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Multi-Pack Disposal Concepts for Spent Fuel (Revision 1)

At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media. Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design. Thermal analysis showed that if “enclosed” concepts are constrained by peak package/buffer temperature, that waste package capacity is limited to 4 PWR assemblies (or 9 BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems. This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).

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Multi-pack Disposal Concepts for Spent Fuel (Rev. 0)

At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media (Hardin et al., 2012). Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design (CRWMS M&O, 1999). Thermal analysis showed that, if “enclosed” concepts are constrained by peak package/buffer temperature, waste package capacity is limited to 4 PWR assemblies (or 9-BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems (EnergySolution, 2015). This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).

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Consolidated Interim Storage Advantages and Disadvantages from Prior Reports and Studies

The question of whether centralized storage of civilian spent nuclear fuel (SNF) should be part of the federal waste management system as an intermediate step before permanent disposal has been debated for more than four decades. Since the Nuclear Waste Policy Act (NWPA) (U.S. House of Representatives 1982) was enacted, the need for interim storage facilities (ISFs) as part of the federal waste management system has been analyzed and documented in numerous reports and studies. This paper summarizes the advantages and disadvantages as described in those reports of incorporating an ISF into the waste management system.

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Waste Generation Overview Refresher, Course 21464

This course, Waste Generation Overview Refresher (COURSE 21464), provides an overview of federal and state waste management regulations, as well as Los Alamos National Laboratory (LANL) policies and procedures for waste management operations. The course covers the activities involved in the cradle-to- grave waste management process and focuses on waste characterization, waste compatibility determinations and classification, and the storage requirements for temporary waste accumulation areas at LANL.

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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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Building Effective Collaborations to Bring Innovation into Waste Management and Decommissioning - 16477

A strong record of research collaboration has been developing between the UK nuclear industry and key universities with nuclear specialism over the past 15 years or so. The DISTINCTIVE program, focusing on decommissioning, immobilization and storage solutions for nuclear waste inventories is an example. It is a consortium of eleven universities funded by the Engineering and Physical Sciences Research Council, National Nuclear Laboratory, Nuclear Decommissioning Authority and Sellafield Limited. The motivation for industry collaboration with universities is explained and the current progress of DISTINCTIVE in the context of these motivations is assessed. (authors)

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New High-Throughput Measurement Systems for Radioactive Wastes Segregation and Free Release - 16166

The European Metrology Research Program (EMRP) is a coordinated research-and-development program that pursues the closer integration of national research programs. The EMRP Metrology for Decommissioning Nuclear Facilities 'MetroDecom' Joint Research Project ENV54 (JRP ENV54), a 36-month project started in September 2014, delivers research products addressing all aspects of the decommissioning process. This includes the characterization of waste materials present on the decommissioning site, pre-selection prior to free release or repository acceptance measurement, free-release measurement, the measurement of waste packages thermal power prior to the repository storage, and the monitoring of stored wastes and repositories in the very long term. This paper describes the track record of two JRP ENV54 working packages, which address the measurement facilities for pre-selection of waste materials prior to measurement for repository acceptance or possible free release (segregation measurement system); and free release (free release measurement system), based on a single standardized concept characterized by unique, patented lead-free shielding. The key objective is to improve the throughput, accuracy, reliability, modularity and mobility of segregation and free-release measurement. This will result in more reliable decision-making with regard to the safe release and disposal of radioactive wastes into the environment and, resulting in positive economic outcomes. (authors)

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Development of the computer code for multi-calculations of source term of spent fuel - 15351

Many PWR spent fuels have been transferred between the neighboring nuclear power plants at Kori, Hanbit, Hanwul site, and from nuclear power plant to the KAERI(Korea Atomic Energy Research Institute) for the research purpose in Korea. In case of the PHWR type spent fuels, these have been transferred from the spent fuel pool to an on-site dry storage facility. Total radioactivity of spent fuels has to be calculated before transport in advance by the regulation requirement. In case of PWR, KN-12 transport cask can transfer 12 spent fuel assemblies at one time and in case of PHWR, Hi-STAR 63 transport cask can transfer 120 spent fuel assemblies at one time. Therefore, total radioactivity of all transported spent fuel assemblies that are transported generally are calculated by using the ORIGEN-ARP to meet this regulatory requirement. Basically, ORIGEN-ARP calculates the one fuel assembly at one time. SCALE code including this code provides the batch mode to perform the multi-calculations of the source term. In this study, in order to use the batch mode of ORIGEN-ARP, ARP-auto program was developed to generate input files and batch file of ORIGEN-ARP automatically and to make the summary file extracting the result from output files of ORIGEN-ARP. This code enhances the efficiency at the field. The working time to complete the calculation of source terms of 120 spent fuel assemblies take only 20 minutes. In order to confirm the credibility of this code, total radioactivity of 12 spent fuel assemblies of 17x17 Westinghouse type were calculated and compared to the previous result. The initial enrichments of these spent fuels are 3.46 ∼ 3.50 w/o and burnups are 29 ∼ 43 GWD/MTU. As the results, the relative errors are about 0.9 ∼ 3.0%. (authors)

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Feasibility study for detection of fuel assemblies state inside sealed dry storage casks using external gamma flux measurements - 25445

The license for the storage of spent nuclear fuel assemblies in dry casks is limited to 40 years. In Germany, a site for a final repository of spent fuel is not yet available. Therefore, this license has to be renewed for prolonged storage period. Currently, there is no experience with dry cask storage worldwide that goes beyond a period of 40 years. Beside regulatory and security aspects, there are questions concerning the long-term integrity of the spent fuel assemblies as this is of relevance for final transportation and reloading to final waste repository casks. Once the cask is filled and sealed, the knowledge about the state of the fuel assemblies is limited. This study investigates the feasibility of using external gamma flux measurements to detect the state of the spent fuel assemblies inside the cask. Monte-Carlo simulations were performed to evaluate the gamma flux distribution outside the sidewall of a cask with all intact fuel assemblies and a cask with one damaged fuel assembly, mimicking a fuel assembly with expanded rods and fuel relocation. The evaluation was performed for different location of the damaged fuel assembly and different storage times. The results of the investigation showed that the intensity and the shape of the gamma flux outside the sidewall of the cask are sufficient to identify the damaged fuel assembly, in case it is located near the inner wall of the cask. The detection of the inner damaged fuel assemblies, however, is less feasible. Due to the self-shielding of the fuel assemblies, the gamma photons which come from the inner fuel assemblies stopped within the fuel basket and assemblies materials. The gamma photons which do survive the fuel assemblies self-shielding reach the inner cask wall with not enough energy to escape the shielding of the thick cast iron wall. In general, it can be concluded that the more the damaged fuel assembly is close to the cask inner walls and that the damage is in a large scale (i.e., significant fuel relocation), the more it can be detected by external gamma flux measurements. (authors)

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Use of Multiple Remote Systems for Calcine Retrieval Operations at Idaho - 16592

There are approximately 5000 m{sup 3} of radioactive calcine material stored in six Calcine Solids Storage Facilities (CSSFs) at the DOE Idaho site. Each CSSF contains from three to twelve stainless steel tanks surrounded by a concrete vault. In accordance with the Idaho Settlement Agreement (ISA), this waste has to be retrieved and immobilized by December 31 2035 although no final decision has yet been made on how the calcine wastes are going to be retrieved from the CSSF's. One of the primary issues with the retrieval is that, in accordance with the ISA, any existing risers into the tanks have to be used 'to the maximum extent possible' in the retrieval process. Since these risers are only 6-8'' in diameter, this limits the types of systems that can be used for the retrieval. In addition, each bin containing the calcine has a series of obstructions which need to be navigated around and so inflexible systems such as conventional hose-based vacuum type systems are not feasible. In addition, the retrieval system selected has to be able to manage multiple forms of calcine (powder, granular, agglomerated, chunks etc) and operate in a very highly radioactive environment. This paper describes the progress made to date in developing an approach based on the use of multiple remote system platforms to gain entry into the bin-sets and to safely and cost-effectively retrieve the calcine wastes. Much of the work has focused on the application of the OC Robotic snake-arm robot system although significant effort has also been allocated to understanding the deployment mechanism of the snake-arm and to the ancillary supporting systems such as those for contamination control, cutting and welding new risers into the bin-sets and vacuum hose management within the bin-set. Future work will include the construction of a scaled mock up to demonstrate the field implementation of the system. (authors)

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Neutron Residual Stress Mapping for Spent Nuclear Fuel Storage Canister Weldment

Many radioactive materials within the nuclear fuel cycle present a significant hazard. Such materials include spent nuclear fuel, high-level waste, legacy waste and other nuclear materials that have no current outlet than storage. These materials are often held in long-term storage as an interim stage within their lifecycle. Lifecycles can include reuse or disposal. Safety and security of spent nuclear fuel (SNF) interim storage installations are very important, due to a great concentration of fission products, actinides and activation products. Stress corrosion cracking of interim storage containers has been identified as a high priority data gap by the Department of Energy. However, little has been done with regards to the canister weld material properties and their impact on stress corrosion, until recently from Sandia National Laboratory canister mockup program. Furthermore, because no post-weld heat treatment was required, the associated high tensile residual stress within these canister welds can drive the initiation and growth of stress corrosion cracking (SCC) cracks. In this work, we carried out in-situ neutron residual stress mapping on a weld plate sectioned from Sandia mock-up canister wall. Significant residual stress profiles were identified from the received longitudinal weld sample. The details of the weldment residual stress program development at ORNL are presented in this progress report. The generated residual stress profile from neutron diffraction technique is similar to that of Sandia residual stress data obtained from the deep hole drilling approach.

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How to increase the efficiency of the electrical discharge method for destruction of nonconductive solid materials

The paper deals with the relevance of using electrical discharge technology for construction works in the city. The technical capabilities of the installation for the multi-borehole electro-discharge splitting off and destruction of rocks and concrete are described. The ways to increase the efficiency of the electrical discharge method for the destruction of solids are proposed. In order to augment the discharge circuit energy, the energy storage is separated into two individual capacitor batteries. The throttle with the inductance of 28.6 μH is installed in one of the batteries, which increases the duration of the channel energy release to 400 μs and the efficiency of electrical discharge splitting off of concrete.

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Investigations of dual-purpose canister direct disposal feasibility - 15106

Direct disposal of commercial spent nuclear fuel (SNF) in dual-purpose canisters (DPCs) could avoid the cost and complexity of re-packaging spent fuel into smaller purpose-built containers, avoid disposing of the DPC hulls as low-level waste, and potentially decrease worker dose. Technical objectives for disposal are: 1) safety of workers and the public, 2) engineering feasibility; 3) thermal management; and 4) criticality control after permanent closure of a repository. The postclosure safety case for DPC direct disposal would resemble that for any repository-waste isolation would be enhanced by choosing host geology in which radionuclide transport is diffusion dominated, and for which transport properties and pathways are insensitive to the expected temperature history. Containment functions would be assigned to the disposal overpack, and published data show that suitable materials exist for most possible disposal environments. Earlier studies showed that DPC-based waste packages would be only slightly larger and heavier than those proposed for a repository in volcanic tuff. Engineered solutions are available for transporting and emplacing them underground, although some could be the largest of their kind. The cost of DPC direct disposal in various geologic settings has been compared to options involving a change to loading standardized multi-purpose canisters, or continuing to load DPCs and re-packaging all of them for disposal. In general DPC direct disposal would cost tens of billions of dollars less, in packaging and disposal costs. Calculations for emplacement in hard rock and sedimentary rock show that host rock peak temperature targets of 100 to 200 deg.C could be met by managing fuel age and burnup, with repository closure at up to 150 years from reactor discharge. For disposal in salt, DPC-based waste packages could be emplaced at only 50 to 100 years from discharge, depending on burnup. Sedimentary media such as shales pose special challenges especially if they have low thermal conductivity. If backfill is used around DPC-based packages its peak temperature could be 150 to 200 deg.C. Without flooding of waste packages with ground water, criticality can never occur. However, some small number of disposal overpacks could fail within the postclosure performance period due to corrosion or disruptive events. Once flooded, the aluminum-based neutron absorber materials used in most DPCs would readily degrade. Analysis has shown that even without neutron absorbers, virtually all DPCs could be subcritical if flooded with chloride brines such as would occur in a salt repository. The licensing basis for existing DPCs involves criticality analysis that may be conservative compared to fuel that is actually loaded, and the difference provides uncredited reactivity margin. Uncredited margin can offset increases in reactivity associated with flooding and basket degradation. For flooding with fresh water, with loss of neutron absorbers, many DPCs have been shown to be subcritical using uncredited margin. Significantly fewer are subcritical if the fuel basket also degrades. Thus, the materials of DPC basket construction (e.g., aluminum or carbon steel) could be deleterious if they fully corrode, along with the neutron absorbers, during the performance period. Logistical simulations are used to better understand the relationship between the needed DPC decay storage time for disposal, and the timing of future events such as the repository opening date, or a transition to loading multi-purpose (disposable) canisters at nuclear power plants. Results show that the maximum benefit from implementing a new canister design, with respect to shortening the cooling time for all SNF including that in DPCs, is associated with the earliest repository start date. As time passes more of the total SNF inventory will be in DPCs such that the value of a transition to multi-purpose canisters will decline, and the remaining options are DPC direct disposal and re-packaging in purpose-designed disposal canisters. Fuel age at emplacement is of interest to evaluate potential impacts from future changes in the fuel management system that could place fuel age limits on dry storage or transportation. The minimum fuel age at emplacement (best-case) is obtained by re-packaging all DPCs into smaller canisters for disposal, thus drastically decreasing the needed decay storage time for any geologic disposal setting. If the industry transitions to smaller, multi-purpose canisters, and without re-packaging, the fuel age at emplacement would be comparable to the best case if: 1) the emplacement power limit is high enough, or 2) both the transition and the repository start date occur early (e.g., 2036 was analyzed). Technical feasibility evaluation continues to show that direct disposal could be safe and cost effective, for some DPCs and in some geologic settings. Direct disposal of a substantial fraction of existing DPCs, limited mainly by thermal management, uncredited criticality margin, materials of DPC basket construction, and the salinity of ground water in the disposal environment, could be part of the overall waste management system. (authors)

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First Road / River Barge Security Transport of SNF under the New German Transport Security Guidelines - 19102

A transfer of 342 spent fuel assemblies from NPP Obrigheim to NPP Neckarwestheim, both located alongside the river Neckar in the south of Germany, is required to promote the decommissioning process of NPP Obrigheim, which started in 2008 after the NPP had been shut down three years before. The spent fuel storage facility in Neckarwestheim offers sufficient storage space to accept the 342 SNF assemblies, stored in fifteen dry casks type CASTOR{sup R} 440/84 mvK. After several transport feasibility studies had been conducted, Daher Nuclear Technologies GmbH was awarded the commission to plan, prepare and conduct the SNF transport. Since the combined road / river barge transport promised to be the most suitable mode of transportation Daher Nuclear Technologies started to prepare its security concept, a prerequisite to apply for and attain the transport license mandatory for transporting spent nuclear fuel in Germany. In March 2014 the company applied for the transport license, which, after more than three years and numerous amendments to the security concept, was finally granted to Daher Nuclear Technologies on May 16, 2017 by the German Federal Office for the Safety of Nuclear Waste Management (successor organization of the BfS). Simultaneously to the license application phase the transport equipment - amongst others the heavy-duty trucks and trailers, the pusher tugs, the river barge and transport frames - had to be provided and modified to meet the intensified requirements of the New German Transport Security Guidelines Road / Rail as well as several other legal regulations referring to the transport of dangerous goods by road and inland waterways (e.g. ADR - European Agreement concerning the International Carriage of Dangerous Goods by Road and ADN - European Agreement concerning the International Carriage of Dangerous Goods by Inland Waterways). The planning process involved various meetings with representatives of different authorities and administrative bodies incl. accredited experts as well as subcontractors. Tremendous training efforts had to be undertaken to train more than 100 transport participants regarding security measures, health, safety and environmental measures, as well as the proper handling of radioactive material. This also included the training of the security staff, that had to be qualified in compliance with the new Transport Security Guidelines Road and Rail. Finally, in June 2017, when all preparations where complete, all requirements met, and all permits obtained, the first ever security SNF transport on a German river was performed. Three laden CASTOR{sup R} 440/84 mvK transport and storage casks were transported in one batch via heavy-duty truck-trailer from the NPP Obrigheim to the NPP's own harbor, to be rolled on to the river barge, that would carry the three casks upstream the river Neckar to the approx. fifty kilometer distant NPP Neckarwestheim. The journey on the river took about eleven hours, in which several bridges and locks had to be passed. The entire transport was escorted by qualified security staff and auxiliary police forces on waterways and alongside on roads. It attracted nationwide public attention, due to its pioneer status and the fact, that there hadn't been any SNF or HLW shipments in Germany since 2011. After the convoy safely arrived at NPP Neckarwestheim's pier the three laden CASTOR{sup R} 440/84 mvK casks were rolled off the river barge via heavy-duty truck-trailers to be stored in the on-site interim storage facility. This first transport was a success and was followed by four additionally successful security transports in September, October, November and December 2017. The public attention receded significantly during the transport campaign, which ended on December 19, 2017 with the transfer of the last three of fifteen casks to the on-site interim storage facility in Neckarwestheim. With the last cask safely transferred to the interim storage facility Daher Nuclear Technologies successfully completed the transport campaign on schedule, meeting one of EnBW's fundamental objectives to finish the transfer of the 342 SNF assemblies by the end of 2017. (authors)

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Making the Case For Safe Storage of Used Nuclear Fuel For Extended Periods of Time: Combining Near-Term Experiments and Analyses with Longer-Term Confirmatory Demonstrations

The need for extended storage of used nuclear fuel is increasing globally as disposition schedules for used fuel are pushed further into the future. This is creating a situation where dry storage of used fuel may need to be extended beyond normal regulatory licensing periods. While it is generally accepted that used fuel in dry storage will remain in a safe condition, there is little data that demonstrate used fuel performance in dry storage environments for long periods of time. This is especially true for high burnup used fuel.

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An Experimental Feasibility Study on Applying Neutron Tomography to Encapsulated Spent Nuclear Fuel - 20050

Visual inspection makes easier to ensure the integrity and safety of spent nuclear fuel (SNF) than any classical techniques. Various classical techniques have been applied but there are no reliable methods to qualitatively and quantitatively verify spent fuel in dry storage. Thus, the present authors have developed the prototype safeguards apparatus for dry storage employing the array of He-4 gas scintillation detectors (S670E, Arktis Radiation Detectors Ltd., Switzerland), newly designed to simultaneously measure thermal and fast neutrons without any moderators. The S670E detector has a cylindrical shape with a diameter of 52 mm and active length of 600 mm (total length: 875 mm). The detector is filled by He-4 gas with an approximate pressure of 180 bar for fast neutron detection, and its inner wall is coated by Li-6 for thermal neutron detection. The scintillation lights generated via Li-6 nuclear reaction and elastic scattering are collected by 24 SiPMs linearly paired at the center of the detector. The detector delivers a TTL (Transistor-Transistor Logic) output for pulse readout and UART (Universal Asynchronous Receiver Transmitter) for device control. In order to assess feasibility of the apparatus, an experimental system has been designed, built, and optimized via computational studies. Cf-252 neutron sources and linearly arrayed detectors, working as a single detector, were occupied for this study due to the difficulties in working with SNF. The laboratory scale cask (diameter: 0.67 m, height: 1.5 m), minimized by a factor of 10 compared to the actual thickness of a commercial TN-32 cask, was also manufactured. The detector array was designed to rotate the lab-scale cask and obtain 36 image profiles at every 10 degrees. All profiles were aligned in single frame image called a sinogram, and the cross-sectional image was then fabricated by the inverse radon transform algorithm. These experiments have been repeated with different configurations and numbers of sources. Some gamma-ray sources were also measured with neutron sources in order to distinguish between neutron and gamma-ray pulses. Basically, a He-4 detector is designed to run on Linux OS so it is difficult to directly apply to Windows-based equipment widely used in S. Korea. Therefore, a new data acquisition board working on Windows OS was designed and built. The board mainly consists of FPGA (Field Programmable Gate Array) and SoC (System on Chip) for TTL pulse readout, sorting measured data, and transferring data to a user interface. In conclusion, the tomographic system at lab-scale has shown considerable potential to detect a partial or gross defect of encapsulated assemblies in dry storage. Next steps of this study will be to 1) repeatedly carry out experiments to demonstrate scientific reliability and validity, and 2) numerically integrate signals with weight factors to enhance the image quality since the suggested system based on passive interrogation method requires longer measurement time. Finally, the system will apply to a commercial dry storage phased out soon in S. Korea. (authors)

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Resource Conservation and Recovery Act General Contingency Plan for Hazardous Waste Treatment, Storage, and Disposal Units at the Y-12 National Security Complex. Revision 12

This document, Y/TS-1015/R12, Resource Conservation and Recovery Act General Contingency Plan for Hazardous Waste Treatment, Storage, and Disposal Units at the Y-12 National Security Complex, is for reference purposes. It is a consolidation of the August 10, 2018, versions of Attachment 6, “Contingency Plan,” from Tennessee Hazardous Waste Management Permits TNHW-122 and TNHW-127. It is not intended to supersede any requirements in these permits and any differences are unintentional. The decision to implement this contingency plan, as part of the Emergency Plan, depends upon whether an imminent or actual incident involving a RCRA treatment, storage, or disposal unit could threaten human health or the environment.

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Blue Ribbon Commission, Yucca Mountain Closure, Court Actions - Future of Decommissioned Reactors, Operating Reactors and Nuclear Power - 13249

Issues related to back-end of the nuclear fuel cycle continue to be difficult for the commercial nuclear power industry and for the decision makers at the national and international level. In the US, the 1982 NWPA required DOE to develop geological repositories for SNF and HLW but in spite of extensive site characterization efforts and over ten billion dollars spent, a repository opening is nowhere in sight. There has been constant litigation against the DOE by the nuclear utilities for breach of the 'standard contract' they signed with the DOE under the NWPA. The SNF inventory continues to rise both in the US and globally and the nuclear industry has turned to dry storage facilities at reactor locations. In US, the Blue Ribbon Commission on America's Nuclear Future issued its report in January 2012 and among other items, it recommends a new, consent-based approach to siting of facilities, prompt efforts to develop one or more geologic disposal facilities, and prompt efforts to develop one or more consolidated storage facilities. In addition, the March 2011 Fukushima Daiichi accident had a severe impact on the future growth of nuclear power. The nuclear industry is focusing on mitigation strategies for beyond design basis events and in the US, the industry is in the process of implementing the recommendations from NRC's Near Term Task Force. (authors)

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