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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Validation Study for Crediting Chlorine in Criticality Analyses for US Spent Nuclear Fuel Disposition

Spent nuclear fuel (SNF) management practices in the United States rely on dry storage systems that include both canister- and cask-based systems. The United States Department of Energy Used Fuel Disposition Campaign is examining the feasibility of direct disposal of dual-purpose (storage and transportation) canisters (DPCs) in a geological repository. One of the major technical challenges for direct disposal is the ability to demonstrate the subcriticality of the DPCs loaded with SNF for the repository performance period (e.g., 10,000 years or more) as the DPCs may undergo degradation over time. Specifically, groundwater ingress into the DPC (i.e., flooding) could allow the system to achieve criticality in scenarios where the neutron absorber plates in the DPC basket have degraded. However, as was shown by Banerjee et al., some aqueous species in the groundwater provide noticeable reactivity reduction for these systems. For certain amounts of particular aqueous species (e.g., chlorine, lithium) in the groundwater, subcriticality can be demonstrated even for DPCs with complete degradation of the neutron absorber plates or a degraded fuel basket configuration. It has been demonstrated that chlorine is the leading impurity, as indicated by significant neutron absorption in the water that is available in reasonable quantities for the deep geological repository media under consideration. This paper presents the results of an investigation of the available integral experiments worldwide that could be used to validate DPC disposal criticality evaluations, including credit for chlorine. Due to the small number of applicable critical configurations, validation through traditional trending analysis was not possible. The bias in the eigenvalue of the application systems due only to the chlorine was calculated using TSURFER analysis and found to be on the order of 100 percent mille (1 pcm = 10 -5 k eff ). This study investigated the design of a series of critical configurations with varying amounts of chlorine to address validation gaps. Such integral experiments would support the crediting of the chlorine neutron-absorption properties in groundwater and the demonstration of subcriticality for DPCs in deep geologic repositories with sufficient chlorine availability.

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DOE OSTI2015

In-Field Performance Testing of the Fork Detector for Quantitative Spent Fuel Verification

Expanding spent fuel dry storage activities worldwide are increasing demands on safeguards authorities that perform inspections. The European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) require measurements to verify declarations when spent fuel is transferred to difficult-to-access locations, such as dry storage casks and the repositories planned in Finland and Sweden. EURATOM makes routine use of the Fork detector to obtain gross gamma and total neutron measurements during spent fuel inspections. Data analysis is performed by modules in the integrated Review and Analysis Program (iRAP) software, developed jointly by EURATOM and the IAEA. Under the framework of the US Department of Energy–EURATOM cooperation agreement, a module for automated Fork detector data analysis has been developed by Oak Ridge National Laboratory (ORNL) using the ORIGEN code from the SCALE code system and implemented in iRAP. EURATOM and ORNL recently performed measurements on 30 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel (Clab), operated by the Swedish Nuclear Fuel and Waste Management Company (SKB). The measured assemblies represent a broad range of fuel characteristics. Neutron count rates for 15 measured pressurized water reactor assemblies are predicted with an average relative standard deviation of 4.6%, and gamma signals are predicted on average within 2.6% of the measurement. The 15 measured boiling water reactor assemblies exhibit slightly larger deviations of 5.2% for the gamma signals and 5.7% for the neutron count rates, compared to measurements. These findings suggest that with improved analysis of the measurement data, existing instruments can provide increased verification of operator declarations of the spent fuel and thereby also provide greater ability to confirm integrity of an assembly. These results support the application of the Fork detector as a fully quantitative spent fuel verification technique.

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DOE OSTI2013

US NRC-Sponsored Research on Stress Corrosion Cracking Susceptibility of Dry Storage Canister Materials in Marine Environments - 13344

At a number of locations in the U.S., spent nuclear fuel (SNF) is maintained at independent spent fuel storage installations (ISFSIs). These ISFSIs, which include operating and decommissioned reactor sites, Department of Energy facilities in Idaho, and others, are licensed by the U.S. Nuclear Regulatory Commission (NRC) under Title 10 of the Code of Federal Regulations, Part 72. The SNF is stored in dry cask storage systems, which most commonly consist of a welded austenitic stainless steel canister within a larger concrete vault or overpack vented to the external atmosphere to allow airflow for cooling. Some ISFSIs are located in marine environments where there may be high concentrations of airborne chloride salts. If salts were to deposit on the canisters via the external vents, a chloride-rich brine could form by deliquescence. Austenitic stainless steels are susceptible to chloride-induced stress corrosion cracking (SCC), particularly in the presence of residual tensile stresses from welding or other fabrication processes. SCC could allow helium to leak out of a canister if the wall is breached or otherwise compromise its structural integrity. There is currently limited understanding of the conditions that will affect the SCC susceptibility of austenitic stainless steel exposed to marine salts. NRC previously conducted a scoping study of this phenomenon, reported in NUREG/CR-7030 in 2010. Given apparent conservatisms and limitations in this study, NRC has sponsored a follow-on research program to more systematically investigate various factors that may affect SCC including temperature, humidity, salt concentration, and stress level. The activities within this research program include: (1) measurement of relative humidity (RH) for deliquescence of sea salt, (2) SCC testing within the range of natural absolute humidity, (3) SCC testing at elevated temperatures, (4) SCC testing at high humidity conditions, and (5) SCC testing with various applied stresses. Results to date indicate that the deliquescence RH for sea salt is close to that of MgCl{sub 2} pure salt. SCC is observed between 35 and 80 deg. C when the ambient (RH) is close to or higher than this level, even for a low surface salt concentration. (authors)

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DOE OSTI2013

Unique Chernobyl Cranes for Deconstruction Activities in the New Safe Confinement - 13542

The devastation left behind from the Chernobyl nuclear power plant (ChNPP) Unit 4 accident which occurred on April 26, 1986 presented unparalleled technical challenges to the world engineering and scientific community. One of the largest tasks that are in progress is the design and construction of the New Safe Confinement (NSC). The NSC is an engineered enclosure for the entire object shelter (OS) that includes a suite of process equipment. The process equipment will be used for the dismantling of the destroyed Chernobyl Nuclear Power Plant (ChNPP) Unit. One of the major mechanical handling systems to be installed in the NSC is the Main Cranes System (MCS). The planned decontamination and decommissioning or dismantling (D and D) activities will require the handling of heavily shielded waste disposal casks containing nuclear fuel as well as lifting and transporting extremely large structural elements. These activities, to be performed within the NSC, will require large and sophisticated cranes. The article will focus on the unique design features of the MCS for the D and D activities. (authors)

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DOE OSTI2013

Facility Utilization and Risk Analysis for Remediation of Legacy Transuranic Waste at the Savannah River Site - 13572

Savannah River Nuclear Solutions (SRNS) completed the Accelerated TRU Project for remediating legacy waste at the Savannah River Site with significant cost and schedule efficiencies due to early identification of resources and utilization of risk matrices. Initial project planning included identification of existing facilities that could be modified to meet the technical requirements needed for repackaging and remediating the waste. The project schedule was then optimized by utilization of risk matrices that identified alternate strategies and parallel processing paths which drove the overall success of the project. Early completion of the Accelerated TRU Project allowed SRNS to pursue stretch goals associated with remediating very difficult TRU waste such as concrete casks from the hot cells in the Savannah River National Laboratory. Project planning for stretch goals also utilized existing facilities and the risk matrices. The Accelerated TRU project and stretch goals were funded under the American Recovery and Reinvestment Act (ARRA). (authors)

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2013

Safety of interim storage solutions of used nuclear fuel during extended term

In 2013, the total amount of stored used nuclear fuel (UNF) in the world will reach 225,000 T HM. The UNF inventory in wet storage will take up over 80% of the available total spent fuel pool (SFP) capacity. Interim storage solutions are needed. They give flexibility to the nuclear operators and ensure that nuclear reactors continue to operate. However, we need to keep in mind that they are also an easy way to differ final decision and implementation of a UNF management approach (recycling or final disposal). In term of public perception, they can have a negative impact overtime as it may appear that nuclear industry may have significant issues to resolve. In countries lacking an integrated UNF management approach, the UNF are being discharged from the SFPs to interim storage (mostly to dry storage) at the same rate as UNF is being discharged from reactors, as the SFPs at the reactor sites are becoming full. This is now the case in USA, Taiwan, Switzerland, Spain, South Africa and Germany. For interim storage, AREVA has developed different solutions in order to allow the continued operation of reactors while meeting the current requirements of Safety Authorities: -) Dry storage canisters on pads, -) Dual-purpose casks (dry storage and transportation), -) Vault dry storage, and -) Centralized pool storage.

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DOE OSTI2015

EURATOM safeguards efforts in the development of spent fuel verification methods by non-destructive assay

The back end of the nuclear fuel cycle continues to develop. The European Commission, particularly the Nuclear Safeguards Directorate of the Directorate General for Energy, implements Euratom safeguards and needs to adapt to this situation. The verification methods for spent nuclear fuel, which EURATOM inspectors can use, require continuous improvement. Whereas the Euratom on-site laboratories provide accurate verification results for fuel undergoing reprocessing, the situation is different for spent fuel which is destined for final storage. In particular, new needs arise from the increasing number of cask loadings for interim dry storage and the advanced plans for the construction of encapsulation plants and geological repositories. Various scenarios present verification challenges. In this context, EURATOM Safeguards, often in cooperation with other stakeholders, is committed to further improvement of NDA methods for spent fuel verification. In this effort EURATOM plays various roles, ranging from definition of inspection needs to direct participation in development of measurement systems, including support of research in the framework of international agreements and via the EC Support Program to the IAEA. This paper presents recent progress in selected NDA methods. These methods have been conceived to satisfy different spent fuel verification needs, ranging from attribute testing to pin-level partial defect verification. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Underwater characterization of control rods for waste disposal using SMOPY

Storage of spent fuel assemblies in cooling ponds requires careful control of the geometry and proximity of adjacent assemblies. Measurement of the fuel burnup makes it possible to optimise the storage arrangement of assemblies taking into account the effect of the burnup on the criticality safety margins ('burnup credit'). Canberra has developed a measurement system for underwater measurement of spent fuel assemblies. This system, known as 'SMOPY', performs burnup measurements based on gamma spectroscopy (collimated CZT detector) and neutron counting (fission chamber). The SMOPY system offers a robust and waterproof detection system as well as the needed capability of performing radiometric measurements in the harsh high dose - rate environments of the cooling ponds. The gamma spectroscopy functionality allows powerful characterization measurements to be performed, in addition to burnup measurement. Canberra has recently performed waste characterisation measurements at a Nuclear Power Plant. Waste activity assessment is important to control costs and risks of shipment and storage, to ensure that the activity level remains in the range allowed by the facility, and to declare activity data to authorities. This paper describes the methodology used for the SMOPY measurements and some preliminary results of a radiological characterisation of AIC control rods. After describing the features and normal operation of the SMOPY system, we describe the approach used for establishing an optimum control rod geometric scanning approach (optimum count time and speed) and the method of the gamma spectrometry measurements as well as neutron check measurements used to verify the absence of neutron sources in the waste. We discuss the results obtained including {sup 60}Co, {sup 110m}Ag and {sup 108m}Ag activity profiles (along the length of the control rods) and neutron results including Total Measurement Uncertainty evaluations. Full self-consistency checks were performed and these demonstrate the validity of the techniques. The results are described and analysed in the context of the measurement performance of the equipment. Different casks were fully characterized using a 60 mm{sup 3} CZT detector, to determine the total activities and spatial profiles. A total activity range measurement of 1x10{sup 8} - 1x10{sup 13} Bq/cm was found to be achievable. Finally, comments are made, based on our measurements, on the ability of this equipment for performing in-situ characterisation of wastes in the harsh environments typical of fuel assembly and waste storage ponds and silos. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

The fuel integrity analysis of transfer processes for Chinshan dry-storage system by FRAPTRAN/CFD

The dry cask storage operation is separated into 3 stages: handling, transfer, and storage. In this study, the handling and transfer stages were established by CFD (Computational Fluid Dynamics). In order to obtain the burnup information of a single rod in the core, a single rod of Chinshan NPP was modeled by FRAPCON-3.5 and FRAPTRAN-1.5. Then, using the boundary condition from the CFD analysis, the fuel behaviors and performances were analyzed in detail with the FRAPTRAN-1.5 code. Three different thermal loads were considered in this paper: 7 kW, 14 kW, and 21 kW. The simulation results showed that, in all the cases: first, during the handling and transfer stages, the maximum cladding temperature was in the range of 477 to 653 K; secondly, the maximum cladding hoop strain was in the range of 0.0018 to 0.0025; and thirdly, the structural radial gap was almost kept as a constant. Based on these results, the fuel cladding integrity was maintained in the handling and transfer stages. These results help evaluating the safety issue of the dry-storage system for Chinshan NPP. (authors)

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DOE OSTI2015

Comparison of German and American fabrication standards for transportation and storage containments of high level radioactive material type B packages - 15055

As the competent authority for the evaluation of German radioactive material (RAM) transport and storage packages, BAM (the German Federal Institute for Materials Research and Testing) has evaluated various international cask manufacturers regarding compliance with the German Type B(U) licensing requirements. The US NRC has recommended paragraphs in the ASME Boiler and Pressure Vessel (B and PV) Code, Section III, Division 3, as the measure to quantify the safety requirements of the US regulations regarding fabrication of RAM transportation and storage containments. The objective of this study is to compare the fabrication criteria of the German Dangerous Goods Procedural Rules (BAM-GGR) and the German Institute for Standardization (DIN) with the US quantitative measures for fabrication of RAM package containment vessels. This comparison will cover major similarities and differences between the requirements of the German standards and the ASME Code requirements for Type B package containment fabrication QC (quality control). While the material most often used for German RAM containment applications is ductile cast iron (DCI), in the US forged austenitic stainless steel is typically used. This study focuses on fabrication QC measures required to ensure reproducible and acceptable properties for DCI containments. (authors)

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DOE OSTI2015

Load planning in the dark: RH-TRU waste container load planning at Sandia National Laboratories/New Mexico - 15116

Sandia National Laboratories (Sandia) performed several sets of experiments during the 1970's and 1980's with fresh and irradiated mixed oxide (MOX) and uranium oxide (UO{sub 2}) materials designed to simulate various fuel responses under varying reactor conditions. These research materials were from light water and breeder mixed oxide reactors. The experimental vessels and extraneous piece parts resulting from the experiments were identified as eligible for disposal at the Waste Isolation Pilot Plant (WIPP) as remote-handled (RH) transuranic (TRU) waste. Once Sandia received the approval to dispose at WIPP, discussions began with the Carlsbad Field Office (CBFO) and Nuclear Waste Partnerships, LLC (NWP) to provide Central Characterization Program (CCP) support for certified acceptable knowledge, visual examination, and dose-to-curie for the repackaging effort. In addition to the WIPP support, Sandia was required to prepare a load plan describing the configuration of the parent containers and how they could be repackaged to meet WIPP and other regulatory requirements. However, all of these containers were packaged as material twenty (20) or thirty (30) years ago and the records of the packaging process and configuration were not documented as waste. Therefore, it seemed that the load plans were prepared with limited information regarding parent container configuration, hence in the dark. The design of the Auxiliary Hot Cell Facility (AHCF) in technical area (TA) V, where the repackaging was to take place, along with the shipping requirements for the RH 72-B cask, limited the options for repackaging. This configuration required that 0.114 m{sup 3} (30-gallon) containers be packaged in the hot cell, removed and placed into 0.208 m{sup 3} (55-gallon) containers to mitigate contamination issues. The material had been tracked as accountable nuclear material and Sandia had verified Material Control and Accountability (MC and A) data on each parent container. Based on the MC and A data, scaling factors could be determined and sampling and analysis was not required. However, it meant that parent containers could not be split into two daughter containers, though two or more parent containers could be combined into one daughter. (authors)

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DOE OSTI2015

Andra's strategy and approach for operational safety of French underground facility Cigeo - 15127

According to the French Act on radioactive waste of 28 June 2006, Andra shall design a reversible repository. In response to this demand, Andra is developing the industrial project namely 'Cigeo', a reversible geological disposal facility for HLW and ILW located in Meuse/Haute-Marne. The previous safety cases of 2005, 2009 and 2012 give a sound basis to assess Cigeo's safety both for the operational and post-closure periods. Regarding operational safety, in this new stage of the project, the challenging issues for the preparation of the safety case are the following: - To adapt the various regulatory frameworks (nuclear and non-nuclear) to the underground facility - To design the underground facility and handling equipment's to reduce at first risks in the underground facility as such radioactive particles release, gas explosion, fire and co-activity and limit their consequence. - To establish waste acceptance criteria to be fulfilled by waste producers in line with operational safety and design options. The objective of the paper is to present the experience of Andra in operation safety of the underground facility of Cigeo. Operational safety for Cigeo is assessed as much as possible according to the existing French regulations and guidance for nuclear facilities. Indeed, compared to existing nuclear facilities, a geological repository is a specific case due to the existence of underground facilities with particularities such as: - ventilation management in a large scale underground facility, - fire risk management. As an input for the industrial design of Cigeo for operation, Andra defined a fire guidance and a specific reference safety guidance, based on the adaptation of the French nuclear regulatory texts for and when feasible to mines and tunnels regulatory texts and on its own experience gained through safety operational reports since 2005 and their reviews by the Safety Authority. - underground co-activity of nuclear operation with construction works. According to the common 'defense-in depth' principle for nuclear facilities, these guidelines for the industrial design of Cigeo for operation consider the successive levels of defense of: - Preventing the risk and its spreading by design and construction; - Detecting incidents and early fighting by design and operation dispositions (detection and control systems, survey procedure); - Controlling incidents and limiting of their consequences by safeguarding systems, specific procedures. At that stage, design options for Cigeo constitute an adequate response to the operational safety functions and requirements: - Two levels of containment are set up: the first one is ensured by the waste package, the second one is ensured from the surface facility to the disposal vaults by the transfer cask during the transfer of waste packages in the disposal vaults by the disposal vault itself and a nuclear ventilation with HEPA filter; - Favorable options for fire risk prevention and mitigation: transport system minimizing heat load (cable car for the transfer from the surface, electric transfer engines), twin access tunnels to ease fire-fighting and evacuation...; - Permanent complete physical separation of nuclear zone and construction zone (including ventilation systems). In practice, operation safety assessment of Cigeo, designed following the aforesaid reference safety guidelines, follows a common risk analysis (internal aggressions such as auxiliary losses, chemical and toxic risks, fall or collision, fire, human factors and external aggressions such as earthquake, air traffic, flooding) with the following steps: - Identification of the major risks and the design basis scenarios that have to be considered for dimensioning of the installations and the safety devices; - On the basis of these different situations, those who will have to be quantified are identified; one or more representative and envelop situations are chosen in order to evaluate robustness of the installation. Quantitative evaluation of these situations permits to: - identify the consequences on man, environment and installation, and in particular to gain confidence in the maintaining of the safety functions; - identify or define the technic and/or organisational dispositions classified as 'important elements for protection' (IEP); - Situations beyond dimensioning permit to verify the robustness of the installation. They permit, on one hand to identify possible complementary material or organisational dispositions; on the other hand, to define the 'hard core' i.e. structures, systems and components that permit to maintain the confinement function of radioactive substance. (authors)

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DOE OSTI2015

System impacts of accelerated transfer of used nuclear fuel to dry storage - 15220

In the wake of the September 11, 2001 terror attacks, and again after the March 11, 2011, accident at the Fukushima Dai-Ichi station from the Great East Japan Earthquake/tsunami, concern was expressed regarding security of current spent fuel storage practices in the United States. An approach recurrently proposed to reduce the potential for fire and the release of radionuclides in an accident or sabotage scenario is to reduce the loading of the reactor fuel pools, which are frequently maintained near their full capacity at many reactor sites in the U.S. We assess the impact on the waste management system of different scenarios for accelerating the schedule of used nuclear fuel (UNF) transfer from reactor pools to dry storage casks. We consider scenarios where the spent fuel pools are drawn down to 50% or 25% of full capacity over 5- or 10-year schedules. The scenarios include status quo, where UNF continues to remain in dry storage at the reactor sites, as well as scenarios with UNF transfer to interim storage and eventually to a mined geologic repository. The primary purpose is to evaluate how the different scenarios for accelerating the transfer of UNF to at-reactor dry storage would affect the performance of the entire UNF management system. Areas that were considered include an estimate of the additional system costs that could be incurred and where the existing infrastructure at the reactor sites would come under stress from the transfer demands. This study considered only the fuel-handling logistics and rough-order-of-magnitude cost at reactor sites, interim storage facility, and operations to package UNF into disposal canisters. The effects of minimum fuel age that could be transferred and the accelerated transfer starting date are also examined, but found to have relatively little effect. Impacts on safety and worker dose are not evaluated quantitatively, although effects can be inferred from increases in handling operations, etc. from the scenarios considered. (authors)

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DOE OSTI2015

The German product quality control of the compacted metallic nuclear waste - 15311

The treatment and safe disposal of spent nuclear fuel is a key challenge of nuclear energy utilization and has to be accomplished regardless the future nuclear of electricity production or phase-out policies. Until June 2005 approx. 6662 tons of spent fuel from German nuclear power plants was sent to the La Hague (France) and Sellafield (UK) plants for reprocessing. An equivalent amount of nuclear waste compounds has contractually to be returned to Germany, which is currently in progress. The returnable waste streams comprise mainly highly-active vitrified waste containing a major part of fission products and actinides, and intermediate-active super-compacted metallic waste containing reprocessed fuel element's structural materials and some technological waste from the reprocessing process itself. Shearing and dissolution of spent fuel leaves scraps of fuel metal cladding, the end-pieces and spacers which are contaminated, activated and irradiating from their life in a reactor core. In fact, this metallic intermediate level radioactive waste makes up a bit more than half of the total volume of the reprocessed waste. At La Hague the metallic waste is super-compacted. Several, typically eight, compacted discs are stacked in a stainless steel container, named CSD-C as for 'Colis Standard de Dechets Compactes'. 4104 containers of such CSD-Cs in 152 transport/storages casks have to be returned to Germany. Each one of them has to be checked and quality certified for its conformity with German national product specifications and acceptance criteria by independent national product quality control inspectors (PKS). The residue package documentation has to be complete and the characterization methods have to be comprehensive. Due to the heterogeneity of metallic waste the characterization and control of its properties is very complex. The nuclear characterization is performed by means of non-destructive measurements, where only a few key nuclides are measured directly. In fact, the proper calculation of the whole nuclear inventory relies on numerous supplementary simulations. The measurement and calculation methodology is scrutinized by PKS in detail and key points are benchmarked with our own simulations for the sake of an independent nuclear inventory declaration check. With respect to the vast amount of documentation and parameters, which have to be checked and verified, PKS had to optimize their checking procedures. For this purpose dedicated proof tools have been developed that shall be applied for selected indicative cross-checks of the relevant parameters. This contribution is dedicated to the procedure of quality control by PKS and examples of actually performed computational verifications of declared nuclear inventory. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

French management model for sustainable transportation of used fuel - 15376

Safe and efficient transportation of radioactive material is key to the success of our industry. Used fuel transports are inherently complex due to the nature of the material and the high visibility. Only a few countries have developed comprehensive used fuel transportation systems that require a wide range of equipment, services and expertise. In the many countries, such as the Unites States, where interim consolidated storage, recycling or geological repository are not available yet, similar systems will have to be implemented. For more than 50 years AREVA TN has safely shipped more than 7,000 used fuel transport casks. The transportation system that was initially developed in the 1970's has been adapted and enhanced over the years to meet more restrictive regulatory requirements and evolving customer needs, and to address public concerns. The numerous 'lessons learned' have offered data and guidance that have allowed for also efficient and consistent improvement over the decades. Based on AREVA TN extensive expertise, this paper will describe the different phases and milestones that need to be met to set up, license and operate a successful used fuel transportation management system. Transportation of spent fuel in France could not be performed effectively and efficiently without strong collaboration with Electricite de France (EDF) who plays a critical role as the shipper. EDF will provide a nuclear operator's perspective on the most critical aspects of a successful used fuel transportation program. (authors)

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DOE OSTI2015

Developing generic design alternatives for dry storage of used nuclear fuel at an interim storage facility - 15480

The Department of Energy (DOE) is beginning to lay the groundwork for implementing interim storage of used nuclear fuel (UNF) as recommended by the Blue Ribbon Commission on America's Nuclear Future in their report to the Secretary of Energy published on January 26, 2012. These plans include activities to 1) establish one or more Interim Storage Facilities (ISF) using consent-based siting, and 2) prepare for large-scale transport of UNF. The Administration released its Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (Strategy) on January 11, 2013. The Strategy includes a phased, adaptive, and consent-based approach to siting and implementing a comprehensive management and disposal system. The Strategy calls for an operational Pilot ISF by 2021, a Larger ISF by 2025, and a geological repository by 2048. The DOE Nuclear Fuels Storage and Transportation Planning Project (NFST) prepared a report in April 2013 titled, -A Project Concept for Nuclear Fuels Storage and Transportation.- This report developed the functions and requirements necessary for development of a Pilot and Larger ISF to meet the framework outlined in the Strategy as follows: - The UNF inventory is limited for the Pilot and Larger ISF as described in the Strategy. - It is assumed that bare fuel will be received in reusable transportation casks in significant quantities for the larger ISF and will not be solely limited to the receipt of fuel in dry storage canister (DSC) as is the Pilot ISF. This assumption was made because the method for accepting UNF from the utilities once the larger ISF begins operating has not yet been decided. - UNF storage alternative concepts identified in previous reports must be more fully developed to support alternative analysis. The purpose of this topic is to report on the development of the generic design alternatives that have been evaluated for the receipt and storage of the UNF at the initial Pilot ISF and Larger ISF. The generic design alternatives have been developed over the past few months by collaboration between the DOE and industry experts experienced in UNF facilities. The design alternatives will enable volunteer host communities to investigate generic design alternatives that could be fully developed for a site-specific location based on their unique design requirements. The design alternatives provide flexibility for additional capabilities to be developed and executed in a modular fashion such that expansion from a Pilot ISF to a Larger ISF could be achieved in an orderly manner. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Streamlining analysis capabilities for SNF management - 15490

A comprehensive, integrated data and analysis tool-the Used Nuclear Fuel-Storage, Transportation and Disposal Analysis Resource and Data System (UNF-ST and DARDS) -is being developed for the US DOE Office of Nuclear Energy (DOE-NE) Nuclear Fuels Storage and Transportation Planning Project (NFST). The overarching goal of UNF-ST and DARDS is to provide a comprehensive controlled source of technical data integrated with key analysis capabilities to characterize inputs to the overall US waste management system from reactor power production through ultimate disposition. UNF-ST and DARDS seamlessly integrates a unified SNF system relational database and key analysis capabilities to simplify and automate the performance of accurate and efficient SNF related analyses to support numerous DOE waste management and fuel cycle activities. Hence, UNF-ST and DARDS enables decision making relative to design, safety, and licensing of SNF systems and facilities by providing the best information available. Automated analysis sequences have been deployed to characterize the discharge inventory and perform criticality, thermal, and dose analyses using time-dependent data. UNF-ST and DARDS provides cask-specific as-loaded safety analysis including criticality, thermal and dose evaluations, as well as results visualization. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Preliminary evaluation of removing used nuclear fuel from shutdown sites - 15568

A preliminary evaluation of removing spent nuclear fuel (SNF) from 12 shutdown nuclear power reactor sites was conducted. 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, and San Onofre. 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 and conditions relevant to transportation of SNF and GTCC waste, (3) an evaluation of the near-site transportation infrastructure and experience relevant to shipping transportation casks containing SNF and GTCC waste, including identification of gaps in information, and (4) an evaluation of the actions necessary to prepare for and remove SNF and GTCC waste. Every 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. It is expected that additional site visits will be conducted to add to the information presented in the evaluation. (authors)

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