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Preliminary Evaluation of Removing Used Nuclear Fuel From Nine Shutdown Sites

The Blue Ribbon Commission on America’s Nuclear Future identified removal of stranded used nuclear fuel at shutdown sites as a priority so that these sites may be completely decommissioned and put to other beneficial uses. In this report, a preliminary evaluation of removing used nuclear fuel from nine shutdown sites was conducted. The shutdown sites included Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion. At these sites a total of 7649 used nuclear fuel assemblies and a total of 2813.2 metric tons heavy metal (MTHM) of used nuclear fuel are contained in 248 storage canisters. In addition, 11 canisters containing greater-than-Class C (GTCC) low-level radioactive waste are stored at these sites. The evaluation was divided in four components: • characterization of the used nuclear fuel and GTCC low-level radioactive waste inventory at the shutdown sites • an evaluation of the onsite transportation conditions at the shutdown sites • an evaluation of the near-site transportation infrastructure and experience relevant to the shipping of transportation casks containing used nuclear fuel from the shutdown sites • an evaluation of the actions necessary to prepare for and remove used nuclear fuel and GTCC low-level radioactive waste from the shutdown sites. Using these evaluations the authors developed time sequences of activities and time durations for removing the used nuclear fuel and GTCC low-level radioactive waste from a single shutdown site, from three shutdown sites located close to each other, and from all nine shutdown sites.

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Savannah River Site Eastern Transportation Hub: A Concept For a DOE Eastern Packaging, Staging and Maintenance Center - 13143

The Department of Energy (DOE) is working to de-inventory sites and consolidate hazardous materials for processing and disposal. The DOE administers a wide range of certified shipping packages for the transport of hazardous materials to include Special Nuclear Material (SNM), radioactive materials, sealed sources and radioactive wastes. A critical element to successful and safe transportation of these materials is the availability of certified shipping packages. There are over seven thousand certified packagings (i.e., Type B/Type AF) utilized within the DOE for current missions. The synergistic effects of consolidated maintenance, refurbishment, testing, certification, and costing of these services would allow for efficient management of the packagings inventory and to support anticipated future in-commerce shipping needs. The Savannah River Site (SRS) receives and ships radioactive materials (including SNM) and waste on a regular basis for critical missions such as consolidated storage, stabilization, purification, or disposition using H-Canyon and HB-Line. The Savannah River National Laboratory (SRNL) has the technical capability and equipment for all aspects of packaging management. SRS has the only active material processing facility in the DOE complex and is one of the sites of choice for nuclear material consolidation. SRS is a logical location to perform maintenance and periodic testing of the DOE fleet of certified packagings. This initiative envisions a DOE Eastern Packaging Staging and Maintenance Center (PSMC) at the SRS and a western hub at the Nevada National Security Site (NNSS), an active DOE Regional Disposal Site. The PSMC's would be the first place DOE would go to meet their radioactive packaging needs and the primary locations projects would go to disposition excess packaging for beneficial reuse. These two hubs would provide the centralized management of a packaging fleet rather than the current approach to design, procure, maintain and dispose of packagings on a project-by-project basis. This initiative provides significant savings in packaging costs and acceleration of project schedules. In addition to certified packaging, the PSMC would be well suited for select designs of 7A Type A packaging and Industrial Packaging. (authors)

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Thermal behavior of polyhalite: a high-temperature synchrotron XRD study

As an accessory mineral in marine evaporites, polyhalite, K 2 MgCa 2 (SO 4 ) 4 ·2H 2 O, coexists with halite (NaCl) in salt formations, which have been considered as potential repositories for permanent storage of high-level nuclear wastes. However, because of the heat generated by radioactive decays in the wastes, polyhalite may dehydrate, and the released water will dissolve its neighboring salt, potentially affecting the repository integrity. Thus, studying the thermal behavior of polyhalite is important. In this paper, a polyhalite sample containing a small amount of halite was collected from the Salado formation at the WIPP site in Carlsbad, New Mexico. To determine its thermal behavior, in situ high-temperature synchrotron X-ray diffraction was conducted from room temperature to 1066 K with the sample powders sealed in a silica-glass capillary. At about 506 K, polyhalite started to decompose into water vapor, anhydrite (CaSO 4 ) and two langbeinite-type phases, K 2 Ca x Mg 2-x (SO 4 ) 3 , with different Ca/Mg ratios. XRD peaks of the minor halite disappeared, presumably due to its dissolution by water vapor. With further increasing temperature, the two langbeinite solid solution phases displayed complex variations in crystallinity, composition and their molar ratio and then were combined into the single-phase triple salt, K 2 CaMg(SO 4 ) 3 , at ~919 K. Rietveld analyses of the XRD data allowed determination of structural parameters of polyhalite and its decomposed anhydrite and langbeinite phases as a function of temperature. Finally, from the results, the thermal expansion coefficients of these phases have been derived, and the structural mechanisms of their thermal behavior been discussed.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Radiolysis Model Formulation for Integration with the Mixed Potential Model

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel (UNF) and high-level radioactive waste. Within the UFDC, the components for a general system model of the degradation and subsequent transport of UNF is being developed to analyze the performance of disposal options [Sassani et al., 2012]. Two model components of the near-field part of the problem are the ANL Mixed Potential Model and the PNNL Radiolysis Model. This report is in response to the desire to integrate the two models as outlined in [Buck, E.C, J.L. Jerden, W.L. Ebert, R.S. Wittman, (2013) “Coupling the Mixed Potential and Radiolysis Models for Used Fuel Degradation,” FCRD-UFD-2013-000290, M3FT-PN0806058]

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MARVEL Decommissioning Plan

The purpose of this document is to update and formally document decommissioning plans for the Microreactor Applications Research Validation and Evaluation (MARVEL) microreactor after two years of operational testing. The operational testing consists of 3 continuous days of operation followed by 4 days of non-operation. This plan is also based on the assumption that all MARVEL fuel will be transported to long-term intermediate storage at the CPP-603 used fuel storage facility, sodium-potassium (NaK) coolant can be disposed as Class B radioactive waste and that the balance of the reactor primary components can be disposed of at an available disposal facility.

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Status and Path Forward for the Department of Energy Used Fuel Disposition Storage and Transportation Program - 12571

The U.S. Department of Energy, Office of Nuclear Energy (DOE/NE) has sponsored a program since Fiscal Year (FY) 09 to develop the technical basis for extended dry storage of used fuel. This program is also working to develop the transportation technical basis for the transport of used fuel after the extended storage period. As this program has progressed, data gaps associated with dry storage systems (e.g., fuel, cask internals, canister, closure, overpack, and pad) have been identified that need to be addressed to develop the technical bases for extended storage and transportation. There has also been an initiation of experimental testing and analyses based on the identified data gaps. The technical aspects of the NE program are being conducted by a multi-lab team made up of the DOE laboratories. As part of this program, a mission objective is to also collaborate closely with industry and the international sector to ensure that all the technical issues are addressed and those programs outside the DOE program can be leveraged, where possible, to maximize the global effort in storage and transportation research. The DOE/NE program is actively pursuing the development of the technical basis to demonstrate the feasibility of storing UNF for extended periods of time with subsequent transportation of the UNF to its final disposition. This program is fully integrated with industry, the U.S. regulator, and the international community to assure that programmatic goals and objectives are consistent with a broad perspective of technical and regulatory opinion. As the work evolves, assessments will be made to ensure that the work continues to focus on the overall goals and objectives of the program. (authors)

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Leaning the tank closure process at the Savannah River Site - 15515

The Savannah River Site liquid waste system consists of multiple facilities that safely receive and store legacy radioactive waste prior to treatment and disposition. The liquid radioactive waste is stored in large underground tanks. The waste inventory currently stored in these tanks is a complex mixture of insoluble metal hydroxide solids and soluble salt supernate. The high-level (radioactive) portion of the waste is vitrified into a glass waste form, while the low-level waste is immobilized into a cementitious grout waste form called salt-stone. Once the waste is retrieved and processed, the tanks are closed. The closure process includes; waste removal, characterizing residuals, isolating the tank systems, and stabilizing the tank with tailored grout formulations. The core of Lean is to eliminate waste or non-value added steps from processes to maximize customer value. From the perspective of the customer, 'value' is defined as any action or process for which a customer would willingly pay. In this case our product is to safely and efficiently remove waste from the underground storage tanks and demonstrate operational closure to the State of South Carolina. Savannah River Remediation (SRR) initiated the use of the Lean process in late 2013 and identified the Tank Closure process as one of the targeted areas to improve. The first step in the Lean process is to 'map' out the value stream and identify areas where non-value added steps are present and then prioritize Rapid Improvement Events (RIE). Based on this mapping process, there were a number of areas that SRR determined warranted further evaluation for improvements, including; standard equipment design, characterization, standard work packages, regulatory documentation, tank characterization, structural evaluations, engineering documentation and equipment re-use. This paper focuses on the application of the Lean process at SRR and more specifically, effective application to the Tank Closure Process. This paper discusses the value stream mapping process for Tank Closures, as well as, summarizes the RIEs and demonstrate the realized benefits. Benefits can come in a variety of forms, from schedule compression to decreased costs. Through the Lean process we have been able to identify close to $10 M in savings over the life cycle of the tank closure process and a 4 year schedule compression. Cost and schedule savings are used to accomplish further risk reduction efforts. In addition to hard benefits, there are also soft benefits. These soft benefits are a result of the process itself and help drive improved efficiencies through better communication with key stakeholders, such as regulators and customers. (authors)

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Challenges in Interim and Final Storage of HLW in Germany - 19060

After the Fukushima event in 2011 Germany has decided to phase out nuclear power. Subsequently, it seemed to be politically attractive also to achieve a broad consensus with respect to final disposal of the high-active heat-generating waste after a long societal conflict lasting for decades. Germany has then decided to start a new site selection program for the determination of a site for high-active and heat-generating waste. In 2013, an overall parliamentary consensus could be achieved to agree on a new Repository Site Selection Act. This law has then been evaluated by a commission and the Repository Site Selection Act became amended. Besides, also organization and responsibilities became adapted as well as the financing system. A fund has been installed, which was financed by the utilities, to secure the availability of money to finance the whole process. In the meanwhile, the organizations for licensing of interim storage and final disposal facilities (BfE), for the operation of interim storage facilities for spent fuel, high-active waste and low and intermediate level waste (BGZ) and for site investigation, construction, operation and decommissioning of repositories for spent fuel and high-active waste as well as for low and intermediate level waste (BGE) had been founded and the site investigation process has just started. (authors)

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Indonesian Highly Enriched Uranium-Bearing Waste Elimination - 17393

The Department of Energy's National Nuclear Security Administration (NNSA) and the Y-12 National Security Complex (Y-12) have been working with PT Industri Nuklir Indonesia (PT Inuki), an Indonesian medical isotope producer, since 2014 to disposition all highly enriched uranium (HEU) stored at the government-owned Research Center for Science and Technology (PUSPIPTEK) near Serpong, Indonesia. To eliminate the 1.5 kilograms of liquid HEU at the facility, Y-12 proposed a method to down-blend the leftover process solutions with a depleted uranium solution made on-site to render low-enriched uranium (LEU). This method is essentially irreversible since the uranium enrichment is degraded using a molecular level of intimate mixing in aqueous solution. This cooperation was tied to a presidential commitment made at the 2016 Nuclear Security Summit to down-blend all remaining HEU in Indonesia by September 2016. While all unirradiated HEU was down-blended to LEU in March 2016, the campaign to down-blend the 1.4 kilograms of irradiated HEU began in late July 2016 and was completed in August, making Indonesia and all of Southeast Asia free of HEU. As part of this cooperation, Y-12 and NNSA also proposed a subsequent disposition option for the LEU solution whereby the irradiated and unirradiated solutions will be chemically manipulated to prepare a dry LEU oxide suitable for interim storage and yield a uranium-free liquid stream that can be discarded as radioactive waste. In the fall of 2016, Y-12 and NNSA plan to begin developing the LEU solution disposition methodology with the Indonesian facility using small-scale batch-wise methods within existing hot cells, involving straightforward precipitation, ion exchange, decantation, and calcination methods. (authors)

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Skyshine Calculations for a Large Spent Nuclear Fuel Storage Facility with SCALE 6.2.3

The SCALE code system developed at Oak Ridge National Laboratory includes state-of-the-art capabilities for radiation source term and radiation transport simulations that can be used in numerous applications, including dose rate analyses of complex consolidated interim storage facilities (CISFs). A licensed CISF could be used to store tens of thousands of tonnes of spent nuclear fuel discharged from commercial power reactors using various cask and storage pad designs. A CISF design must comply with the regulatory requirements provided in 10 CFR Part 72, including requirements related to annual dose limits applicable to real individuals located beyond the area controlled by the licensee. Therefore, calculating a dose to the public is a necessary part of the licensing process for the construction of a CISF. These calculations are very challenging because of the complexity of the CISF design and the low magnitude of dose rate at large distances from the facility. This paper describes detailed far-field dose rate calculations performed for a proposed CISF using MAVRIC, the Monte Carlo radiation shielding sequence in SCALE 6.2.3, with automated variance reduction based on discrete ordinates calculations. The method presented in this paper uses a detailed Monte Carlo radiation transport simulation in one step from source to dose rate. A series of independent simulations was made using the complete site geometry (all casks present), but with only one cask containing radiation sources to obtain the dose rate maps produced by each storage cask. The CISF dose rate map was obtained by adding the dose rate maps produced by the independent individual cask simulations. Ample volumes of air and soil extending beyond the location of interest for dose rate calculation were included in the calculation model to properly simulate important radiation attenuation and scattering events that affect far-field dose rates. Furthermore, a comprehensive sensitivity study is included in this paper to illustrate the importance of selecting appropriate air volume, mass density, and composition for CISF skyshine dose rate calculations. Dry soil and soil containing water were analyzed to determine their effects on groundshine radiation.

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Multinational underground nuclear parks

Newcomer countries expected to develop new nuclear power programs by 2030 are being encouraged by the International Atomic Energy Agency to explore the use of shared facilities for spent fuel storage and geologic disposal. Multinational underground nuclear parks (M-UNPs) are an option for sharing such facilities. Newcomer countries with suitable bedrock conditions could volunteer to host M-UNPs. M-UNPs would include back-end fuel cycle facilities, in open or closed fuel cycle configurations, with sufficient capacity to enable M-UNP host countries to provide for-fee waste management services to partner countries, and to manage waste from the M-UNP power reactors. M-UNP potential advantages include: the option for decades of spent fuel storage; fuel-cycle policy flexibility; increased proliferation resistance; high margin of physical security against attack; and high margin of containment capability in the event of beyond-design-basis accidents, thereby reducing the risk of Fukushima-like radiological contamination of surface lands. A hypothetical M-UNP in crystalline rock with facilities for small modular reactors, spent fuel storage, reprocessing, and geologic disposal is described using a room-and-pillar reference-design cavern. Underground construction cost is judged tractable through use of modern excavation technology and careful site selection. (authors)

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Crystalline Silicotitanate ion exchange media long-term storage evaluation

Crystalline Silicotitanate a (CST) ion exchange media is produced in an engineered form, known as IONSIV™ IE-911. This material has been tested extensively for use in removing 137 Cs from Savannah River Site (SRS) and Hanford site tank waste. This inorganic material is highly selective for removing cesium ions, even in the presence of high concentrations of sodium and potassium ions and is tolerant of the harsh conditions of the varying tank chemistries. A batch of caustic washed IE-911 that had been produced by the supplier and received at SRS in 2001 was tested to verify that it was still in good condition after the extended storage. This testing involved standard batch contacts to measure the removal of Cs from simulated tank waste. This report describes the results of testing performed to verify that the IE-911 matches the performance of previous batches of this media. To confirm that the performance of IE-911 is similar to previous batches and that no degradation had occurred during storage, batch contact testing was performed using SRNL’s standard methodology using a non-radioactive SRS Average Waste Simulant composition utilized in previous studies. Batch contact kinetics testing was also conducted to determine cesium loading on this batch versus time. The cesium equilibrium loading isotherm in SRS simulant at 23°C was compared to ZAM isotherm model predictions and previous test results with IE-911 media. The measured and predicted distribution coefficients and cesium loading (“Q”) are shown in Table ES-1. The physical properties of the material were also determined, including density, moisture content, and particle size.

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Advanced Computational Modeling of High-Level Waste Vitrification at the Hanford Site

The U.S. Department of Energy (DOE) has selected vitrification for stabilizing legacy tank waste at the Hanford site, where radioactive waste from plutonium production was historically stored in underground tanks. This waste will be separated into low-activity waste (LAW) and high-level waste (HLW) fractions and processed at the Waste Treatment and Immobilization Plant (WTP). At WTP, glass melters are used for the vitrification of radioactive tank waste, transforming it into a stable borosilicate glass form for safe long-term storage. The melter vessel is constructed from highly durable and heat-resistant materials, where the vitrification process occurs. The main regions that are modeled are the melt pool, plenum, cold cap, riser/discharge chamber, and surrounding structure with insulation layers. Forced convection induced by air bubblers at the base of the melter ensure uniform temperature distribution and provide heat to the cold cap layer. The cold cap is a region of reacting batch feed that floats on top of the molten glass and is where the batch-to-glass reactions occur. Joule heating provided by electrodes mounted along the vertical walls of the melter and immersed directly in the glass, generates the necessary heat for the net endothermic conversion processes that occur in the cold cap. The high temperatures, radioactivity, and opaque nature of the glass prevent direct observation inside the melters. Therefore, computational models are essential for providing insight into factors that affect melter throughput. Thermocouples in the plenum provide operators with plenum temperature measurements. Operational adjustments include bubbling rate, voltage supplied to the electrodes, feed adjustments, and glass removal rate. Different computational fluid dynamics (CFD) models have been developed, each serving a specific purpose. There are CFD models of different scale melters, as well as models that capture the two-phase flow interfaces of rising bubbles in the molten glass or models with a simplified molten glass region so that the surrounding structure and plenum can be feasibly incorporated. Pilot-scale melter models have been developed to serve as validation of the methods employed in the simulation of the full-scale WTP melters. Models incorporating resolved bubbling are used to develop momentum source terms to implement into a single phase, multi-region, steady-state flow model that is being validated by measured process parameters such as glass production rate, voltage, input power, plenum temperatures, etc. The resolved bubbling model uses the multiphase volume of fluid approach to model the system with a high-resolution interface capturing scheme to maintain sharp interfaces between the molten glass and the air phase. The suite of CFD models is continually being improved to incorporate more realistic physics and achieve faster turnaround time. For example, an incremental controller is implemented to automatically adjust electrode voltage within the simulation to a molten glass set point temperature of 1150°C. Newer models feature improved meshes to ensure conformal meshes between regions and eliminate unnecessary mesh refinement in areas that are not of interest (such as boundary layers in offgas ports). Instead of explicitly modeling the structural, refractory, and insulation layers of the melter, a thermal resistance approach is used with published correlations used for boundary conditions. The development of robust and efficient CFD models will be instrumental in enabling the WTP to successfully fulfill its mission of safely stabilizing legacy nuclear waste.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗

NRC Technical Research Program to Evaluate Extended Storage and Transportation of Spent Nuclear Fuel - 12547

Any new direction proposed for the back-end of spent nuclear fuel (SNF) cycle will require storage of SNF beyond the current licensing periods. The Nuclear Regulatory Commission (NRC) has established a technical research program to determine if any changes in the 10 CFR part 71, and 72 requirements, and associated guidance might be necessary to regulate the safety of anticipated extended storage, and subsequent transport of SNF. This three part program of: 1) analysis of knowledge gaps in the potential degradation of materials, 2) short-term research and modeling, and 3) long-term demonstration of systems, will allow the NRC to make informed regulatory changes, and determine when and if additional monitoring and inspection of the systems is necessary. The NRC has started a research program to obtain data necessary to determine if the current regulatory guidance is sufficient if interim dry storage has to be extended beyond the currently approved licensing periods. The three-phased approach consists of: - the identification and prioritization of potential degradation of the components related to the safe operation of a dry cask storage system, - short-term research to determine if the initial analysis was correct, and - a long-term prototypic demonstration project to confirm the models and results obtained in the short-term research. The gap analysis has identified issues with the SCC of the stainless steel canisters, and SNF behavior. Issues impacting the SNF and canister internal performance such as high and low temperature distributions, and drying have also been identified. Research to evaluate these issues is underway. Evaluations have been conducted to determine the relative values that various types of long-term demonstration projects might provide. These projects or follow-on work is expected to continue over the next five years. (authors)

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International Low Level Waste Disposal Practices and Facilities

The safe management of nuclear waste arising from nuclear activities is an issue of great importance for the protection of human health and the environment now and in the future. The primary goal of this report is to identify the current situation and practices being utilized across the globe to manage and store low and intermediate level radioactive waste. The countries included in this report were selected based on their nuclear power capabilities and involvement in the nuclear fuel cycle. This report highlights the nuclear waste management laws and regulations, current disposal practices, and future plans for facilities of the selected international nuclear countries. For each country presented, background information and the history of nuclear facilities are also summarized to frame the country's nuclear activities and set stage for the management practices employed. The production of nuclear energy, including all the steps in the nuclear fuel cycle, results in the generation of radioactive waste. However, radioactive waste may also be generated by other activities such as medical, laboratory, research institution, or industrial use of radioisotopes and sealed radiation sources, defense and weapons programs, and processing (mostly large scale) of mineral ores or other materials containing naturally occurring radionuclides. Radioactive waste also arises from intervention activities, which are necessary after accidents or to remediate areas affected by past practices. The radioactive waste generated arises in a wide range of physical, chemical, and radiological forms. It may be solid, liquid, or gaseous. Levels of activity concentration can vary from extremely high, such as levels associated with spent fuel and residues from fuel reprocessing, to very low, for instance those associated with radioisotope applications. Equally broad is the spectrum of half-lives of the radionuclides contained in the waste. These differences result in an equally wide variety of options for the management of radioactive waste. There is a variety of alternatives for processing waste and for short term or long term storage prior to disposal. Likewise, there are various alternatives currently in use across the globe for the safe disposal of waste, ranging from near surface to geological disposal, depending on the specific classification of the waste. At present, there appears to be a clear and unequivocal understanding that each country is ethically and legally responsible for its own wastes, in accordance with the provisions of the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management. Therefore the default position is that all nuclear wastes will be disposed of in each of the 40 or so countries concerned with nuclear power generation or part of the fuel cycle. To illustrate the global distribution of radioactive waste now and in the near future, Table 1 provides the regional breakdown, based on the UN classification of the world in regions illustrated in Figure 1, of nuclear power reactors in operation and under construction worldwide. In summary, 31 countries operate 433 plants, with a total capacity of more than 365 gigawatts of electrical energy (GW[e]). A further 65 units, totaling nearly 63 GW(e), are under construction across 15 of these nations. In addition, 65 countries are expressing new interest in, considering, or actively planning for nuclear power to help address growing energy demands to fuel economic growth and development, climate change concerns, and volatile fossil fuel prices. Of these 65 new countries, 21 are in Asia and the Pacific region, 21 are from the Africa region, 12 are in Europe (mostly Eastern Europe), and 11 in Central and South America. However, 31 of these 65 are not currently planning to build reactors, and 17 of those 31 have grids of less than 5 GW, which is said to be too small to accommodate most of the reactor designs available. For the remaining 34 countries actively planning reactors, as of September 2010: 14 indicate a strong intention to precede with introduction of nuclear power; 7 are preparing but haven't made a final decision, 10 have made a decision and are preparing infrastructure, 2 have ordered a new nuclear power plant, and 1 has a plant under construction. In all countries interested in pursuing nuclear power, it is necessary for the governments to create an environment for investment and advancement of nuclear power, including development of a professional and independent regulatory framework and regime, knowledge and refinement of skills in nuclear safety and control, definition of policies on nuclear waste management and decommissioning, and participation in international non-proliferation measures. Specifically related to radioactive waste management, nuclear facilities and industries that utilize radioactive material work to well-established safety standards for the management of their waste.

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Probability of Pinching Failure in PWR SNF after Long-Term Storage

A probabilistic assessment was performed on pinching failure of PWR SNF rod cladding under accidental SNF cask transport conditions. a set of 3,000 rod pinching scenarios was analyzed, with different geometrical, material, and loading parameters, using a Latin Hypercube Sampling (LHS). The investigated accidents include train derailments leading to cask collisions or cask drops. Pinching failure was related to cladding offset strain, defined as the plastic deformation interval in radial cladding direction normalized with respect to the outer cladding diameter. Finite element (FE) models were used to compute SNF rod deformations and the offset strain demand OSD in each of the assessed pinching scenarios. The numerical models account for the contribution of the degraded fuel pellets to rod pinching stiffness. The offset strain capacity OSC was computed using regression and interpolation models based on ring compression (RCT) test results, and account for hydride-related effects on cladding ductility. The results show a low probability of pinching failure. Hydride-related cladding embrittlement affects the structural integrity of PWR SNF rods after long-term dry storage of up to 300 years. Radial hydride patterns or large amounts of circumferentially oriented hydrides reduce the cladding ductile deformation capability when pinched at low material temperatures. The probability of hydride-related pinching failure of PWR cladding is estimated after long-term dry storage to address: How do PWR SNF rod conditions like hydrogen content, vacuum drying peak cladding hoop stress σ{sub θ}, and cladding material temperature affect cladding ductility? Is hydride-related cladding embrittlement a material degradation mechanism that affects the structural integrity of PWR SNF rods under pinching loading? What are the controlling parameters for deformation demands in PWR SNF rods under pinching loading? Is the pinching load amplitude an important parameter that controls the probability of pinching failure? How do parameters like pellet-cladding gap or fuel pellet degradation affect the pinching loading capacity of PWR SNF rods? Statistical modelling of PWR SNF rod conditions in U.S. inventory: A pseudo rod database was generated based on EIA database. Analysis of train accident scenarios: a probability density function (pdf) of a pinching load distribution P{sub 9m} of a 9-m cask drop is linearly scaled to generate a more representative: distribution P{sub AS}. The new distribution considers train collision velocities, cask drops for on- and off-bridge accidents, and cask impact orientation. LHS generates 3,000 rod pinching scenarios based on nine independent geometrical, material, and loading parameters. Probabilistic FE Analysis: The LHS realizations are used to generate 3,000 FE models. The FE analysis is conducted for P{sub 9m}, and P{sub AS}. The workflow is automatized. The OSC is computed with regression and interpolation models, as a function of the hydrogen content C{sub H}, vacuum drying peak σ{sub θ}, and SNF temperature T{sub FR}: The regression data was obtained in RCTs on PWR cladding. The OSD is derived from the cladding deformations in the FE models. Cladding failure was identified via a comparison of OSC and OSD. A 300 and a 30 years-of-dry-storage duration is investigated: The dry-storage period affects the expected SNF temperature and consequently, the OSC and the OSD. The material temperature LHS population is adapted to account for a SNF temperature variation. The probability of hydride-related, ductile cladding failure during a severe transportation accident is not significantly affected by the pinching load amplitude: The probability is slightly reduced from a range of less than 0.0333% to 0.1667% for P{sub 9m}, to a range of less than 0.0333% to 0.1333% for P{sub AS}, depending on the vacuum drying peak cladding hoop stress modelling assumptions. Further, the SNF dry-storage duration does not significantly affect the probability of ductile pinching failure. Based on available train accidents data, the pinching failure probability for a 5,000 km long transportation route ranges from less than 1.62 E-7 to a value of 6.47 E-7, depending on the vacuum drying peak cladding hoop stress modelling assumptions. The pinching load amplitude is not a controlling parameter for the probability of ductile cladding pinching failure, and does not control the cladding deformations once pellet-cladding contact is established, The pellet-cladding gap width, however, plays a key role. Fuel pellet degradation only affects the deformation capacity, if the pellet stiffness is significantly reduced (more than 90%). Only highly embrittled cladding is likely to fail. The deformation demands on the cladding under pinching loading are small.

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Developing Tomorrows Nuclear Professionals - Apprentice and Graduate Development at PFCS, Sellafield - 19013

The Pile Fuel Cladding Silo (PFCS) at Sellafield, originally built in 1949/50 to receive and store fuel cladding from reprocessing Windscale Pile fuel, has been used to store this material and other waste from the UK nuclear programme ever since. However, the Silo is unsuitable for indefinite storage of the waste; its age, condition and inventory, combined with significant uncertainty in all these factors, rank the facility among the most hazardous in Western Europe. In 2012, Sellafield Ltd began a highly collaborative relationship with its supply chain partners Bechtel Cavendish Nuclear Solutions (BCNS), to undertake a suite of projects to create a capability for retrieving waste. In doing so, the team has discovered a shared passion for graduate student and apprenticed trades development, which has led to the team goal of becoming the number one destination for training secondments within the associated businesses. This paper describes how the team has leveraged the unique challenge of the project, the extensive experience of the delivery team, and the capabilities of the contributing organisations to create a structure secondment experience aimed at maximizing learning and generating enthusiasm for nuclear clean-up. The scheme was developed and implemented, based on some very simple but consistent feedback from the trainee community on what differentiates a good and bad secondment. The paper includes three case studies from individuals who have participated in the programme and the analysis of overall feedback from all programme trainees. Universally, the scheme has been well received with a significant number of trainees coming back to the project to begin their professional careers. The simple differentiators between a good and bad secondment provide a good basis for those wishing to develop similar schemes. (authors)

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Crystalline Ceramic Waste Forms: Reference Formulation Report

The research conducted in this work package is aimed at taking advantage of the long term thermodynamic stability of crystalline ceramics to create more durable waste forms (as compared to high level waste glass) in order to reduce the reliance on engineered and natural barrier systems. Durable ceramic waste forms that incorporate a wide range of radionuclides have the potential to broaden the available disposal options and to lower the storage and disposal costs associated with advanced fuel cycles. Assemblages of several titanate phases have been successfully demonstrated to incorporate radioactive waste elements, and the multiphase nature of these materials allows them to accommodate variation in the waste composition. Recent work has shown that they can be successfully produced from a melting and crystallization process. The objective of this report is to explain the design of ceramic host systems culminating in a reference ceramic formulation for use in subsequent studies on process optimization and melt property data assessment in support of FY13 melter demonstration testing. The waste stream used as the basis for the development and testing is a combination of the projected Cs/Sr separated stream, the Trivalent Actinide - Lanthanide Separation by Phosphorous reagent Extraction from Aqueous Komplexes (TALSPEAK) waste stream consisting of lanthanide fission products, the transition metal fission product waste stream resulting from the transuranic extraction (TRUEX) process, and a high molybdenum concentration with relatively low noble metal concentrations. In addition to the combined CS/LN/TM High Mo waste stream, variants without Mo and without Mo and Zr were also evaluated. Based on the results of fabricating and characterizing several simulated ceramic waste forms, two reference ceramic waste form compositions are recommended in this report. The first composition targets the CS/LN/TM combined waste stream with and without Mo. The second composition targets with CS/LN/TM combined waste stream with Mo and Zr removed. Waste streams that contain Mo must be produced in reducing environments to avoid Cs-Mo oxide phase formation. Waste streams without Mo have the ability to be melt processed in air. A path forward for further optimizing the processing steps needed to form the targeted phase assemblages is outlined in this report. Processing modifications including melting in a reducing atmosphere, and controlled heat treatment schedules are anticipated to improve the targeted elemental partitioning.

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