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DOE OSTI2021

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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DOE OSTI2024

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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DOE OSTI2017

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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DOE OSTI2017

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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DOE OSTI2015

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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DOE OSTI2023

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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DOE OSTI2018

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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DOE OSTI2019

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

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DOE OSTI2016

Evaluation of Bentonite Engineered Barrier Performance under Repository Conditions

Nuclear energy has been an important source of power throughout the world for several decades, but with that power comes substantial amounts of waste. The USA alone is projected to produce 77,100 MTHM (34,700 MCi) of commercial spent fuel by the year 2020. There has been a growing sense of urgency to determine a method of disposal for all types of nuclear waste. Underground storage has been proposed as the safest means of long term storage and disposal by both the Academy of Sciences and the Blue Ribbon Commission with a focus on deep geologic disposal. Because deep geological repositories (DGR) pose as the safest method of disposal due to the isolation of the waste and the low-permeability of the surrounding material, it is crucial to understand the science behind these repositories. Many scientists have concern with DGRs because of all the unanswered scientific questions. With a global inventory of 300,000 MTHM, it is imperative to understand how DGRs will retain the nuclear waste over a time span of up to a projected 1 million years. Because of the amount of nuclear waste increases each year, there is an inherent need to study the viability of a repository in order to begin implementation of that repository. Based on the half-lives of each radionuclide in nuclear waste, after approximately 100 years of storage, the heat production from the decay of radionuclides becomes dominated by the long-lived actinides. Because the decay of the nuclear waste results in an elevated temperature within the canister, That mixed with water intrusion can corrode the canister and allow radionuclides to leach into the surrounding environment. In an effort to retain these radionuclides within the repositories, most all deep geologic repository designs incorporate an engineered multi-barrier approach. Once the radionuclides begin to leach out of the canisters, the next barrier to retain them is a layer of compacted clay such as a smectite. A mechanistic understanding of mobility at elevated temperatures is crucial in order to evaluate the performance of the engineered clay barrier system within a deep geological repository. Because the clay barrier is compacted so tightly, the only method of mobility for radionuclides is through molecular diffusion. Assessing the factors that affect the rates at which various radionuclides will diffuse through these clays is important in the performance assessment of the repository system as a whole. Neptunium-237 has been identified as a major actinide of concern for long term storage, dominating the radioactivity emitted from the waste between 10,000 to 100,000 years of storage. Under oxidizing conditions, this radionuclide exists mainly as Np(V). In this oxidation state, the neptunium exists chemically as the NpO{sub 2}{sup +} ion which exhibits enhanced mobility characteristics such as high solubility and low affinity for sorption to surfaces. Because of this, {sup 237}Np(V) has been identified as the most mobile radionuclide in nuclear waste and thus a major radionuclide of concern, motivating research into the mobility of {sup 237}Np(V) through the compacted clay barrier. Bentonite is one of the most common backfill materials used in repositories. This clayey rock's main constituent is the mineral Na-montmorillonite [(Ca{sub 0.06}Na{sub 0.14}K{sub 0.05})(Al{sub 1.57}Fe{sup 3+} {sub 0.17}Mg{sub 0.25})(Al{sub 0.04}Si{sub 3.96}O{sub 10} - (OH){sub 2})] which gives this material the advantageous properties required for a repository such as a high swelling capacity. Because of this high affinity for absorbing water, montmorillonite is the ideal candidate for backfill material in a deep geological repository. Adsorption of radionuclides onto the surface of the compacted clay is a major mechanism for retardation of mobility. Because diffusion and adsorption are integrated into this process, it is important to characterize Np adsorption onto the surface of compacted montmorillonite. Previous studies have quantified the adsorption of Np onto montmorillonite at variable ionic strengths and temperatures. Furthermore, several studies have focused on the diffusion of Np through compacted clay, but no other studies have determined diffusion coefficients at elevated temperatures of up to 80 deg. C. (authors)

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DOE OSTI2024

The DECOVALEX international collaboration on modeling of coupled subsurface processes and its contribution to confidence building in radioactive waste disposal

Abstract The long-lived radiotoxicity of the high-level radioactive waste generated by nuclear power plants requires safe isolation from the biosphere for many hundreds of thousands of years. An international consensus has emerged that such isolation can best be provided by disposal in mined geologic repositories, a strategy that today is pursued by most countries dealing with radioactive waste. However, the need to predict the performance of such repositories over very long time periods generates large uncertainties that have to be accounted for in safety assessments. The findings from such safety assessments need to be conveyed to all stakeholders in a clear way, such that public confidence in geologic disposal solutions can be achieved. It is suggested here that close international collaboration on the technical aspects of geologic waste disposal has helped, and will continue to help, building trust and increasing confidence. This paper discusses a particular international collaboration initiative referred to as DECOVALEX, which brings together multiple teams and disciplines to collectively tackle complex experimental and modeling challenges related to geologic disposal. By describing how DECOVALEX works and by providing joint research examples, a case is made that such international collaboration contributes to knowledge transfer and confidence building in radioactive waste disposal science.

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DOE OSTI2023

Analysis of Transient Postclosure Criticality

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 the 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. For an evaluation of the long-term performance of the repository, one of the events that may need to be considered is the SNF achieving a critical configuration during the postclosure period. Of particular interest is the potential behavior of SNF 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.

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DOE OSTI2022

Transient Postclosure Criticality Model Development

A key objective of the United States Department of Energy’s (DOE) Office of Nuclear Energy’s Spent Fuel and Waste Science and Technology (SFWST) Campaign is 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 the 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. For an evaluation of the long-term performance of the repository, one of the events that may need to be considered is the SNF achieving a critical configuration during the postclosure period. Of particular interest is the potential behavior of SNF in dualpurpose canisters (DPCs), which are currently licensed and being used to store and transport SNF but were not designed for permanent geologic disposal.

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DOE OSTI2018

Safety Concept and Safety Assessment for a HLW Repository in Bedded Salt in Germany - 18253

The German site selection act for a repository for vitrified HLW and SNF demands the consideration of various host rock formations for deep geological disposal. One of the eligible and presently investigated host rock formations is bedded rock salt. Within the framework of the national R and D project KOSINA the scientific bases for the design and safety assessment of a repository for HLW and SNF in bedded salt formations were provided. The research findings presented in this paper include the formulation of a safety concept and safety demonstration concept as well as exemplary radiological consequences analysis for a generic repository in bedded rock salt formations. The safety concept describes in a verbal, argument-based way how the natural conditions, processes and technical measures at large yield a status of safety. Fundamental principle of the safety concept is the concentration and containment of the radioactive and other pollutants in the waste in the containment providing rock zone (CRZ), which ensures isolation from the biosphere. Based on the principle of safe containment inside the CRZ guiding principles, design objectives and strategic measures are derived. The safety demonstration concept describes the means, such as analyses and arguments, which are used in the safety case to demonstrate the safety of a repository system in bedded salt on the basis of the safety concept and its guiding principles and design objectives. The decisive elements are: the demonstration of integrity of the geological barrier, the demonstration of integrity of the geotechnical barriers, the scenario analysis, and the evaluation of the identified scenarios. These elements are complemented in the safety demonstration by concepts on how to consider criticality, non-radiological protection goals and operational safety. The radiological consequences analysis primarily consists of the calculation of the radiological safety indicator for different test cases and scenarios for the various repository design and waste emplacement options. Within the KOSINA project the disposal concepts of drift disposal, horizontal and vertical borehole disposals were investigated in detail by means of computational flow and transport modelling of the repository mine (near field) and assessment of the radiological consequences (biosphere model). The applied software package RepoTREND is introduced; the modelling concept and employed radiological indicators are explained. Examples of the modelling results for the reference cases of the drift disposal repository concept for bedded salt will be given, demonstrating the capabilities of the model and the applicability to close-to-reality repository systems for deep geological disposal. (authors)

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DOE OSTI2018

Pathways towards Valid Data: Assessment of Optimized Packaging Strategies for Core Components - 18158

Every company that runs nuclear power plants needs a strategy for disposal of the radioactive waste and for decommissioning and dismantling of its nuclear reactors. The dismantling progress of several German NPPs is already in an advanced stage. Hence, particularly highly activated core components have to be disposed of as radioactive waste. Nevertheless, this is also an issue during the operating phase due to necessary replacements of highly activated core components such as PWR control elements or RPV heads. The conditioning of waste packages for final disposal in the German repository Konrad requires the core components to be dismantled, packaged and dried. Therefore, a qualified packaging strategy is needed to ensure a reliable compliance with the protection objectives of the repository and an optimized utilization of prescribed contingents. Exemplified by PWR control elements a pathway is drawn how to perform an optimized packaging strategy in compliance with the protection objectives of the repository based on only few relevant radionuclides derived from activation calculations. For these calculations, particularly the influence of trace impurities in the materials has to be considered and their amounts have to be verified or analyzed by sampling. Furthermore, low nuclide activities from contamination cannot simply be neglected because some of them are obliged to be declared or their radiological relevance has not been sufficiently determined. Due to various influence factors, qualified sampling strategies need carefully considered concepts. Therefore, detailed planning facilitates the qualification and approval process and reduces the costs for analyses and the number of containers. (authors)

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DOE OSTI2015

Simultaneous Thermal Analysis of WIPP and LANL Waste Drum Samples: A Preliminary Report

On Friday, February 14, 2014, an incident in P7R7 of the WIPP underground repository released radioactive material into the environment. The direct cause of the event was a breached transuranic (TRU) waste container, subsequently identified as Drum 68660. Photographic and other evidence indicates that the breach of 68660 was caused by an exothermic event. Subsequent investigations (Britt, 2015; Clark and Funk, 2015; Wilson et al., 2015; Clark, 2015) indicate that the combination of nitrate salts, pH neutralizing chemicals, and organic-based adsorbent represented a potentially energetic mixture. The materials inside the breached steel drum consisted of remediated, 30- to 40-year old, Pu processing wastes from LANL. The contents were processed and repackaged in 2014. Processing activities at LANL included: 1) neutralization of acidic liquid contents, 2) sorption of the neutralized liquid, and 3) mixing of acidic nitrate salts with an absorber to meet waste acceptance criteria. The contents of 68660 and its sibling, 68685, were derived from the same parent drum, S855793. Drum S855793 originally contained ten plastic bags of acidic nitrate salts, and four bags of mixed nitrate and oxalate salts generated in 1985 by Pu recovery operations. These salts were predominantly oxalic acid, hydrated nitrate salts of Mg, Ca, and Fe, anhydrous Na(NO 3 ), and minor amounts of anhydrous and hydrous nitrate salts of Pb, Al, K, Cr, and Ni. Other major components include sorbed water, nitric acid, dissolved nitrates, an absorbent (Swheat Scoop®) and a neutralizer (KolorSafe®). The contents of 68660 are described in greater detail in Appendix E of Wilson et al. (2015)

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DOE OSTI2019

A Python Package for Fuel Cycle and Waste Management Analysis - 19545

The future of nuclear power utilization in the United States may involve a transition from the once-through light water reactor fuel cycle to a so-called advanced fuel cycle. An advanced fuel cycle may offer improvements in natural resource utilization, reactor safety, and environmental impact. In addition, adoption of an advanced fuel cycle will result in changes in the amount, composition, and disposal of the nuclear wastes they produce. Characteristics intrinsic to the wastes themselves, such as mass and radioactivity at specific times, were used as waste management metrics in the Department of Energy 2014 Fuel Cycle Evaluation and Screening (FCES) study. Binning of these metrics informed fuel cycle according to relative potential benefit. However, waste properties alone cannot qualify management strategy. Therefore, comparing fuel cycles from the perspective of waste management can provide additional insights. To that end, an object-oriented Python package has been developed to perform fuel cycle waste management calculations. The package, named nwpy (short for Nuclear Waste analysis in Python), mimics the mass flow operations reported in the FCES study and interfaces with the ORIGEN-S code to assess metric data as reported in the FCES. In addition, this Python package determines the loading of waste streams into forms suitable for final disposal in a geological repository and uses such data for repository performance assessment. This work describes the Python package methodology, assumptions, and capabilities. As input data, the package requires information about the fuel cycle such as the mass of fuel discharged, the cooling time between reactor discharge and subsequent operations, and the separations process efficiencies and waste forms it produces. In addition, the isotopic composition of the reactor discharge stream, such as that available from the FCES database, is required. With that, it can perform prescribed mass flow and decay operations throughout the back end of the fuel cycle, the latter carried out by the ORIGEN-S code distributed with the SCALE6.2 software suite. The result is time-dependent data characterizing the waste on a per-package basis. Package loading for waste streams depends on the stream form (i.e. UNF vs. HLW, or glass HLW from aqueous reprocessing vs. ceramic HLW from electrochemical processing) and is based on literature review and engineering assumption. As an example of the application of the code, calculations for the areal footprint of direct-contact repositories for different fuel cycles is shown. Development is ongoing, and the package is expected to grow incorporating capabilities for waste management analyses such as waste transportation and radionuclide transport through the biosphere. (authors)

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DOE OSTI2014

Investigations of Dual-Purpose Canister Direct Disposal Feasibility (FY14)

Results reported here continue to support the FY13 conclusion that direct disposal of DPCs is technically feasible, at least for some DPCs, and for some disposal concepts (geologic host media). Much of the work performed has reached a point where site-specific information would be needed for further resolution. Several activities in FY14 have focused on clay/shale media because of potential complications resulting from low thermal conductivity, limited temperature tolerance, and the need to construct hundreds of kilometers of emplacement drifts that will remain stable for at least 50 years. Technologies for rapid excavation and liner installation have significantly advanced in the past 20 years. Tunnel boring machines are the clear choice for large-scale excavation. The first TBM excavations, including some constructed in clay or shale media, are now approaching 50 years of service. Open-type TBMs are a good choice but the repository host formation would need to have sufficient compressive strength for the excavation face to be self-supporting. One way to improve the strength-stress relationship is to reduce the repository depth in soft formations (e.g., 300 m depth). The fastest construction appears to be possible using TBMs with a single-pass liner made of pre-fabricated concrete segments. Major projects have been constructed with prefabricated segmented liner systems, and with cast-in-place concrete liners. Cost comparisons show that differences in project management and financing may be larger cost factors than the choice of liner systems. Costs for large-scale excavation and construction in clay/shale media vary widely but can probably be limited to $10,000 per linear meter, which is similar to previous estimates for repository construction. Concepts for disposal of DPC-based waste packages in clay/shale media are associated with thermal management challenges because of the relatively low thermal conductivity and limited temperature tolerance. Peak temperature limits of 100°C or lower for clay-rich materials have been selected by some international programs, but a limit above 100°C could help to shorten the duration of surface decay storage and repository ventilation. The effects of locally higher peak temperatures on repository performance need to be evaluated (in addition to the effects at lower temperatures). This report describes a modeling approach that couples the TOUGH2 and FLAC3D codes to represent thermally driven THM processes, as a demonstration of the types of models needed.

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DOE OSTI2016

Site Selection in Germany - Status of the Evaluation of the Site Selection Law - 16122

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 active heat producing waste from scratch. The consequence of this law is a delay of at least two decades to start operation of a final disposal site and additional costs of at least EUR 2.7 billion. 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 and to develop basic principles for site selection, including safety requirements and selection criteria for rock formations. The commission includes representatives from the parliament, academia, civil society organizations, industry, the environmental organizations and trade unions and should forward its recommendations after a delay in starting the evaluation process now in summer 2016. The present law will then be reviewed. The site selection then might start after the next federal election in 2017 at the earliest probably based on a new site selection law. A new repository site should be determined till 2031 and for this site the more detailed site investigation will take place followed by a detailed safety analysis, before the erection of the repository can start. Based on the present procedural steps it seems to be rather unlikely to determine a repository site till 2031. As shown in the publication, there will be a delay of at least 20 years compared to the schedule given in the site selection law until a repository site can be determined. (authors)

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