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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)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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)

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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)

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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)

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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↗
DOE OSTI2015

Andra's strategy and approach for operational safety of French underground facility Cigeo - 15127

According to the French Act on radioactive waste of 28 June 2006, Andra shall design a reversible repository. In response to this demand, Andra is developing the industrial project namely 'Cigeo', a reversible geological disposal facility for HLW and ILW located in Meuse/Haute-Marne. The previous safety cases of 2005, 2009 and 2012 give a sound basis to assess Cigeo's safety both for the operational and post-closure periods. Regarding operational safety, in this new stage of the project, the challenging issues for the preparation of the safety case are the following: - To adapt the various regulatory frameworks (nuclear and non-nuclear) to the underground facility - To design the underground facility and handling equipment's to reduce at first risks in the underground facility as such radioactive particles release, gas explosion, fire and co-activity and limit their consequence. - To establish waste acceptance criteria to be fulfilled by waste producers in line with operational safety and design options. The objective of the paper is to present the experience of Andra in operation safety of the underground facility of Cigeo. Operational safety for Cigeo is assessed as much as possible according to the existing French regulations and guidance for nuclear facilities. Indeed, compared to existing nuclear facilities, a geological repository is a specific case due to the existence of underground facilities with particularities such as: - ventilation management in a large scale underground facility, - fire risk management. As an input for the industrial design of Cigeo for operation, Andra defined a fire guidance and a specific reference safety guidance, based on the adaptation of the French nuclear regulatory texts for and when feasible to mines and tunnels regulatory texts and on its own experience gained through safety operational reports since 2005 and their reviews by the Safety Authority. - underground co-activity of nuclear operation with construction works. According to the common 'defense-in depth' principle for nuclear facilities, these guidelines for the industrial design of Cigeo for operation consider the successive levels of defense of: - Preventing the risk and its spreading by design and construction; - Detecting incidents and early fighting by design and operation dispositions (detection and control systems, survey procedure); - Controlling incidents and limiting of their consequences by safeguarding systems, specific procedures. At that stage, design options for Cigeo constitute an adequate response to the operational safety functions and requirements: - Two levels of containment are set up: the first one is ensured by the waste package, the second one is ensured from the surface facility to the disposal vaults by the transfer cask during the transfer of waste packages in the disposal vaults by the disposal vault itself and a nuclear ventilation with HEPA filter; - Favorable options for fire risk prevention and mitigation: transport system minimizing heat load (cable car for the transfer from the surface, electric transfer engines), twin access tunnels to ease fire-fighting and evacuation...; - Permanent complete physical separation of nuclear zone and construction zone (including ventilation systems). In practice, operation safety assessment of Cigeo, designed following the aforesaid reference safety guidelines, follows a common risk analysis (internal aggressions such as auxiliary losses, chemical and toxic risks, fall or collision, fire, human factors and external aggressions such as earthquake, air traffic, flooding) with the following steps: - Identification of the major risks and the design basis scenarios that have to be considered for dimensioning of the installations and the safety devices; - On the basis of these different situations, those who will have to be quantified are identified; one or more representative and envelop situations are chosen in order to evaluate robustness of the installation. Quantitative evaluation of these situations permits to: - identify the consequences on man, environment and installation, and in particular to gain confidence in the maintaining of the safety functions; - identify or define the technic and/or organisational dispositions classified as 'important elements for protection' (IEP); - Situations beyond dimensioning permit to verify the robustness of the installation. They permit, on one hand to identify possible complementary material or organisational dispositions; on the other hand, to define the 'hard core' i.e. structures, systems and components that permit to maintain the confinement function of radioactive substance. (authors)

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DOE OSTI2015

The decommissioning imperative - motivating progress in us used fuel management - 15185

When the Yucca Mountain project was terminated in 2010 there was approximately 3500 metric tons of used fuel stored at Independent Spent Fuel Storage Installations (ISFSIs) at which there was no operating reactor (10 shutdown plant sites and the GE Morris facility). By 2015, four additional sites with five reactors will have shut down, causing the amount of used fuel 'stranded' at sites where there is no operating reactor to nearly double to approximately 6900 metric tons which will be stored at a total of 15 sites. At each of these sites, once the reactors are decommissioned, the only thing preventing the land from being returned to communities for other uses is the remaining presence of the used fuel. The cost of maintaining 15 stand-alone ISFSIs is born by electric ratepayers and taxpayers. Going forward, the costs of maintaining stand-alone ISFSIs is certain to grow even if the number of stand-alone ISFSIs does not grow beyond the already announced shutdown of the Oyster Creek reactor in 2019 (which would increase the amount of stranded used fuel by another 10% and create a 16. stand-alone ISFSI). One reason for this cost escalation will be the need for ISFSI owners to renew storage licenses, and implement aging management programs to support these renewals, due to the continued delay in the federal program for removing the used fuel. Storage licenses at 10 of 16 stand-alone ISFSIs will expire by 2021, which is the earliest projected date at which the Department of Energy (DOE) has indicated it could have a pilot interim storage facility available - and nearly 3 decades before DOE projects a repository will be available. The burden and cost of continuing to renew storage licenses - and manage aging storage facilities - at 16 stand-alone sites instead of moving stranded used fuel to a single consolidated storage location or repository will likely be significant. Industry resources are increasingly being channeled into dry storage aging management activities as public and regulator interest in potential long-term degradation mechanism continues to grow. Whether the ultimate destination for used fuel is Yucca Mountain or some other location, the current stalemate preventing progress in the federal program to move it must be broken - and the increasing prominence of the decommissioned reactor quandary may very well be a pivotal factor in breaking that stalemate. (authors)

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DOE OSTI2015

Transport of French PWR steam generators to very low level waste disposal - 15203

When considering the complexity of IAEA nuclear transport regulations, the transport of a Steam Generator (SG) in one un-packaged block from a dismantling plant to a nuclear waste disposal is a challenge. The choices made, the difficulties encountered and the feedback that has been learnt from it are described. The disposal of decommissioning waste requires taking into account not only the constraints of acceptance of the repository but also the specific constraints of transportation from the dismantling plant to the final disposal site. In France, these criteria and regulations are respectively under the supervision of the French National Radioactive Waste Management Agency (ANDRA) and the French Nuclear Safety Authority (ASN). In the course of the dismantling of the EDF PWR plant Chooz A, the four decontaminated SGs were shipped to the ANDRA Very Low Level Waste (VLLW) repository. After removal, the SGs were decontaminated, allowing them to attain the criteria of VLLW and thus be accepted by the ANDRA. However, to transport these decontaminated SGs as single un-packaged blocks, it was necessary to ensure compliance with the Surface Contaminated Object SCO-I regulation criteria. For this, transport safety files were written and surface contamination and radiation criteria checked. One of the four SGs had to be decontaminated again and another one had to be transported under special arrangement. Due to the exceptional nature of the transportation, specialized means of transport and stowage were also studied. (authors)

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DOE OSTI2015

Outcomes of the full scale seal experiment: a seal industrial prototype for the Cigeo project - 15207

The Full Scale Seal (FSS) Experiment is one of various experiments implemented by Andra, within the frame of the French HLW and IL-LLW Deep Geological Repository Project (aka Cigeo) development, to demonstrate the technical feasibility and check the performance of the seals which will be constructed in the shafts, ramps, drifts and disposal vaults, at time of Cigeo underground facilities progressive closure. The FSS experimental objectives, principles and rationale were presented at WM'2013 (Paper 1306): FSS is a technological demonstrator built on surface, inside a concrete made horizontal drift model (also called the 'test box'). FSS is part of the European DOPAS (Demonstration of Plugs and Seals) [2] project, under the umbrella of the Implementing Geological Disposal - Technology Platform (IGD-TP) [3] project. FSS construction activities started in July 2012 and were completed in September 2014. The construction compliance assessment work activities are now ongoing (and will be continued until mid-2015). This paper presents the construction and monitoring story of the FSS components, starting with the construction of a first containment wall based on a low pH self-compacting concrete (SCC), followed by the swelling clay core backfilling sequence and finishing with the erection of a second containment wall based on a low pH shotcrete. The FSS experiment also enabled to qualify the commissioning methods applicable to an industrial sealing work at time of underground operations in the Repository. These commissioning methods are described and the results discussed. The equipment and methods used for the purpose of seal component construction and monitoring are presented. The physical characteristics obtained for each component of the seal are detailed and the compliance of measured values with the initial requirements is commented. The post-mortem work planned (e.g. coring and wire sawing of low pH concrete containment walls, use of penetrometer for the swelling clay core...) is also explained. FSS is a first of a kind experiment, by its size, by the technical choices made. The first technical outcomes were presented to the French Nuclear Authority and to its Technical Support Organization. The successful implementation of the work carried out in the FSS drift model was positively assessed by the evaluators and deemed representative of the real underground conditions (industrial feasibility is proven). This article concludes with a critical analysis of the methods and equipment used, of the results obtained and provides some perspectives of improvement for the construction of the first Cigeo industrial prototype seals scheduled around 2030 in ramps and horizontal drifts (provided the Cigeo construction is authorized in late 2020, and following the licensing request filing scheduled in late 2017). (authors)

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

The transportation factor in site evaluations for spent nuclear fuel storage and disposal facilities - 15460

New facility siting studies will be necessary to fulfill the Blue Ribbon Commission (BRC) on America's Nuclear Future 2012 recommendation to develop consolidated interim storage (CIS) for spent nuclear fuel in the next decade. Future site evaluations for used nuclear fuel (UNF) storage and disposal facilities must address transportation requirements, especially access to mainline railroads, early in the site selection process. The U.S. Department of Energy (DOE) in 2013 proposed a strategy for managing spent nuclear fuel in accordance with the BRC recommendations by establishing a pilot interim storage facility by 2021 and a consolidated interim storage facility by 2025. The authors demonstrate that currently available geographic information systems and already adopted siting criteria (defining favorable and unfavorable conditions) allow the identification of the transportation portion of a storage facility siting process to begin without delay. Ready access to the mainline rail system is absolutely critical since existing rail access, or easily constructed access, is a highly favorable condition for successful siting. Despite widespread rail abandonments over the past three decades, vast areas of the United States retain mainline rail service or could connect to existing mainlines by building new rail spurs of less than 25-50 miles. This precondition for transportation planning is critical. Planners need to start with the beginning - availability of rail transportation to and from the interim storage facility/facilities. A second step in transportation screening would be to evaluate those potential site locations within 50 miles of existing mainline railroads using other appropriate variables. This effort could begin with the application of the 1984 repository siting guidelines for transportation (10 CFR 960.5-2-7)[1]. Repository surface facilities have essentially the same transportation requirements as storage facilities. The guidelines in 10 CFR 960 provide a readily available template for evaluating sites based on transportation risks, impacts and costs. Initial site screening, after considering proximity to mainline railroads as the precursory step, should evaluate: 1) proximity to highly populated areas; 2) proximity to interstate highways; 3) proximity to environmentally sensitive areas and lands protected by federal, state, or tribal governments; and 4) proximity to military installations, hazardous materials facilities, or other facilities that might impact transportation safety and security.Several challenges exist which prevent effective transportation planning. The diversity of canister and cask types and SNF characteristics and conditions, the increasingly limited rail access at the shipping sites and the ease with which public attention can be marshalled against a waste management program, all suggest that future efforts need to identify potential storage sites that reduce transportation impacts and costs by reducing near-site transportation construction and reduce total system transportation impacts. This can most easily be accomplished by reducing shipping distances in a process that is easily defined and implemented. The paper concludes that the DOE should develop a site screening and evaluation program with transportation as a central consideration. (authors)

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

Engineering management challenges of the CIGeO Project - 15576

After being involved for years in the conceptual studies on the French Deep Geological Repository's (DGR) design for ANDRA, INGeROP (www.ingerop.com) is carrying out since 2012 the global project management and the technical integration of the preliminary, basic and detailed design of the Cigeo project, within the framework of GAIYA consortium. The article presents some particular organizational aspects that turned out to be more challenging than foreseen in the course of our activities. Starting from the presentation of real examples, some lessons learned are derived as well as practices of interest in solving analogue issues in similar projects, with focus on their application to the engineering phase of geological repositories in different countries. Finally, the paper concludes with some avenues for reflection on the 'reversibility' - including 'retrievability' - requirement asked for by the French public and government and its consequences on the engineering choices, on the expected lifetime of components and structures and on the (increased) complexity of CIGeO systems. (authors)

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

Applying fast calorimetry analysis on spent fuel calorimeter - 15620

Swedish Nuclear Fuel and Waste Management Company (SKB) is responsible for the back-end of the Swedish nuclear industry. This includes transport of spent fuel by boat from the power plants to the Central Interim Storage Facility for Spent Fuel (Clab), and in the future, encapsulation of the spent fuel and further transportation to a geological repository. In order to ensure safety for the planned geological repository, the decay heat has to be well known. For this purpose the SKB has performed calorimetric measurements on over 200 pressurized water reactor (PWR) and boiling water reactor (BWR) assemblies since 2003. The results from earlier analyses have been used to improve computer codes predicting heat as well as to acquire deeper knowledge about the fuel inventory. The work complements research being done as part of the Next Generation of Safeguards Initiative - Spent Fuel project, whose purpose is to develop and test nondestructive assay (NDA) technologies to improve safeguards measurements of spent nuclear fuel. The project is working to accomplish multiple goals, one of which is to determine the heat content of spent fuel assemblies. To shorten the measurement time, the temperature increase method was used instead of the equilibrium temperature method. In the temperature increase method, the temperature inside of the calorimeter is lowered below the surrounding water temperature before the measurement, and the temperature is allowed to increase due to the higher surrounding water temperature during the measurement. The rate of the temperature increase when the temperature is the same as the surrounding temperature is then compared to a calibration curve where an electric heater has been used in the same way. The price for the higher throughput is less accuracy and dependency of fuel geometry since the heat flow will be different for different fuel types, especially between PWR and BWR assemblies. Recently at Los Alamos National Laboratory, more sophisticated analysis methods have been considered for the use of calorimetry for treaty verification. These methods aim to predict the equilibrium temperature based on early data and therefore be able to shorten the measurement time while maintaining good accuracy and eliminate the dependency on fuel geometry. Even though the techniques show promise for future calorimetry of spent fuel, certain design features of the current system have been identified to interfere with the accuracy of equilibrium measurements in general such as changing pool temperatures and heat leakage during calibration. (authors)

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

The Bukov underground research facility as part of the Czech Republic's DGR development programme - 15644

The Czech Republic's deep geological repository (DGR) policy sets out a number of important milestones including the selection of candidate sites by 2018 and the selection of a final site by 2025. The final commissioning of the Czech DGR is planned for 2065. In 2010 a screening process was completed employing geological and other criteria that resulted in the identification of 7 potentially suitable sites, all situated in crystalline - granite host rock environments. The next step of the process will result in a reduction both in the number of sites and the surface area of remaining sites based on preliminary safety evaluations and socioeconomic, political and environmental criteria. In order to better understand the behaviour of crystalline host rock environments for DGR safety case purposes, the Radioactive Waste Repository Authority (RAWRA) made decided to enhance the extent of in-situ measurements and research at a depth corresponding to that of the future DGR. In the second half of 2013, RAWRA commenced the construction of the Bukov Underground Research Facility (Bukov URF) in the southern section of the Rozna uranium mine at a depth of 600 m below the earth's surface in a geologically well-mapped area of a Moldanubic crystalline rock formation which is composed principally of migmatised gneisses, migmatites and granulites. The Bukov URF will be, first and foremost, a research facility at which work will be conducted focusing on the taking of geophysical measurements, petrographic, mineralogic, geochemical research and the in-situ determination of the physical - mechanical, geochemical and hydrogeological properties of the host rock massif. The research results will form input data which will be used to both support and improve the safety of the future DGR. (authors)

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DOE OSTI2016

Cigeo, The French Geological Disposal Facility Project: Preliminary Design (APS) Phase - 16038

The Cigeo project has been in the works for 25 years. Numerous studies have been conducted, increasingly associated with the Meuse/Haute-Marne site, with further specific research thanks to direct access to the Callovo-Oxfordian clay formation from the underground laboratory of Bure-Saudron. These studies and research initially aimed to demonstrate the feasibility of the repository. They also helped gain a high level of understanding of phenomena to support design studies and demonstrate safety. Transition to the industrial phase began with the development of a plan for delivering waste to the facility for disposal. The plan introduced sequencing for the various types of waste to be disposed of, and was optimised to determine the size of inspection, transfer and handling facilities. In describing the life of the repository and therefore the vision for its operation, it has become obvious that our generation should not impose choices on future generations. We must provide them with reference technical solutions, with the financial resources to implement them. It is also our duty to begin the construction and initial operating phases. However, because the facility will operate over several generations, we must leave a degree of flexibility so that they may reassess the options that we define and adopt their own solutions, as necessary. They will also benefit from operational experience gathered as facility operations develop. This is the context in which the preliminary design phase is being finalised in preparation for the detailed design phase, with the aim of gradual commissioning during the latter part of the next decade. (authors)

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DOE OSTI2016

Applicability of Wireless Power Transfer for Monitoring Technology of Radioactive Waste Geological Disposal - 16156

To assure the engineered barrier of a high-level radioactive waste repository functions correctly, long-term monitoring inside the backfilled deposition holes and tunnels is planned. Cables are used to power monitoring equipment and send signals however complete isolation will not be possible. Previously developed monitoring systems send information wirelessly and used batteries. Due to the limited battery life, these systems cannot be used for long-term monitoring. In this paper a wireless power transfer method based on magnetic resonant coupling is presented. The necessary transmitter and receiver coils have been developed. The performance of the individual coils has been tested. The performance of the system of transmitter and receiver with both air and bentonite as medium was tested. The results showed that the system with bentonite performed similarly well as the system with air. Wireless power transfer based on magnetic resonant coupling is therefore a feasible method for high-level radioactive waste repository monitoring applications which could in principle work for a period of 100 year or longer. (authors)

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DOE OSTI2016

An Evaluation of Standardized Canisters in the Waste Management System - 16317

Development of a standardized canister system represents an opportunity to develop an integrated approach to addressing storage, transportation, and disposal issues in the waste management system. However, regardless of timing and method, deployment of such a system would have the potential to cause significant system-wide impacts. This evaluation expands upon previous standardized-canister system evaluations and compares continued loading of dual-purpose canisters (i.e., the status quo) with loading of standardized canister systems in the near term, that is, before repository requirements are known and/or before operating reactors shut down. This evaluation quantitatively compares order-of-magnitude costs and logistics for different standardization scenarios with those of status quo scenarios, provides insight into quantifiable impacts of loading standardized canister systems in the near term, tests system-level analysis tools and associated input, and identifies scenarios for further analysis. This is a technical paper that does not take into account the contractual limitations under the Standard Contract (10 CFR Part 961). As work on assessing standardized canister system scenarios has progressed, new data for at-reactor and interim storage facility (ISF) operations have been collected and/or generated to provide more realism at the system level, and past scenarios were re-evaluated with this new information. In addition, new scenarios were developed and analyzed that focused on (1) bare fuel transported from reactors to an ISF in re-usable bare fuel casks and deferring deployment of a standardized canister system to the ISF, (2) alternative acceptance strategies based on conclusions from previous systems architecture studies, and (3) the impacts of having to accommodate spent fuel in differently sized canisters for different potential repository geologies. Results of this evaluation indicate that shifting the loading of standardized canisters from the reactor to the ISF does not result in a significant change in total system cost. Results also show that some alternative acceptance strategies could reduce the number of years that spent nuclear fuel (SNF) stays on-site at reactors for cases involving standardized canisters (consistent with previous results that looked at dual-purpose canisters). Finally, incorporating the new at-reactor loading data of small standardized canisters shows a significant system-wide cost reduction when compared to previous estimates of at-reactor loading of small canisters. (authors)

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