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At least 613 records · Page 34

Machine learning surrogates for surface complexation model of uranium sorption to oxides

Abstract The safety assessments of the geological storage of spent nuclear fuel require understanding the underground radionuclide mobility in case of a leakage from multi-barrier canisters. Uranium, the most common radionuclide in non-reprocessed spent nuclear fuels, is immobile in reduced form (U(IV) and highly mobile in an oxidized state (U(VI)). The latter form is considered one of the most dangerous environmental threats in the safety assessments of spent nuclear fuel repositories. The sorption of uranium to mineral surfaces surrounding the repository limits their mobility. We quantify uranium sorption using surface complexation models (SCMs). Unfortunately, numerical SCM solvers often encounter convergence problems due to the complex nature of convoluted equations and correlations between model parameters. This study explored two machine learning surrogates for the 2-pK Triple Layer Model of uranium retention by oxide surfaces if released as U(IV) in the oxidizing conditions: random forest regressor and deep neural networks. Our surrogate models, particularly DNN, accurately reproduce SCM model predictions at a fraction of the computational cost without any convergence issues. The safety assessment of spent fuel repositories, specifically the migration of leaked radioactive waste, will benefit from having ultrafast AI/ML surrogates for the computationally expensive sorption models that can be easily incorporated into larger-scale contaminant migration models. One such model is presented here.

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Reduced diffusion and enhanced retention of multiple radionuclides from pore structure characterization of barrier materials for enhanced repository performance

Fluid flow and chemical transport in porous media are the macroscopic consequences of pore structure, which integrates geometry (e.g., pore size and surface area, pore-size distribution) and topology (e.g., pore connectivity). Low-permeability geological media whose pores are poorly interconnected will exhibit the characteristics of anomalous diffusion and sample size-dependent effective porosity, which will strongly impact long-term net diffusion and retention of radionuclides in geological repository settings involving different host rocks and barrier materials. A suite of innovative and complementary experimental approaches is utilized to study the microscopic pore structure and macroscopic fluid flow & chemical transport for a range of host rocks and barrier materials, in addition to standard clay minerals and reference rocks. With a particular focus on quantifying the presence and magnitude of “isolated” pores for a reduced effective porosity in low-permeability geomedia, the integrated methodologies for basic properties and pore structure characterization of these geomedia include X-ray diffraction, thin section petrography, grain size distribution, water immersion porosimetry after vacuum-pulling for full saturation, mercury intrusion porosimetry, nitrogen physisorption, scanning electron microscopy, X-ray computed tomography, and (ultra-)small angle neutron (X-ray) scattering. In addition, custom-designed gas diffusion, tracer recipe involving a range of anionic and cationic chemicals with subsequent analyses by laser ablation and inductively coupled plasma-mass spectrometry, along with batch sorption, column transport, and imbibition tests were conducted for coupled effects of pore structure and chemical retention/transport. From the perspectives of pore structure in conjunction with multiple and complementary approaches to examining a range of sample sizes under different observational scales, we find that the poor pore connectivity is prevalent in low-permeability media (mudstone and crystalline rock) that is related to geological processes (e.g., compaction, diagenesis and thermal maturation). For example, the deep and organic matter-rich mudstones have a much smaller effective porosity than the total porosity (as a result of poor pore connectivity) and associated diffusion coefficient, and the effective porosity & diffusion coefficients are also dependent upon the sample sizes used in the measurement. Similarly, most of the pore space in the shallow mudstone is also controlled by pore-throat diameters in the 5-50 nm range of intergranular pore types from its fine-grained nature, but with an overall good pore connectivity. However, the nm-sized pore space (physically pore-network architecture) and strong sorption capacities (chemical retention from clay minerals) of both shallow and deep mudstones lead to the synergistic retention of cationic radionuclides and their utilities as effective host rocks and barrier materials. Our unique approaches of studying how the micro-scale pore structure affect macro-scale fluid flow, diffusion & retention, and chemical transport produce improved mechanistic understanding, and realistic quantification, of diffusion and retention of typical radionuclides in a range of generic host rocks and barrier materials (clay/shale, salt, crystalline rock, and tuff), with the overall results leading to scientifically-based understanding of enhanced isolation (from both diffusion and retention) of radionuclides and improved confidence on the long-term performance of geological repository to store high-level radioactive wastes. In addition to the training of 25 undergraduates, graduates, and postdocs of UTA, the scientists (organizations) involved in performing this work (e.g., discussion, sample sharing, and operation of SANS and SAXS instruments) include Ed Matteo, Yifeng Wang, and Kristopher Kuhlman (Sandia National Laboratories), Jens Birkholzer, Liange Zheng, Tim Kneafsey, and Sharon Borglin (Lawrence Berkeley National Laboratory), Mavrik Zavarin (Lawrence Livermore National Laboratory), Yukio Tachi and Yuta Fukatsu (Japan Atomic Energy Agency), Mieke de Craen (Euridice, Belgium), Markus Bleuel (NIST), Wei-Ren Chen, Gergely Nagy, Changwoo Do, William Heller, Larry Anovitz, and Kenneth Littrell (ORNL), as well as Jan Illvsky, Ivan Kuzmenko, Ju-Sang Park and Jon Almers (ANL). Key deliverables include a total of 13 peer-reviewed journal articles (nine published and three under review), 23 presentations at scientific conferences (AAPG, AAPG Southwest Section, AGU, Asian Clay Conference, GSA, GSA South-Central Section, IHLRWM, InterPore, International Conference on Chemistry and Migration Behavior of Actinides and Fission Products in the Geosphere, International Conference on Coupled Processes in Fractured Geological Media: Observation, Modeling and Application), and academic institutions (UTA, New Mexico State University; University of Poitiers, France; University of Helsinki, Finland; Uppsala University, Sweden; Istanbul Technical University, Turkey) and other organizations (Andra, France; Posiva Oy, Finland).

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The Enhanced Sealing Project, Monitoring a Full-Scale Shaft Seal at Canada's Underground Research Laboratory (2009-2015) - 16070

A full-scale shaft seal was designed and installed in the 5-m-diameter main shaft at Canada's Underground Research Laboratory (URL) in 2009. Located where the main shaft intersects a water-bearing fracture zone at a depth of approximately 273 m, a 6-m-thick clay component was sandwiched between two 3-m-thick concrete components. The clay component was 40:60 mixture of bentonite and quartz/feldspar sand and spans approximately 1 m beyond the maximum identified vertical extent of the fracture zone. Its purpose is to limit the mixing of saline groundwater from the deeper region with the fresher, near-surface groundwater. Its construction was part of Canada's Nuclear Legacy Liabilities Program (NLLP). The 1. phase of the Enhanced Sealing Project (ESP), a jointly-funded Research, Development and Demonstration (RD and D) project was started in 2009 to monitor the saturation of the shaft seal until the end of 2013. The 2. phase, extended monitoring of the ESP is supported by Posiva (Finland), Andra (France) and CNL (Canada) and this will run until the end of 2016. The slow evolution of the seal and its surroundings and complete removal of the surface facilities at the URL site in 2014, has meant that the ESP monitoring has evolved to a stand-alone, lower frequency data collection program. Information gained from the ESP monitoring is applicable to support the development of sealing technology to safely close a deep geological repository for the permanent disposal of radioactive wastes. (authors)

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Dehalogenation reactions between halide salts and phosphate compounds

Reactions between phosphoric acid [H 3 PO 4 ] or ammonium hydrogen phosphates [i.e., NH 4 H 2 PO 4 , (NH 4 ) 2 HPO 4 ] and halide salts can be used to dehalogenate (remove halides from) salt-based waste streams, where the process of removing halides yields products that have more efficient disposal pathways for repository storage. In this context, the term efficiency is defined as higher waste loadings and simplified immobilization processes with potential for recycle of certain salt components (e.g., 37 Cl as H 37 Cl or NH 4 37 Cl). The main streams identified for these processes are nuclear wastes generated during electrochemical reprocessing of used nuclear fuel as well as used halide salts from molten salt reactor operation. The potential byproducts of these reactions are fairly consistent across the range of halide species (i.e., F, Cl, Br, I) where the most common are hydrogen halides [e.g., HCl (g) ] or ammonium halides (e.g., NH 4 Cl). However, trihalide compounds (e.g., NCl 3 ), nitrogen triiodide ammine adducts [NI 3 ·(NH 3 ) x ], and ammonium triiodide (NH 4 I 3 ) are also possible. Several of these byproducts (i.e., NCl 3 , NBr 3 , NI 3 , and NH 4 I 3 ) are shock-sensitive contact explosives so their production in these processes must be tracked and carefully controlled, which includes methods of immediate neutralization upon production such as direct transport to a caustic scrubber for dissolution. Several benefits arise from utilizing H 3 PO 4 as the phosphate additive during dehalogenation reactions for making iron phosphate waste forms including more oxidized iron (higher Fe 3+ :Fe 2+ ratios), higher chemical durabilities, and the avoidance of trihalides, but the byproducts are hydrogen halides, which are corrosive and require special handling.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Computational scheme for burnup credit applied to long term waste disposal - 14448

The application of burnup credit (BUC) to criticality safety evaluations (CSE) of geological repositories for long term disposal of spent nuclear fuel (SNF) is under consideration in Switzerland. Therefore, in collaboration with the national waste management organization, PSI is conducting research towards the development of a BUC methodology tailored for Swiss specific applications. As part of this research, a computational sequence from the in-core burnup phase, through the interim storage, until final disposal is being established. The tools selected for this methodology, referred to as BUCSS-R (Burnup Credit System for Swiss reactors and Repositories), are CASMO for the in-core fuel depletion, the decay module of SERPENT for the decay calculations, and MCNP for the criticality calculations. In connection to BUCSS-R, a computational interface to validated reference core models of the Swiss reactors is being developed. The purpose is to apply reconstructed depletion histories based on real core operation conditions with CASMO assembly calculations, to transfer the pin wise nuclide compositions at discharge to the subsequent SERPENT decay calculations, and afterwards to the final MCNP criticality calculations. An objective of this paper is to present the elements of the BUCSS-R methodology, and to show a first application to so-called bounding analyses of a preliminary Swiss disposal canister design loaded with fuel assembly types operated in a Swiss PWR reactor. The rationale for these bounding analyses is that before starting the integration of methodologies for bias and uncertainty quantifications, the span of the k-eff variation between fresh and burnt fuel loadings should first be quantified in a realistic manner. The obtained results are reported in this paper together with studies performed in order to evaluate the effects from burnup levels and nuclides credit at flooded conditions on the k-eff values. (authors)

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Performance assessment modeling of a generic SNF/HLW repository in salt with coupled thermal-hydrologic effects - 15423

This paper describes advances in performance assessment modeling of deep geologic repositories facilitated by a massively parallel, high-performance computing (HPC) environment. Our new Generic Disposal System Analysis (GDSA) Framework utilizes the massively parallel PFLOTRAN multi-physics code to simulate repository performance in the presence of coupled thermal-hydrologic-chemical processes, linked to DAKOTA, an HPC uncertainty sampling and propagation code that provides nested parallelism for multi-realization performance assessment and sensitivity analysis. This enhanced performance assessment (PA) modeling capability is demonstrated with deterministic and probabilistic simulations of a generic repository for SNF and HLW in bedded salt host rock, by comparing repository performance between a case with heat-generating waste ('thermal' case) and a case without heat generation ('isothermal' case). The simulation results provide preliminary insights into the effect of multi-physics processes and thermal-hydrologic coupling on the long-term behavior of a reference-case salt repository. (authors)

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Unsaturated alluvium disposal modelling with improved geological realism

This report describes specific activities in the Fiscal Year (FY) 2023 associated with the Geologic Disposal Safety Assessment (GDSA) Repository Systems Analysis (RSA) work package funded by the Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Spent Fuel and Waste Disposition (SFWD). The overall objective of the GDSA RSA work package is to develop generic deep geologic repository concepts and system performance assessment (PA) models in several potential host-rock environments, and to simulate and analyze these generic repository concepts and models using the GDSA Framework toolkit and other tools as needed. The specific GDSA goal addressed in this report is reference case development, simulation, and analysis for the unsaturated alluvium (UZ), one of the four potential host-rocks considered by the GDSA. Further, we aim to exercise the simulation tools and methodologies under development by GDSA for PA modelling.

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Framework for The Analysis of US Permanent Geologic Disposal Options for Radioactive Waste

The U.S. study of permanent geologic disposal options of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) provided a technical basis for informing policy decisions regarding strategies for the management and disposal of radioactive waste requiring geologic isolation through the evaluation of potential impacts of waste forms on the feasibility and performance of representative generic concepts for geologic disposal. The goal of the study was to help inform relevant policy questions including: Is a "one-size-fits-all" repository a good strategic option for disposal? Do different waste types and forms perform differently enough in different disposal concepts that they warrant different treatment? Do some disposal concepts perform significantly better with or without specific waste types or forms?

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Prediction of the thermal-hydraulic-mechanical response of a geological repository at large scale and sensitivity analyses

To better understand the Thermal-Hydraulic-Mechanical (THM) response of the rock caused by the high-level radioactive waste (HLW) released heat in the Callovo-Oxfordian formation (COx) in the near-field and far-field areas, a series of coupled THM modelling is performed. This study presents a case study based on the French geological repository for HLW (Cigéo project) to better assess modelling of deep geological repositories (DGR) within the DECOVALEX-2019 framework. In this study, the proposed coupled THM model, implemented in the platform of COMSOL, is validated by comparing the modelled THM results for a point heat source in an infinite rock mass with an analytical solution. It is also validated by comparing its thermal component for the Base Case with accurate results calculated using another numerical method. To shorten the calculation time, the DGR is modelled using six cuboid blocks representing 162 HLW cells and only six central HLW cells of interest modelled in detail. This simplification is validated. This model can easily be used to model not only the near-field THM response of the rock but also the far-field THM response. The influence of the boundary conditions applied on the gallery wall, HLW cell walls, and external surfaces of the model and the influence of different vertical dimensions of the geometry are studied. The sensitivity analyses of the THM parameters of COx on the THM response at different locations are performed. Finally, the influence of using a 2-dimensional (2D) model to represent a 3-dimensional (3D) repository is also investigated.

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U.S. Senate Bill S.854-IS - A Maladjusted Politicized Maze for Consent-Based Siting of New HLW-Repositories - 16019

At the end of 2015, the USA's HLW-disposal program had been on hold since 2010 pending the enactment of enabling legislation for one or both of: 1) The Yucca Mountain HLW repository; and/or 2) The law proposed in Senate Bill S.854-IS 'to establish a new organization to manage nuclear waste, provide a consensual process for siting nuclear waste facilities, ensure adequate funding for managing nuclear waste, and for other purposes'. But, S.854-IS does not acknowledge the existence of the Yucca Mountain HLW repository, which, if re-enabled, would affect both when another HLW repository would be needed and its disposal capacity. Furthermore, it neither timely nor meaningfully integrates waste generators, facility-host(s) and other directly affected parties with the new 'implementing' organizations deemed to be more susceptible to political influence and interferences than those plaguing the Yucca Mountain HLW repository program since 1987 that resulted in a multitude of contentions, lawsuits, delays, and increased costs. For these and other reasons very compelling and conclusive to the author, the legal status of the YM HLW repository should be 'definitive' before a law based upon S.854-IS is enacted. Instead, the near-term focus should be on: 1. Defining 'consent-based' quantitatively. 2. Promptly siting and opening at least one large, expandable, 'consolidated' HLW-storage facility allowing the government to begin taking title to commercially-generated HLW; most-expeditiously accomplished by amending existing legislation. 3. Updating S.854-IS based on lessons learned about siting, developing, and licensing nuclear-waste-storage and -disposal facilities in the USA and abroad the past 30+ years. (authors)

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Modeling and Analysis of the Transport and Disposal of Beryllium Moderator Blocks and Greater than Class C (GTCC) Waste

Radioactive waste in the United states is categorized based on its source, radioactivity, and security risk. The categorization method employed by the Nuclear Regulatory Committee (NRC) for low-level radioactive waste is not currently applied to any waste generated by the Department of Energy (DOE) or disposed of in DOE facilities. As a result, there is a large volume of DOE-generated waste with characteristics similar to the NRC’s “Greater than Class C” (GTCC) waste category which presently do not have a path for long-term disposal. Much of this waste cannot undergo the standard commercial disposition process due to it being categorized as Transuranic (TRU) waste under DOE guidelines, due to elevated concentrations of key fission products. Possible solutions to this dilemma are explored in the Environmental Impact Statement for the “Disposal of Greater-Than-Class-C Low Level Radioactive Wave” (EIS-0375). This paper analyzes several EIS possible paths for disposal for this “orphaned” waste. One example that highlights this categorization issue is the spent beryllium cladding that has been extracted from the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) throughout its operational lifetime. These beryllium blocks surround the reactor’s main chambers and components and serve as neutron moderators. This paper will use this particular waste form to analyze the economic and technological viability of transporting and storing this material under the storage and transportation criteria of the Waste isolation Pilot Plant (WIPP), which was deemed the most feasible disposition path according to the EIS. The example of this waste form used in this paper that highlights this categorization issue is the spent beryllium cladding extracted from Idaho National Laboratory’s (INL) Advanced Test Reactor (ATR). These beryllium blocks surround the reactor’s components and serve as neutron moderators, and thus have accumulated high concentration of high-activity transuranic isotopes. This paper analyzed the viability of transporting and storing this material under the storage and transportation criteria of the five primary disposition paths covered in EIS-0375. An array of calculations was carried out to assess the viability of the various disposition paths for the beryllium shipments as well as other waste shipments that fall within the GTCC category. Modeling with MCNP 6.2 was conducted to determine whether the beryllium blocks could be safely stored and transported within a 72-B cask; the standard shipping container for remote-handled, transuranic waste, while also meeting regulatory limits at various disposition paths. A cost analysis of the transportation and long-term disposal paths mentioned in the EIS was also carried out using available data and information from similar waste shipments. Lastly, a geochemical analysis of the various geological repository discussed in the EIS was also carried out using the Geochemists Workbench Release 14. Long-term disposal of the beryllium blocks at the Waste Isolation Pilot Plant (WIPP) proved to be the most cost-effective long-term disposition option of the ones considered in the EIS. A dose rate calculation at both contact and remote-handling distances indicate that the analyzed beryllium shipments should not exceed the exposure limits at any of the considered locations. Greater-than-class-C waste has been left in a regulatory state of limbo for years, resulting in backlogs of inventory across several research sites in the United States. Due to its high activity and presence of transuranic isotopes, it is imperative to ensure that it remains inaccessible and sequestered both in its short-term interim as well as a long-term geologic time scale. The information and assessments done in the paper could potentially serve as a reference for any future shipments of this waste form at the WIPP facility as well as other disposition paths that may be considered in the future.

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Impacts of Fast Critical Assembly Fuel Discards on Liquid Waste Processes

The Savannah River Nuclear Solutions (SRNS) Fast Critical Assembly (FCA) mission is reestablishing the electrolytic dissolver for processing of Pu and Pu-U materials clad in stainless steel (SS). H-Canyon is planning to dissolve and neutralize FCA fuel without recovering the special nuclear material (i.e., Pu) prior to discarding to the Concentration, Storage, and Transfer Facilities (CSTF) operated by the Savannah River Mission Completion (SRMC) Liquid Waste (LW) Organization. The FCA discards will be combined with sludge in the CSTF after Low Temperature Aluminum Dissolution (LTAD) if needed. The combined waste will be washed, concentrated, and vitrified at the Defense Waste Processing Facility (DWPF). The high level waste canisters produced will be temporarily stored in the Glass Waste Storage Buildings prior to transfer to a future federal repository. Decants from LTAD and sludge washing will be combined with DWPF recycle and dissolved salt cake to prepare salt batches for processing in the Salt Waste Processing Facility. The resulting decontaminated salt solution will be processed in the Saltstone Production Facility.

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Cold crucible induction melter studies for making glass ceramic waste forms: A feasibility assessment

Glass ceramics are being developed to immobilize fission products, separated from used nuclear fuel by aqueous reprocessing, into a stable waste form suitable for disposal in a geological repository. This work documents the glass ceramic formulation at bench scale and for a scaled melter test performed in a pilot-scale (approximately 1/4 scale) cold crucible induction melter (CCIM). Melt viscosity, electrical conductivity, and crystallization behavior upon cooling were measured on a small set of compositions to select a formulation for melter testing. Property measurements also identified a temperature range for melter operation and cooling profiles necessary to crystallize the targeted phases in the waste form. Bench scale and melter run results successfully demonstrate the processability of the glass ceramic using the CCIM melter technology.

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Defense waste processing facility (DWPF) liquids model: revisions for processing higher TIO 2 containing glasses

Radioactive high level waste (HLW) at the Savannah River Site (SRS) has successfully been vitrified into borosilicate glass in the Defense Waste Processing Facility (DWPF) since 1996. Vitrification requires stringent product/process (P/P) constraints since the glass cannot be reworked once it is poured into ten foot tall by two foot diameter canisters. A unique “feed forward” statistical process control (SPC) was developed for this control rather than statistical quality control (SQC). In SPC, the feed composition to the DWPF melter is controlled prior to vitrification. In SQC, the glass product would be sampled after it is vitrified. Individual glass property-composition models form the basis for the “feed forward” SPC. The models transform constraints on the melt and glass properties into constraints on the feed composition going to the melter in order to guarantee, at the 95% confidence level, that the feed will be processable and that the durability of the resulting waste form will be acceptable to a geologic repository. This report documents the development of revised TiO 2 , Na 2 O, Li 2 O and Fe 2 O 3 coefficients in the SWPF liquidus model and revised coefficients (a, b, c, and d).

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Site Selection Restart in Germany Results of the Site Selection Commission - 17017

Since 2011 Germany is pursuing a phase-out strategy concerning the use of nuclear power for electricity production. This decision was strongly influenced by the Fukushima event. In 2013 the federal government announced that they also had achieved an agreement with the Federal States in Germany on a law to restart the site selection for a repository for spent fuel and high-level heat-generating waste from scratch. The new law was passed in July 2013. At first a 34-member commission had been installed in April 2014 to evaluate the Site Selection Law [1] and to develop basic principles for site selection, including safety requirements and selection criteria for rock formations. The commission included representatives from the parliament, civil society organizations, industry, environmental organizations, trade unions and scientists. The commission forwarded its recommendations in summer 2016. The evaluated law should come into force in 2017. The stepwise approach for finding a suitable site for a repository for Germany's heat-generating waste might then start in 2017. As shown in publication [2], the site selection process will take several decades. The preferred solution based on the present state of the art is the final disposal in deep geological formations in a mine with the option 'Reversibility and Retrievability (R and R)'. The scientific challenges and the upcoming R and D program are complex. Therefore a 'step-back option' (learning process) is foreseen in the program. The R and D program has to react flexibly to new findings, demands or modifications of criteria or requirements. (authors)

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Underground test coordination at the waste isolation pilot plant - 15229

Underground test coordination is a critical component of safe and efficient conduct of testing operations in a working underground research laboratory (URL). This is particularly important when planning and implementing large, multi-organizational field-scale test programs, when the URL is co-located with an operating repository, or when the host facility, such as the Waste Isolation Pilot Plant (WIPP), is recovering from off-normal events. As such, the Department of Energy Carlsbad Field Office developed an Integrated Project Team (IPT) charter in late 2013 establishing the WIPP underground Test Coordination Office (TCO). The IPT was established to ensure that underground science activities are coordinated, implemented, and managed in an integrated, consistent, efficient, cost-effective, safe, and environmentally compliant manner that maintains quality assurance standards. This is done while ensuring testing operations do not interfere with the WIPP's primary mission, the disposal of transuranic (TRU) waste, and at present, the recovery of WIPP operations. This paper reports the current status of the WIPP underground science and testing programs within the framework of the WIPP stabilization and recovery activities and describes the role of the TCO for current activities as well as for the long-term success of an underground research laboratory. As a result of the underground fire and radiological incident of early February 2014, during this recent period of WIPP recovery, the TCO is ensuring the requirements of the scientific programs are met and that work activities are planned and implemented in compliance with enhanced work control processes and corrective actions imposed at WIPP. This paper addresses the impacts of the accidents at WIPP on the many science experiments that were in process when the events occurred. It presents the plans for their recovery in concert with WIPP's recovery plans, and summarizes the impacts of new underground processes and procedures that will be imposed on the science work as a result. (authors)

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Postclosure Criticality Analysis Results

The United States Department of Energy’s (DOE) Office of Nuclear Energy’s Spent Fuel and Waste Science and Technology Campaign seeks to better understand the technical basis, risks, and uncertainty associated with the safe and secure disposition of spent nuclear fuel (SNF) and high-level radioactive waste. Commercial nuclear power generation in the United States has resulted in thousands of metric tons of SNF, the disposal of which is the responsibility of DOE (Nuclear Waste Policy Act of 1982, as amended). Any repository licensed to dispose of SNF must meet requirements regarding the long-term performance of that repository. The evaluation of long-term performance of the repository may need to consider the SNF achieving a critical configuration during the postclosure period. Of particular interest is the potential for this situation to occur in dual-purpose canisters (DPCs), which are currently licensed and being used to store and transport SNF but were not designed for permanent geologic disposal. DOE has been considering disposing of SNF in DPCs to avoid the costs and worker dose associated with repackaging the SNF currently stored in DPCs into repository-specific canisters. This report examines the consequences of postclosure criticality to provide technical support to DOE in developing a disposal plan.

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Germany's First Steps in the Restarted Site Selection Process - 18015

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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