational fluid dynamics (CFD) model of the unsealed canisters containing aluminum-clad spent nuclear fuel from INL’s ATR in the irradiated fuel storage facility. This 3D CFD model considers fully resolved fuel plates in an unsealed canister with small mass exchange with surrounding air as time evolves in the simulation and contains a network of chemical reactions to consider due to several of the side reactions occurring in the presence of water vapor, nitrogen, oxygen and carbon dioxide rather than in helium backfilled DOE standard canisters. Prior to the CFD modeling, the full set of chemical reactions under these conditions was analyzed, reduced, and refit to a more manageable number of equations for the computational efficiency of the 3D CFD simulation. To deal with the long time-scale of interest for tracking the possible buildup of trace species such as nitric acid, NOx, and hydrogen, a segregated time stepping algorithm is used, wherein the velocity field is only updated periodically, and assume to be in a pseudo steady state. This study first examines the theory of the model, workflow to establish radiolysis reaction network, and initial simulations of the evolutions of thermal fields, hydrogen gas and nitric acid concentrations within unsealed canisters in conditions relevant to the INL INTEC CPP-603 facility over a long period of time.

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Spent Nuclear Fuel Transportation: Rail Conveyance Transmissibility

Equipos Nucleares, S.A. (ENSA) and the US Department of Energy (DOE) are preparing a full scale Spent Nuclear Fuel (SNF) transportation test. This transportation testing will include road, rail, coastal, and trans-Atlantic shipments. The test campaign will use a full scale commercial transportation package and cradle. The package will be loaded with two instrumented fuel assemblies to measure strains at cladding locations and accelerations on the fuel assemblies, and thirty “dummy” assemblies. This testing is designed to closely match an actual SNF shipment. Accelerometers will also be used at various locations throughout the full conveyance system (i.e. rail car/truck, cask, transport cradle, and basket) to study the transmission of loads through the system and to provide validation for numerical models. Previous testing and modeling work has shown how the structural transmissibility of the transport system can affect the magnitude of these loads, and the importance of modeling all aspects of the transport system (i.e. rail car/truck, transport cradle, cask, basket, and fuel) (Ref 1, 2). This paper describes preliminary models that were constructed to estimate load transmission during rail transport, from the bottom of the cradle to an individual fuel rod within the package. The modeling studies in this paper evaluate the system response to postulated shock pulses and random vibration loads. These models describe the transmissibility of the conveyance, and demonstrate how loads can be amplified or attenuated as they are transmitted through the structure. This is done by coupling the rail vehicle dynamics code NUCARS, to the general finite element modeling code ANSYS, and the explicit dynamics code LS-DYNA. Models such as the ones presented herein will be used during the test campaign to help analyze and evaluate the test data as it is collected.

Cladding, Spent Fuel and Radiactive Waste, Used Fu↗

A compact breed and burn fast reactor using spent nuclear fuel blanket

A long-life breed-and-burn (B and B) type fast reactor has been investigated from the neutronics points of view. The B and B reactor has the capability to breed the fissile fuels and use the bred fuel in situ in the same reactor. In this work, feasibility of a compact sodium-cooled B and B fast reactor using spent nuclear fuel as blanket material has been studied. In order to derive a compact B and B fast reactor, a tight fuel lattice and relatively large fuel pin are used to achieve high fuel volume fraction. The core is initially loaded with an LEU (Low Enriched Uranium) fuel and a metallic fuel is used in the core. The Monte Carlo depletion has been performed for the core to see the long-term behavior of the B and B reactor. Several important parameters such as reactivity coefficients, delayed neutron fraction, prompt neutron generation lifetime, fission power, and fast neutron fluence, are analyzed through Monte Carlo reactor analysis. Evolution of the core fuel composition is also analyzed as a function of burnup. Although the long-life small B and B fast reactor is found to be feasible from the neutronics point of view, it is characterized to have several challenging technical issues including a very high fast neutron fluence of the structural materials. (authors)

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The Environmental Protection Agency's Safety Standards for Disposal of Spent Nuclear Fuel: Potential Path Forward in Response to the Report of the Blue Ribbon Commission on America's Nuclear Future - 13388

Following the decision to withdraw the Yucca Mountain license application, the Department of Energy created a Blue Ribbon Commission (BRC) on America's Nuclear Future, tasked with recommending a national strategy to manage the back end of the nuclear fuel cycle. The BRC issued its final report in January 2012, with recommendations covering transportation, storage and disposal of spent nuclear fuel (SNF); potential reprocessing; and supporting institutional measures. The BRC recommendations on disposal of SNF and high-level waste (HLW) are relevant to the U.S. Environmental Protection Agency (EPA), which shares regulatory responsibility with the Nuclear Regulatory Commission (NRC): EPA issues 'generally applicable' performance standards for disposal repositories, which are then implemented in licensing. For disposal, the BRC endorses developing one or more geological repositories, with siting based on an approach that is adaptive, staged and consent-based. The BRC recommends that EPA and NRC work cooperatively to issue generic disposal standards-applying equally to all sites-early in any siting process. EPA previously issued generic disposal standards that apply to all sites other than Yucca Mountain. However, the BRC concluded that the existing regulations should be revisited and revised. The BRC proposes a number of general principles to guide the development of future regulations. EPA continues to review the BRC report and to assess the implications for Agency action, including potential regulatory issues and considerations if EPA develops new or revised generic disposal standards. This review also involves preparatory activities to define potential process and public engagement approaches. (authors)

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FATE Unified Modeling Method for Spent Nuclear Fuel and Sludge Processing, Shipping and Storage - 13405

A unified modeling method applicable to the processing, shipping, and storage of spent nuclear fuel and sludge has been incrementally developed, validated, and applied over a period of about 15 years at the US DOE Hanford site. The software, FATE{sup TM}, provides a consistent framework for a wide dynamic range of common DOE and commercial fuel and waste applications. It has been used during the design phase, for safety and licensing calculations, and offers a graded approach to complex modeling problems encountered at DOE facilities and abroad (e.g., Sellafield). FATE has also been used for commercial power plant evaluations including reactor building fire modeling for fire PRA, evaluation of hydrogen release, transport, and flammability for post-Fukushima vulnerability assessment, and drying of commercial oxide fuel. FATE comprises an integrated set of models for fluid flow, aerosol and contamination release, transport, and deposition, thermal response including chemical reactions, and evaluation of fire and explosion hazards. It is one of few software tools that combine both source term and thermal-hydraulic capability. Practical examples are described below, with consideration of appropriate model complexity and validation. (authors)

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