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DOE OSTI2017

Compressible Arch Segments for the Cigeo Access Ramps, Drifts and Vaults - A Field Test - 17011

Andra is currently implementing the detailed engineering phase of the step-wise design development of Cigeo (the French HLW and IL-LLW Deep Geological Repository). Various technical issues are at stake, including the requirements linked to 'Reversibility / Retrievability', at different stages of the repository life cycle (from its construction start-up to its final closure and monitoring). These reversibility requirements have a significant impact on the way Cigeo is designed, will be progressively developed, operated, later closed and monitored. On a shorter term, one must demonstrate that the waste container retrieval operations are technically safe and feasible. In that sense, the structural dimensioning of the ramps and mains, as well as that of the IL-LLW disposal vaults, is critical, since the concrete liner supporting the rock walls must last over a period of some 100 to 150 years (lapse of time during which retrieval operations are likely to take place). The present paper starts with a description of the geomechanical conditions prevailing in the Callovo-Oxfordian (COX) clay-stone formation (aka the argillite). The structural dimensioning of the IL-LLW disposal vault liner, as calculated at the end of the basic engineering phase is indicated. The thickness values obtained so far have led to considering an alternative design in which a compressible material is positioned between the concrete liner intrados and the rock walls: in such situation, the 'creeping effect' of the rock is mitigated, for a time, by the compressible material (nearly a perfect elastoplastic material), hence reducing the stress level exerted inside the concrete liner. This situation enables to significantly reduce the liner thickness and provides potentially significant saving in the construction of the works concerned (preliminary sensitivity results are given). The following part of the paper is then focused on a practical case story: a technological demonstration (still ongoing) of a tunnel boring experiment, implemented underground (in a dedicated drift of the Meuse Haute Marne URL infrastructures, at scale 1:2 of the Cigeo drifts/vaults). In this excavation and construction experiment, the tunnel boring machine (TBM) lays (via an erector) 'classical' concrete arch segments (forming a ring) in a first portion of the drift and 'compressible' ones (the compressible material is positioned on their extrados) in a second portion of the drift. The technical demonstration objective is two-fold: the capacity to lay compressible arch segments with a TBM is checked, their structural behavior (with time) is compared with that of 'classical' ones. The way the compressible material is produced and then fixed on the arch concrete segment is also documented, c/w a description of the complementary testing campaign to be carried out on a test bench (on surface) of a concrete ring (at a scale 1:2 of Cigeo) formed of compressible arch segments. This paper concludes with a timeline of the experiment. Should the results obtained provide some perspectives of technical and economic improvement (i.e. saving vis-a-vis the 'classical' solution) a decision could be taken by Andra to incorporate this alternative solution in the 'design freeze' of Cigeo underground infrastructures. The positive outcomes of this technical test campaign would also pave the way for specifying the Cigeo TBM characteristics (dimensions and handling equipment), scheduled for manufacturing order around 2020. (authors)

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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 OSTI2017

Westinghouse Quiver - A simple and safe system for the complete handling and storing of failed BWR and PWR fuel rods and fragments - 17049

Background: Extracted leaking fuel rods from fuel assemblies have been stored in the spent fuel pools at the Nuclear Power Plants for many years. Sending those fuel rods to an intermediate storage has been very difficult due to the uncertainty of the technical requirements. The Westinghouse Quiver is a container that can be used from the first moment when the leaking fuel rod is extracted from the Fuel Assembly. The leaking fuel rod can be placed in the Westinghouse Quiver. When the Quiver is filled, it can be sealed and dried and then treated as a regular fuel assembly during intermediate storage and final repository. Purpose of System: The purpose of the system is to store leaking fuel rods, broken rods and fuel rod fragments leak tight in the fuel pool and after vacuum drying and permanent sealing at an intermediate storage facility. The Quiver is designed to be fully compatible with a regular fuel assembly. That means all handling of the Quiver in a Nuclear Power Plant, during transportation or storage is also identical to that of a regular fuel assembly. Fulfilling these criteria together with functional testing, the Quiver is considered acceptable for long time storage of leaking fuel rods, both in dry or wet condition. The Quiver has been qualified in Switzerland for use in the spent fuel pool, intermediate storage, and for a final repository. The system will be implemented as well in Sweden during 2016-2019. (authors)

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DOE OSTI2017

Does the Perfect Waste Package Exist? Development of a Decisional Tool Addressing this Challenge - 17053

The definition of a safe, technically, economically and societally acceptable waste management strategy is highly dependent on the Waste Acceptance Criteria in each country and on the already available disposal routes. It is the reason why there is not a single solution but a combination of optimized solutions depending on the type of waste and on each country's considerations. In fact, when operating a nuclear plant or performing its decommissioning or legacy waste management, one has to consider a large waste diversity in term of natures, types, volumes and radioactivity spectra. All nuclear operators have to take-up the challenge of implementing solutions for conditioning its waste in packages suitable for transport short to long term interim storage and final disposal of it (or prepare for final disposal if the repository is not yet available). Operators have then to answer the following dilemma: either define a complete strategy packaging the waste for final disposal, or 'containerising' it temporarily, pending final disposal. The strategy is highly country dependent; in the first case, one has to take into account the available information and future orientations relative to acceptability of the packages in order to define a robust solution. The approach implemented in France requires a close dialogue between the Operators, the Safety Authority, the Administration and the stakeholders. Moreover it pushes forward whenever possible standardization, contributing to cost minimization. If the second approach has the advantage of leaving the options open, and reducing the initial investments, it inevitably requires future retrieval and re-packaging with potential evolutions/degradation of the initial waste form. It also induces the potential risk to produce additional secondary waste. As an additional drawback, it introduces uncertainties and unknowns related to the future waste management criteria and to the costs. The general idea is that, depending on context, a given route (i.e., treatment, conditioning, and transport), could be considered as efficient or not. Assessing a waste package is certainly highly context dependent. However it should be possible to recognize and to lead an evaluation based on a number of fixed criteria, whatever the context, to be addressed when assessing the development of a final waste package in connection to the overall management route. A waste package is always a combination of a waste container and a waste form and will have to be transported one day from the production site to the final repository or to off-site interim storage facilities. AREVA's approach consists in defining an evaluation grid based on a set of criteria which aim (i) to evaluate whether the different waste treatments, conditionings, transport and packages deserve to be considered within different national contexts, and (ii) to evaluate R and D actions to be performed for improving existing solutions or developing new ones. One of the key lessons to be drawn from AREVA's experience in developing waste packages is that the perfect waste doesn't exist even though the vitrified canister is probably the most efficient one. The performance of a package resulting from a compromise depends on context and is a strength combination between the waste container and the waste form. The approach is to evaluate a waste package development against different set of criteria and to draw optimization diagrams. In such diagram, a given waste package is assessed with respect to selected criteria such as waste form and waste container, stability, level of waste incorporation, integrity lifetime, secondary waste generation, transport, long term behavior, and economical effectiveness, etc. Addressing a waste form is a very key point. This evaluation grid and comparative approach will be presented through examples and comparison providing from AREVA feedback on waste package development related to complex waste management such as problematic waste or legacy waste. (authors)

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DOE OSTI2017

Re-evaluation of Waste Isolation Pilot Plant's Passive Institutional Controls Program - Status - 17116

The United States government set forth in 40CFR194.43 requirements for 'Passive Institutional Controls' for nuclear waste repositories. This regulation stated compliance applications for a repository 'shall include detailed descriptions of the measures that will be employed to preserve knowledge about the location, design and contents of the disposal system.' In the early 1990's the U.S. Department of Energy developed and proposed in the Compliance Certification Application of 1996 a robust and complex technical and engineering Passive Institutional Controls Program including such things as records management, large surface markers, salt berms, buried and above-ground information rooms and small subsurface markers/plaques. The Waste Isolation Pilot Plant Passive Institutional Control program meets the regulatory requirements and is certified by the U.S. Environmental Protection Agency. Presently this program is being re-evaluated in light of the Nuclear Energy Agency's, Records Knowledge and Memory initiative, shrinking government budgets, and the ethical/moral considerations of risks versus expenditures. This paper will discuss progress made to date in re-evaluating the Waste Isolation Pilot Plant Passive Institutional Controls program and expected plans for the future. (authors)

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DOE OSTI2017

Flow Rate Dependency of Permeability Changes with the Deposition of CSH under the Highly Alkaline Ca-rich Condition - 17175

The formation of calcium silicate hydrate (CSH) around the geological repository is one of the notable factors for the assessment of the radionuclide migration. This study discussed the clogging effect of CSH in flow path with the change in the flow rate based on the flow experiments using micro flow cell. In the experiment, the change of the permeability with the alteration of granite under the highly alkaline and Ca-rich condition was examined. As a result, the permeability in flow path gradually decreased with the periodical increase and decrease of the permeability. In addition, under the condition of the lower flow rate (0.06 ml/h), the permeability became much smaller than those for other conditions of the higher flow rates. These suggest that the clogging of flow path will progress with the local deposition/moving of CSH, in particular, under the condition of the lower flow rate. Retardation effect of radionuclide migration will be expected because it is prospected that flow rate of groundwater is lower than that of this study in the actual condition of repository. (authors)

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DOE OSTI2017

The Reactor Graphite Handling. Problems and Solutions - 17208

The problem of industrial uranium-graphite reactors and NPP units decommissioning is one of the urgent tasks that require efficient and cost-effective solutions. To date, only Russia has already shut down and subject to decommissioning more than 120 objects. World-wide, more than 260,000 ton of irradiated graphite have arisen from the commercial nuclear power operations, research reactors, and defense production reactors. Under the circumstances the problem to develop cost-effective methods of the irradiated graphite management has appeared. Alternative disposal options for the irradiated graphite were discussed for at least 25-30 years. To date, various methods of irradiated graphite reprocessing have been proposed. They are based on the application of different physical and chemical processes, most of which involve incineration of the entire mass of graphite. However, they all have disadvantages, and do not solve the whole complex of problems. The primary goal of the research presented here was to carry out a study aimed at the development of a more simple technology (as compared to the incineration of the entire mass of graphite) that would provide for the graphite purification and allow the possibility of its transfer to the less hazardous waste category followed by the disposal of it in a near-surface repository. To achieve it is proposed to consider two directions, each of them reducing the cost of graphite reprocessing and the waste volume disposed of in the repository. The first alternative is based on the principle of partitioning radionuclides of different nature from the graphite, thus allowing to transfer it into the LLW category and minimize the volume of secondary waste products. The second option to change the reprocessing technology is the idea to use reactor graphite as the starting material to produce stable final waste form compositions intended for the long-term storage of radioactive waste including transuranic elements. (authors)

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DOE OSTI2017

Compact System for Operating and Decommissioned Reactor Used Fuel Management - 17243

Most nuclear power plants in operation today were designed with limited wet storage for their discharged used nuclear fuel (UNF), predicated on the assumption that either recycling or geologic storage would be available within the reactor's lifetime. While that has proven to be the case in some countries such as France with its recycling, it has not for most. In the USA, dry storage of UNF at the plant site at an Independent Spent Fuel Storage Installation (ISFSI) has since the mid-1980s been the primary mode of interim storage so that the plant can off-load its UNF from the pools at the power plants. Initially the space reserved for the ISFSI at a reactor site was based on the assumption that the Department of Energy (DOE) would be transporting that UNF to a geological repository site, such as Yucca Mountain, by 1998. However, due to the delay in the opening of any geological repository the utilities are running out of space on their existing ISFSIs for additional UNF dry storage systems. Since plants prefer to keep their ISFSIs inside the protected area of the plant, options for expanding the ISFSI are becoming limited. At the same time, increasing the capacity of the dry storage systems is also not an option in most cases as the dry storage systems have almost reached their capacity limits with the 37 PWR or 89 BWR fuel assembly canister designs. For plants that have announced plans to shut down, their main priority is to off-load the UNF from the pool to dry storage at an ISFSI so that they can proceed to 'Safe store' or other storage options, and start their decommissioning process as soon as possible. Thus, keeping the size of the ISFSI compact is very important for safety, security and economics. This paper will describe the details of a reduced footprint version of AREVA TN's NUHOMS{sup R} System that addresses the need for maximizing ISFSI capacity. This system not only offers a compact design that stores nearly double the number of UNF assemblies in the same space, it also offers superior radiation protection to the workers and public. In addition, it offers unmatched protection against beyond design basis (BDB) events like the tsunami at Fukushima in Japan, the earthquake at North Anna site, tornadoes and manmade attacks such as the 9/11 attack on the World Trade Center in New York. It also includes significant enhancement for a robust Canister Evolved Safety Program for Aging Management. The utilities in the USA are currently using dry storage systems with canisters of different diameter, different length, and different weights. The fuel assemblies that are stored in these canisters are also different in their fuel types (PWR or BWR),different fuel parameters like burnups, enrichments, and cooling times and different storage orientations in the storage overpacks like horizontal or vertical. The compact version of AREVA TN's NUHOMS{sup R} System offers a 'Universal' Storage Overpack which is designed to accept all these different design dry storage canister available in the industry. As UNF storage continues to drives the industry towards innovations that address the need for consolidated storage, compact ISFSIs and aging management, this new evolution of the AREVA TN NUHOMS{sup R} system offers a significant leap forward in addressing all of these challenges. (authors)

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DOE OSTI2017

Practical Considerations for the Implementation of a World Information Library (WIL) - 17112

The U.S. Department of Energy's (DOE) Waste Isolation Pilot Plant (WIPP) is a deep geologic repository for the permanent disposal of defense-generated transuranic waste. Title 40 Code of Federal Regulations (CFR) Part 194.43, Passive institutional controls, requires that measures be employed, '...to preserve knowledge about the location, design, and contents of the [WIPP] disposal system'. These measures include the, 'Placement of records in the archives and land record systems of local, State, and Federal governments, as well as international archives, that would likely be consulted by individuals in search of unexploited resources'. One possibility for allowing individuals to quickly and easily access information on the location, design, and contents of a disposal system and its relation to unexploited resources is the development of an electronic geographic information system (GIS) that would be used to interpret spatial data. In 2016, R. Patterson proposed such a system, known as the 'World Information Library (WIL)'. Information in this database system would be designed to be publicly accessible through a simple web-based user interface that would focus on spatial databases with categories such as; waste repositories, resources, historical sites, wildlife, infrastructure, medicine, agriculture, and technology. The conceptual WIL would be a compilation of searchable libraries and databases. A category in the WIL would consist of spatial databases created for the purpose of preserving knowledge and providing a means of communicating location information in a GIS-based Land Use Information System (LUIS) user interface. The successful implementation of the WIL would require the robust use of project controls, an effective Quality Assurance program, and the development of international standards in collaboration with industry experts. The purpose of this paper is to identify and discuss known practices and industry standards that could be utilized for the implementation of a WIL as was introduced by Patterson et al. in 'A Conceptual World Information Library (WIL) and Land Use Information System (LUIS)'. (authors)

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DOE OSTI2017

The Waste Factory at the East Tennessee Technology Park - 17424

To successfully support the U.S. DOE Oak Ridge Office of Environmental Management (OREM) in its cleanup at the East Tennessee Technology Park (ETTP), URS - CH2M Oak Ridge LLC (UCOR) recognized the need for the timely disposition of large amounts of waste. To achieve cleanup goals, UCOR implemented the 'Waste Factory' concept. The Waste Factory requires a centralized waste management program, waste management-deployed staff, and numerous outlets for the various waste streams that move the waste from generation to disposition in a timely, compliant manner. Because no waste can be double handled or stored, characterization has to be sound, the disposition path selected based on the characterization, packaging determined, the Waste Acceptance Criteria (WAC) evaluated for the selected repository, and the waste shipped immediately upon generation. Since inception of the UCOR contract, a total of 509,703.24 m{sup 3} (18 million ft{sup 3}) have been dispositioned, totaling more than 60,000 waste loads and 6.5 million safe kilometers (4 million safe miles) driven. The generated waste has been dispositioned onsite and offsite based on the waste characterization and the UCOR waste management hierarchy. Without the timely disposition of the large amounts of waste generated from a massive cleanup project like ETTP, the project could not have achieved the aggressive schedule and performance that has been accomplished to date. The Waste Factory concept can serve as a model for waste management operations across the DOE complex and in other industries where significant quantities of waste must be disposed of. Without the availability of secure, onsite disposal, UCOR would have been forced to send hundreds of millions of pounds of waste by truck to repositories around the country, increasing risks and costs and slowing the overall progress of cleanup. (authors)

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DOE OSTI2017

Deep Borehole Disposal Research in the United States - 17427

The U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) manages a research and development (R and D) program for the disposal of spent nuclear fuel and high-level radioactive waste in the United States. Deep boreholes are being evaluated for disposal of some smaller DOE-managed waste as an alternative to disposal in a mined geologic repository. The deep borehole disposal concept consists of drilling borehole(s) to a total depth of about 5,000 m (16,400 ft - or greater than three miles) into crystalline basement rock, placing waste packages in the lower emplacement zone portion of the borehole, and sealing and plugging the upper portion of the borehole. The required bottom-hole diameter of the borehole(s) depends on the waste package configuration for the reference concept, but ranges from 22 to 43 cm (8.5 to 17 inches). Additional research and development is desired in several important areas for further consideration of deep borehole disposal of radioactive waste. A Deep Borehole Field Test (DBFT) will be conducted by DOE to further assess the viability of deep borehole disposal as an alternative for the disposal of smaller DOE-managed waste forms. The DBFT will be used to evaluate three main areas: 1) evaluation of the capability for drilling and construction of deep, large-diameter boreholes; 2) downhole scientific analyses to assess hydrogeochemical conditions that control waste stability and containment; and 3) engineering activities to assess the viability and safety of waste package emplacement and retrieval in deep boreholes. The DBFT will include first drilling a characterization borehole, followed by an optional field test borehole, to a total depth of about 5,000 m (16,400 feet) into crystalline basement rock in a geologically stable continental location. The initial smaller diameter characterization borehole will be drilled and completed to facilitate downhole scientific testing and analysis. If site conditions are found to be suitable, DOE intends to drill a larger-diameter field test borehole to evaluate emplacement activities using surrogate waste packages. The surrogate waste packages will not contain any radioactive waste, nor will any radioactive waste be used in these tests. One of the advantages of deep borehole disposal is that it would give DOE the flexibility to consider options for disposal of smaller waste forms. Deep borehole disposal also potentially offers earlier disposal of some wastes than might be possible in a mined repository, and could reduce costs associated with projected treatments of some wastes. In order to accomplish the DBFT, DOE will require the support of multiple national laboratories, along with a site contractor to acquire a suitable test site and to manage the site and the required testing activities. DOE issued a competitive solicitation via a request for proposal (RFP) in July, 2015, which requested proposals to provide a test site and site management services. The contract was awarded in January, 2016, to a team led by Battelle Memorial Institute which included a proposed test site in Pierce County, North Dakota. After efforts to acquire both the initial test site in North Dakota and an alternative proposed site in Spink County, South Dakota were unsuccessful, activities were suspended. The experiences in Pierce County, ND and Spink County, SD highlighted the importance of public engagement and support for the DBFT, and that relevant levels of government and other public stakeholders should be involved from the beginning. Using the lessons learned in North and South Dakota, DOE issued a new RFP on August 22, 2016 which emphasized local, state, and tribal (if applicable) government engagement, as well as public and other stakeholder involvement ahead of proposal submittals and throughout the contract execution phases. The new RFP allowed for multiple initial awards and multiple phases of contract execution. In December, 2016, DOE announced the selection of four companies - AECOM, ENERCON, RE/SPEC, and TerranearPMC - to begin exploring the possibility of conducting a DBFT under the terms of the new RFP. Ultimately, only one site will be chosen for the field test. (authors)

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DOE OSTI2017

Recent Developments in Radioactive Waste Management in the Federal Republic of Germany - 17450

In 2013 the Federal Parliament and the Federal Council passed the Act Concerning the Search and Selection of a Site for a Repository for Heat-Generating Radioactive Waste and the Amendment of Other Acts (Site Selection Act). It was a restart for the search for a repository for especially heat-generating waste in Germany. As a first step in May 2014 the Commission 'Storage of High-Level Radioactive Waste' started their work. Their aim was to debate and clarify the principle questions concerning the management of heat-generating radioactive waste. As well as to evaluate and determine the fundamentals for the Site Selection Process e.g. safety requirements, exclusion criteria, minimum requirements and consideration criteria for the site selection. They also evaluated the organizational and procedural arrangements in the scope of waste management. As a first act to implement the suggestions of this Commission the organizational structures in waste management were greatly altered. This reorganization not only had a significant influence on the waste management of high-active, but also on the management of low- and intermediate-active waste management in the Federal Republic of Germany. In 2015 a second Commission started its work focusing on the funding of waste management in Germany. The results of this Commission were presented in April 2016. At the moment the suggestions of this Commission are being implemented, changing the financial responsibilities for waste management in the Federal Republic of Germany. The presentation will focus on changes and the developments of radioactive waste management in Germany and talk about the ongoing process in implementing the suggestions of the two Commissions. (authors)

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DOE OSTI2017

A Study Of Geological Disposal Facility (GDF) Ageing Phenomena In Bentonite - 17605

Since the 1950's, the UK's nuclear activity in energy, defence and medicine has led to the accumulation of a large amount of higher activity radioactive waste (HAW). Most of nuclear waste arises in a wide variety of physical and chemical forms with different radioisotopes and loadings. Due to complexity in HAW, it is imperative that there be a rigorous management system that ensures population and environmental safety over extended periods. The concept of Deep Geological Disposal of waste in a stable geological formation is, nowadays, an internationally accepted solution as the surest and the most viable long term method for management. Any future Geological Disposal Facility (GDF) will comprise, among other artificial and natural engineered barriers, a barrier immediately around the waste containers compounded by a sealing material. The long-term stability of a GDF will be determined by the design and construction of the engineered barriers, and the mechanical, hydraulic and geochemical interactions that may happen between the different engineered components as a result of combined processes such as the heat produced by radioactive decay and the ion water composition of the surrounding host rock. Bentonite clay has been selected as the most promising buffer and backfill material in many disposal programmes because of its swelling capacity, low permeability, plasticity and cation exchange capacity. In the nuclear scope, bentonite testing at different scale-lengths (laboratory and field scale) has been performed and implemented in order to examine the thermo-hydro-mechanical (THM) and thermo-hydro-geochemical (THG) processes that may happen in the engineered barriers and geological settings associated with a GDF. Two major European field scale tests have been performed to better understand bentonite performance and behaviour immediately after HAW emplacements. The Full-Scale Engineered Barriers Experiment (FEBEX) at the Grimsel Test Sit (Switzerland) and the Alternative Buffer Material (ABM) Project at the Aespoe Hard Rock Laboratory (Sweden). The Febex Experiment was a pilot project designed for 'field research' of HAW repository hosted in crystalline rock where surrogate HAW canisters were placed horizontally in drifts, surrounded by compacted blocks of bentonite, sealed by a concrete plug and maintaining a constant canister surface temperature of 100 deg. C. After five years of operation, a partial dismantling was carried out in 2002. Heater processes occurring in the surrounding bentonite. The dismantling of the heater no. 2 (after 18 years of operation) started in late 2014 and is planned to be finished at the end of 2016 (termed FEBEX-DP). The FEBEX Project utilised bentonite from the Cortijo de Archidona deposit (Almeria, Spain). This 'Febex' bentonite is sourced from calcium-magnesium montmorillonite with good properties in relation to swelling capacity, permeability, adsorption capacity, and thermal conductivity. The ABM Project is a large-scale programme that has been running since 2006. The programme is designed to study a suite of possible buffer materials and correlate the physical and chemical properties of eleven different types of bentonites to their fundamental mineralogical properties. The ABM test consists in three 'packages' containing a central heater tube surrounded by different pre-compacted bentonite blocks and instrumentation. Each 'package' was installed in separate boreholes. The target temperature for the surface of the heated tubes in contact with the bentonite was 130 deg. C. Both FEBEX and ABM test the bentonite was saturated with water percolating from surrounding host rock (FEBEX and ABM) and also via artificial supply systems (ABM), being the water itself acting as a long term corrosion agent. For both experiments and emerging interest is in the interaction of bentonite with metallic components, under geological disposal conditions are not considered to be thermodynamically stable and hence arising corrosion products could interact with the buffer. These interactions depend on various factors such as the redox potential, temperature, pH, solute concentration and buffer material characteristics. During the operational phases of the FEBEX and ABM and after its 'closure' a significant volume of air was found to be retained in the bentonite. Under these conditions a period of initial aerobic corrosion by the air in the bentonite barrier would have occurred followed by establishment of anaerobic corrosion would be the dominant process ongoing. The anaerobic corrosion began when the bentonite became water saturated. Under these conditions, the water acted as the major oxidant. The present project focuses on the characterization of samples arising from the FEBEX and ABM Projects as well as performing additional independent experiments designed to complement these prior studies. Characterization will focus mainly on corrosion related alteration combined with thermal effects. Through detailed sample characterization, the intention is to achieve a greater knowledge regarding the changes to the bentonite under the experimental conditions to which it was exposed, and plausible alterations in its THM and THG properties occurring during the experiment due to blended effects of temperature,water content, engineering structural components and solutes are also to be measured. The fulfilment of a large-scale 'in-situ' test is intricate, time-consuming and with an abundance of external variables that are not always well controlled. For this reason, the University of Bristol is performing a series of well constrained time-resolved laboratory tests which bentonite is being subjected to different possible scenarios that may happen whilst in a HAW repository. In virtue of this experiment, the UoB aims to understand the illitisation and cementation process and their effects on bentonite safety properties. (authors)

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DOE OSTI2015

Demonstration of friction stir Welding (FSW) Technology for packaging of used nuclear fuel - 15025

With uncertainty surrounding the disposal of US Used Nuclear Fuel (UNF), the nuclear industry is making an effort to identify an interim storage solution. One of the alternatives is to simply manage UNF on-site until a permanent repository can be established - extended storage. In an effort to provide a technical basis for extended storage, the Department of Energy (DOE) performed a gap analysis identifying several issues that need to be addressed. One of the issues relates to the performance of the UNF dry storage container (DSC). DSC container welds (fabricated using conventional fusion welding processes) are sensitive to Stress Corrosion Cracking (SCC), and their extended-term performance is questionable. Fluor, the Pacific Northwest National Laboratory (PNNL) and the Savannah River National Laboratory (SRNL) believe that Friction Stir Welding (FSW) technology may be able to address the DSC issue. The Fluor team collaborated on a project to demonstrate feasibility of FSW for DSC packaging (closure-welding). This paper reports on a 2-year project in which FSW process conditions were established (for DSC container materials and dimensions) and used in the preparation of mechanical and corrosion specimens for testing and evaluation. In addition to FSW test specimens, fusion welded specimens were prepared and tested in like manner for comparative purposes. The results show excellent mechanical properties (meeting construction code requirements) and significant improvement in corrosion performance - over the fusion welded specimens. It is believed these findings will be of significant value to efforts underway to provide a technical basis for DSC extended storage performance. FSW technology appears to be an acceptable alternative or candidate process, capable of addressing the SCC sensitivity issues associated with fusion-welded DSC fabrication welds. In addition, as an alternative to on-site, extended storage of UNF, the DOE has proposed the development and construction of a Consolidated Interim Storage Facility (CISF). UNF would be accepted from the many, local sites for consolidation at one or more CISFs until a permanent repository could be established. A recent CISF study indicates that UNF currently packaged into DSCs will likely be re-packaged into containers suitable for permanent storage. The study also indicates that re-packaging is best performed under water in a pool. FSW technology has several specific advantages for this application, 1) because it utilizes machine-tool equipment, a welding station can easily perform a cutting operation, and 2) FSW has been successfully demonstrated capable of joining container materials under water. (authors)

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DOE OSTI2015

Technetium incorporation in glass for the Hanford Tank Waste treatment and immobilization plant - 15066

A priority of the United States Department of Energy (U.S. DOE) is to dispose of nuclear wastes accumulated in 177 underground tanks at the Hanford Nuclear Reservation in eastern Washington State. These nuclear wastes date from the Manhattan Project of World War II and from plutonium production during the Cold War. The DOE plans to separate high-level radioactive wastes from low activity wastes and to treat each of the waste streams by vitrification (immobilization of the nuclides in glass) for disposal. The immobilized low-activity waste will be disposed of here at Hanford and the immobilized high-level waste at the national geologic repository. Included in the inventory of highly radioactive wastes is large volumes of {sup 99}Tc (∼9 x 10 E2 TBq or ∼2.5 x 10{sup 4} Ci or ∼1500 kg). A problem facing safe disposal of Tc-bearing wastes is the processing of waste feed into in a chemically durable waste form. Technetium incorporates poorly into silicate glass in traditional glass melting. It readily evaporates during melting of glass feeds and out of the molten glass, leading to a spectrum of high-to-low retention (ca. 20 to 80%) in the cooled glass product. DOE-ORP currently has a program at Pacific Northwest National Laboratory (PNNL), in the Department of Materials Science and Engineering at Rutgers University and in the School of Mechanical and Materials Engineering at Washington State University that seeks to understand aspects of Tc retention by means of studying Tc partitioning, molten salt formation, volatilization pathways, and cold cap chemistry. Another problem involves the stability of Tc in glass in both the national geologic repository and on-site disposal after it has been immobilized. The major environmental concern with {sup 99}Tc is its high mobility in addition to a long half-life (2.1 x 10{sup 5} yrs). The pertechnetate ion (TcO{sup 4-}) is highly soluble in water and does not adsorb well onto the surface of minerals and so migrates nearly at the same velocity as groundwater. Long-term corrosion of glass waste forms is an area of current interest to the DOE, but attention to the release of Tc from glass has been little explored. It is expected that the release of Tc from glass should be highly dependent on the local glass structure as well as the chemistry of the surrounding environment, including groundwater pH. Though the speciation of Tc in glass has been previously studied, and the Tc species present in waste glass have been previously reported, environmental Tc release mechanisms are poorly understood. The recent advances in Tc chemistry that have given rise to an understanding of incorporation in the glass giving rise to significantly higher single-pass retention during vitrification are presented. Additionally, possible changes to the baseline flowsheet that allow for relatively minor volumes of Tc reporting to secondary waste treatment will be discussed. (authors)

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

The fourth decade of underground research at the Grimsel Test Site - what we have learned and where we go from here - 15075

At the Grimsel Test Site, NAGRA and its partners have celebrated thirty years' worth of essential contributions to the characterization of crystalline host rock, and the development and confirmation of safe geological disposal concepts. In operation since 1984, the underground research laboratory in the Swiss Alps shall not accept radioactive waste but can host experiments with radiotracers/radionuclides under strictly controlled conditions. The strong drive for international cooperation at the Grimsel Test Site is reflected in the more than two dozen organizations and research institutes from twelve countries and the European Union which have participated in the research program to date. The program has been designed dynamically to always mirror the current requirements of national programs for the long-term management of radioactive waste, as well as attempt to anticipate the next steps. Early work featured the development of methodologies for formation characterization. Then followed a period concentrating on performance assessment issues, before radionuclide migration and barriers were investigated intensely. The past 15 years have been dominated by engineering demonstration experiments. The research conducted at the Grimsel Test Site has complemented laboratory experiments, surface investigations and numerical modelling studies through the bridging of different scales and drawing off links between laboratory setups and natural analogues. The crystalline rock formations of the Aar Massif have provided an opportunity to do experiments in environments similar to those of future repositories. The facility also serves as a platform for continual knowledge transfer between generations of scientists and fosters synergies among disciplines. However, considering the ambitious schedule for repository implementation in several continuous, one may ask whether the era of generic underground rock laboratories is approaching its end. (authors)

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

Performance assessment modernization at WIPP - 15077

To aid in WIPP PA calculations, the US DOE has undertaken a broad modernization effort that includes: 1) updating the hardware and operating system used; 2) moving the current PA (performance assessment) codes to the new system; and 3) adding a database to store results essential for the analysis of assessments; 4) integrating run-control scripts with a web site to facilitate monitoring of performance assessment calculations; and 5) adding new capabilities via implementation of the PFLOTRAN code to replace multiple codes currently in use. The majority of codes were migrated from an Open VMS/Alpha processor system to a Unix/Intel processor system. Most codes were upgraded to use the dynamic allocation features of Fortran 95 and double precision variables. A few codes were migrated from Linux/Intel systems to the Unix/Intel systems and these code required nearly no changes to the source code. All of the migrated codes were re-qualified under Sandia's WIPP QA program. The implementation of some of the conceptual models using PFLOTRAN is being driven by the expectation that changes in the configuration of the repository will require a 3-dimensional representation of the mined areas. Symmetry in the current configuration of the repository made a 2-D representation adequate. PFLOTRAN is an open-source, state-of-the-art massively parallel subsurface flow and reactive transport code. A version of this code is being developed specifically for WIPP PA in order to maintain the capabilities of the current flow and transport codes, as well as add the capabilities of a parallel code that will treat additional processes, including: 1) miscible multi-phase flow; and 2) heat transfer. PFLOTRAN will replace the two codes (i.e. BRAGFLO and NUTS) currently used for flow and transport calculations, simplifying the PA code structure and speeding up calculations by simultaneously solving for flow and transport with a single code. To ensure that PFLOTRAN reproduces the results of the current code suite, existing test cases have been used for verification purposes. Additional test cases have been developed to establish the new capabilities for WIPP PA. The result of the modernization effort will be a state-of-the-art subsurface flow and transport capability that will serve WIPP PA into the future. (authors)

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

Effectiveness of actinide management in fast reactors over a finite recycling period - 15123

Long-lived nuclear waste primarily consists of transuranic (TRU) isotopes. Full recycling of TRU isotopes can, in theory, lead to a reduction in repository radiotoxicity to reference levels in as little as ∼500 years provided reprocessing and fuel fabrication losses are limited. However, over a limited time frame, the radiotoxicity of the 'final' core can dominate over reprocessing losses, leading to a much lower reduction in radiotoxicity compared to that achievable at equilibrium. In this paper, the fuel cycle code ORION is used to model the recycle of light water reactor (LWR)-produced TRUs in sodium-cooled fast reactors (SFRs). With full actinide recycling, at least 6 generations of SFRs are required in a gradual phase-out of nuclear power to achieve transmutation performance approaching the theoretical equilibrium performance. Break-even SFRs are much less effective from a point of view of reducing waste radiotoxicity. Thorium (Th)-burner fuel cycles are found to result in slightly lower radiotoxicity than U-burner fuel cycles for the same number of generations. Closed break-even Th fuel cycles require ∼3 generations of operation before their waste radiotoxicity benefits result in decay to the reference level in ∼1000 years, provided that Pa is recycled. While this is a very long time frame, it is substantially better than the time frame required by other strategies. The integrated decay heat over the scenario time frame is substantially higher for break-even Th-SFRs than break-even U-SFRs as a result of much higher {sup 90}Sr production. As decay heat affects repository sizing, this weakens the argument for the Th cycle. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗