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At least 1,333 records · Page 74
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DOE OSTI2013

Preliminary Safety Analysis of the Gorleben Site: Geological Database - 13300

The Gorleben salt dome is 4 km wide and nearly 15 km long. It is composed of different salt rock types of the Zechstein (Upper Permian) series and extends to the Zechstein basis in a depth of more than 3 km. In the course of the salt dome formation the salt was moved several kilometers. During the uplift of the salt the initially plane-bedded strata of the Zechstein series were extensively folded. In this process anhydrite as a competent layer was broken to isolated blocks. In the core of the salt dome the Hauptsalz, which is characterized by a particularly high creeping capacity, forms a homogeneous halite body with a volume of several cubic kilometres. The Hauptsalz contains gaseous and liquid hydrocarbons in separated zones of decimeter to meter dimensions. The overall hydrocarbon content is far below 0.01 %. At the flanks the salt dome consists of salt rocks with lower creeping capacities. Brine reservoirs with fluid volumes in the range of liters to hundreds of cubic meters exist in certain regions of this part of the salt dome. The water content of the Hauptsalz is below 0.02 %. Interconnected pores do not exist in the salt rock outside of fluid bearing or fractured areas, i.e. the salt rock is impermeable. The exploration of the Gorleben site as a potential site for a HLW-repository started in 1979 and is still in progress. To date no scientific findings contest the suitability of the site for a safe HLW-repository. (authors)

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DOE OSTI2013

Dry Storage of Research Reactor Spent Nuclear Fuel - 13321

Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. The initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage to dry storage requires integration with current facility operations, and selection of equipment that will allow safe operation within the constraints of existing facility conditions. Examples of such constraints that are evaluated and addressed by the dry storage program include limited basin depth, varying fuel lengths up to 4 m, (13 ft), fissile loading limits, canister closure design, post-load drying and closure of the canisters, instrument selection and installation, and movement of the canisters to storage casks. The initial pilot phase restricts the fuels to shorter length fuels that can be loaded to the canister directly underwater; subsequent phases will require use of a shielded transfer system. Removal of the canister from the basin, followed by drying, inerting, closure of the canister, and transfer of the canister to the storage cask are completed with remotely operated equipment and appropriate shielding to reduce personnel radiation exposure. (authors)

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DOE OSTI2013

Stakeholder Transportation Scorecard: Reviewing Nevada's Recommendations for Enhancing the Safety and Security of Nuclear Waste Shipments - 13518

As a primary stakeholder in the Yucca Mountain program, the state of Nevada has spent three decades examining and considering national policy regarding spent nuclear fuel and high-level radioactive waste transportation. During this time, Nevada has identified 10 issues it believes are critical to ensuring the safety and security of any spent nuclear fuel transportation program, and achieving public acceptance. These recommendations are: 1) Ship the oldest fuel first; 2) Ship mostly by rail; 3) Use dual-purpose (transportable storage) casks; 4) Use dedicated trains for rail shipments; 5) Implement a full-scale cask testing program; 6) Utilize a National Environmental Policy Act (NEPA) process for the selection of a new rail spur to the proposed repository site; 7) Implement the Western Interstate Energy Board (WIEB) 'straw man' process for route selection; 8) Implement Section 180C assistance to affected States, Tribes and localities through rulemaking; 9) Adopt safety and security regulatory enhancements proposed states; and 10) Address stakeholder concerns about terrorism and sabotage. This paper describes Nevada's proposals in detail and examines their current status. The paper describes the various forums and methods by which Nevada has presented its arguments and sought to influence national policy. As of 2012, most of Nevada's recommendations have been adopted in one form or another, although not yet implemented. If implemented in a future nuclear waste program, the State of Nevada believes these recommendations would form the basis for a successful national transportation plan for shipments to a geologic repository and/or centralized interim storage facility. (authors)

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DOE OSTI2013

Walk the Line: The Development of Route Selection Standards for Spent Nuclear Fuel and High-level Radioactive Waste in the United States - 13519

Although storage facilities for spent nuclear fuel (SNF) and high-level radioactive waste (HLRW) are widely dispersed throughout the United States, these materials are also relatively concentrated in terms of geographic area. That is, the impacts of storage occur in a very small geographic space. Once shipments begin to a national repository or centralized interim storage facility, the impacts of SNF and HLRW will become more geographically distributed, more publicly visible, and almost certainly more contentious. The selection of shipping routes will likely be a major source of controversy. This paper describes the development of procedures, regulations, and standards for the selection of routes used to ship spent nuclear fuel and high-level radioactive waste in the United States. The paper begins by reviewing the circumstances around the development of HM-164 routing guidelines. The paper discusses the significance of New York City versus the Department of Transportation and application of HM-164. The paper describes the methods used to implement those regulations. The paper will also describe the current HM-164 designated routes and will provide a summary data analysis of their characteristics. This analysis will reveal the relatively small spatial scale of the effects of HM 164. The paper will then describe subsequent developments that have affected route selection for these materials. These developments include the use of 'representative routes' found in the Department of Energy (DOE) 2008 Supplemental Environmental Impact Statement for the formerly proposed Yucca Mountain geologic repository. The paper will describe recommendations related to route selection found in the National Academy of Sciences 2006 report Going the Distance, as well as recommendations found in the 2012 Final Report of the Blue Ribbon Commission on America's Nuclear Future. The paper will examine recently promulgated federal regulations (HM-232) for selection of rail routes for hazardous materials transport. The paper concludes that while the HM 164 regime is sufficient for certain applications, it does not provide an adequate basis for a national plan to ship spent nuclear fuel and high-level radioactive waste to centralized storage and disposal facilities over a period of 30 to 50 years. (authors)

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DOE OSTI2013

The Polymers for Liquid Radioactive Waste Solidification: a Lost Chapter in the History of Engineering or a Step Forward? - 13529

Ideas on the application of polymers for the liquid radioactive waste immobilization go a way back, and the first studies in the area were published 30-40 years ago. One should admit that regardless of the fairly large number of publications appeared in the past years currently the interest in this work came down greatly. It was the successful assimilation and worldwide implementation of the LRW cementation technology caused a slump in the interest in polymers. But today it's safe to say that the situation slowly changes, particularly due to the market appearance of the high-tech polymers manufactured by Nochar Company, and unique properties of these polymers gradually raise the demand in various countries. The results of multiple experiments performed with the simulated solutions have passed the comprehensive tests with actual waste. The economic effect from the implementation of the new technology is defined by the volume reduction of waste coming onto the repository, by the decline in the cost of transportation and of the repository construction on account of cutting down the construction volume. Interesting results have been obtained during the search for the technical decisions that would allow using the polymer materials in the processing technology of the industrial toxic waste. One more promising area of the possible application of polymers should be pointed out. It is the application of polymer materials as the assets for the emergency damage control when the advantages of the polymers become obvious. (authors)

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DOE OSTI2013

Are We Serious in the US about the Disposal of HLW? - 13561

Since all efforts to date to dispose of HLW in the US have been unsuccessful, the following specific actions need to be taken if we are serious about such disposal: - The requirement in the EPA environmental radiation protection standards to predict the behavior of these unwanted residuals for one million years is meaningless. The Standards must be revisited. - Characterize two sites. There are myriad ways a site can be found to be unacceptable. Additionally, the existing HLW inventory requires a second repository. - Congress should specify incentives to states under consideration for a site. Perhaps 5% of total cost would be appropriate. - An independent technical review group should be established in such states to evaluate a proposed repository similar to the New Mexico Environmental Evaluation Group (EEG) for the WIPP Project because the state's interests are not necessarily the same as DOE's. - Acceptance or rejection of a proposed site should be based on technical issues, not social ones. Professionals in this field should present papers identifying the merits of HLW disposal in their own state. The scarcity of such research suggests Not In My Back Yard (NIMBY) syndrome. - Medical diagnostic ionizing radiation exposure to the US public is now 8,000 times greater than radiation exposure from nuclear energy. People accept this believing the benefits outweigh any risks. A major effort needs to focus on both benefits as well as risks of radioactive waste disposal. - DOE needs to announce preferences of host rock formations, incentives for states, and potential consequences should we fail to act. (author)

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DOE OSTI2013

Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2 discharge reuse. The EM2 waste disposal profile is effectively only fission products, which reduces the mass (about 3% vs LWR), average half life, heat and long term radio-toxicity of the disposal. Widespread implementation of EM2 fuel cycle is highly significant as it would increase world energy reserves; the remaining energy in U.S. LWR SNF alone exceeds that in the U.S. natural gas reserves. Unlike many LWR SNF disposition concepts, the EM2 fuel cycle conversion of SNF produces energy and associated revenue such that the overall project is cost effective. By providing conversion of SNF to fission products the fuel cycle is closed and a non-repository LWR SNF disposition path is created and overall repository requirements are significantly reduced. (authors)

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DOE OSTI2013

Improving the actinides recycling in closed fuel cycles, a major step towards nuclear energy sustainability

Increasing the sustainability of nuclear energy is a longstanding road that requires a stepwise approach to successively tackle the following 3 objectives. First of all, optimize the consumption of natural resource to preserve them for future generations and hence guarantee the energetic independence of the countries (no uranium ore is needed anymore). The current twice-through cycle of Pu implemented by France, UK, Japan and soon China is a first step in this direction and already allows the development and optimization of the relevant industrial processes. It also allows a major improvement regarding the conditioning of the ultimate waste in a durable and robust nuclear glass. Secondly, the recycling of americium could be an interesting option for the future with the deployment of FR fleet to save the repository resource and optimize its use by allowing a denser disposal. It would limit the burden towards the future generations and the need for additional repositories before several centuries. Thirdly, the recycling of the whole minor actinides inventory could be an interesting option for the far-future for strongly decreasing the waste long-term toxicity, down to a few centuries. It would bring the waste issue back within the human history, which should promote its acceptance by the social opinion.

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DOE OSTI2013

The used nuclear fuel problem - can reprocessing and consolidated storage be complementary?

This paper describes our CISF (Consolidated Interim Storage Facilities) and Reprocessing Facility concepts and show how they can be combined with a geologic repository to provide a comprehensive system for dealing with spent fuels in the USA. The performance of the CISF was logistically analyzed under six operational scenarios. A 3-stage plan has been developed to establish the CISF. Stage 1: the construction at the CISF site of only a rail receipt interface and storage pad large enough for the number of casks that will be received. The construction of the CISF Canister Handling Facility, the Storage Cask Fabrication Facility, the Cask Maintenance Facility and supporting infrastructure are performed during stage 2. The construction and placement into operation of a water-filled pool repackaging facility is completed for Stage 3. By using this staged approach, the capital cost of the CISF is spread over a number of years. It also allows more time for a final decision on the geologic repository to be made. A recycling facility will be built, this facility will used the NUEX recycling process that is based on the aqueous-based PUREX solvent extraction process, using a solvent of tri-N-butyl phosphate in a kerosene diluent. It is capable of processing spent fuels at a rate of 5 MT per day, at burn-ups up to 50 GWD per ton of spent fuels and a minimum of 5 years out-of-reactor cooling.

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DOE OSTI2012

Implementing Geological Disposal of Radioactive Waste Technology Platform From the Strategic Research Agenda to its Deployment - 12015

Several European waste management organizations (WMOs) have initiated a technology platform for accelerating the implementation of deep geological disposal of radioactive waste in Europe. The most advanced waste management programmes in Europe (i.e. Finland, Sweden, and France) have already started or are prepared to start the licensing process of deep geological disposal facilities within the next decade. A technology platform called Implementing Geological Disposal of Radioactive Waste Technology Platform (IGD-TP) was launched in November 2009. A shared vision report for the platform was published stating that: 'Our vision is that by 2025, the first geological disposal facilities for spent fuel, high-level waste, and other long-lived radioactive waste will be operating safely in Europe'. In 2011, the IGD-TP had eleven WMO members and about 70 participants from academia, research, and the industry committed to its vision. The IGD-TP has started to become a tool for reducing overlapping work, to produce savings in total costs of research and implementation and to make better use of existing competence and research infrastructures. The main contributor to this is the deployment of the IGD-TP's newly published Strategic Research Agenda (SRA). The work undertaken for the SRA defined the pending research, development and demonstration (RD and D) issues and needs. The SRA document describing the identified issues that could be worked on collaboratively was published in July 2011. It is available on the project's public web site (www.igdtp.eu). The SRA was organized around 7 Key Topics covering the Safety Case, Waste forms and their behaviour, Technical feasibility and long-term performance of repository components, Development strategy of the repository, Safety of construction and operations, Monitoring, and Governance and stakeholder involvement. Individual Topics were prioritized within the Key Topics. Cross-cutting activities like Education and Training or Knowledge Management as well as activities remaining specific for the WMOs were as well identified in the document. For example, each WMO has to develop their own waste acceptance rules, and plan for the economics and the funding of their waste management programmes. The challenge at hand for the IGD-TP is to deploy the SRA. This is carried out by agreeing on a Deployment Plan (DP) that guides organizing the concrete joint activities between the WMOs and the other participants of the IGD-TP. The first DP points out the coordinated RD and D projects and other activities that need to be launched to produce these results over the next four to five years (by the end of 2016). The DP also describes general principles for how the joint work can be organised and funded. (authors)

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DOE OSTI2012

Waste Isolation Pilot Plant Status and Plans - 2012 - 12049

The Waste Isolation Pilot Plant (WIPP), a deep geologic repository for safe disposal of long-lived transuranic radioactive waste related to the nation's defense, is completing its 12. year of operations. WIPP's mission includes coordination of all Department of Energy (DOE) sites to prepare, package and characterize transuranic (TRU) waste for final shipment and emplacement in WIPP. Five of the 10 disposal panels planned have been filled and sealed from ventilation. Additional small quantity sites have been de-inventoried by consolidating their waste through the certified characterization line at the Idaho National Laboratory (INL). New emplacement methods for RH waste in shielded containers are being considered for disposal by WIPP's regulatory authorities. A new large Type B shipping package, was added to the WIPP transportation fleet, and facility modifications to the WIPP waste unloading and emplacement processes for large containers were completed in 2011. Shipments from the Savannah River site in these new large rectangular packages began in August 2011. Licensing efforts are proceeding for a new criticality control over-pack container that will allow almost twice the fissile content to be shipped than previously. This will reduce the number and cost of shipments of Special Nuclear Material (SNM) declared as waste. Modifications to WIPP regulatory requirements for the disposal footprint and disposal unit closure systems are in progress. These, and other developments, make for exciting times at WIPP. This paper provides an up-to-date look at the many aspects of America's only deep geologic long-lived radioactive waste repository, which is completing its 12. year of operations. A record year of safe and compliant shipments to WIPP tops the list of accomplishments in 2011. Four more small quantity sites were de-inventoried by consolidating their waste through the certified characterization line at INL in 2011. A new Type B shipping package, the TRUPACT-III has been added to the transportation fleet, and large waste boxes are being shipped from SRS without the need for repackaging. New emplacement methods for remote-handled waste in shielded containers are undergoing regulatory review. WIPP plans to license a new criticality control payload container that will allow almost twice the fissile content to be shipped than previously, thereby reducing the number and cost of shipments of SNM declared as waste. Other regulatory modifications planned in 2012 include approval of a design change that would replace the disposal concept for panels 9 and 10 from using the common access drifts (the 'mains') with a new footprint south of panels 4 and 5. DOE also plans to change the panel closure design set forth in its certification by EPA and the HWFP by the NMED. The panel closure design change will be a rule making under EPA's procedures and a class 3 permit modification request under NMED procedures. Plans for achieving 90% of legacy TRU waste retrieval and emplacement in WIPP by 2015 have been developed. Key to the success of this so-called 90/15 plan is adequate funding, both for WIPP operations, as well as for TRU retrieval programs at the generator sites. (authors)

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DOE OSTI2012

Radioactive Spent Ion-Exchange Resins Conditioning by the Hot Supercompaction Process at Tihange NPP - Early Experience - 12200

Spent ion-exchange resins are considered to be problematic waste that, in many cases, requires special approaches and precautions during their conditioning to meet the acceptance criteria for disposal. In Belgium, for economical reasons, the Volume Reduction Factor is a key criterion. After Tractebel Engineering performed a technical and economical comparison of the industrially available systems, Tihange NPP decided to install a spent ion-exchange resins hot supercompaction unit with Tractebel Engineering in the role of architect-engineer. The treatment and conditioning unit processes the spent ion-exchange resins through the following steps: dewatering of the resins, drying the resins under deep vacuum, discharging the dried resins into compactable drums, super-compacting the drums to generate pellets, grouting the pellets into standard 400 litres waste drums (overpacks) licensed for final disposal in the near-surface repository in Belgium. Several developments were required to adapt the reference process and equipment to PWR spent ion-exchange bead resins and Belgian radioactive waste acceptance criteria. In order to avoid cracks on the compacted drum, and external surface contamination from resin leaks, some improvements were achieved to minimize spring-back as well as the risk of cracking the drum wall. Placing the compactable drum inside a second, slightly larger drum, guarantees clean and reproducible pellets. Currently the commissioning phase is on-going. Numerous process validation tests have been completed. An acceptance file was transmitted to the Belgian Waste Management Authority recently. This will be followed by demonstration tests necessary to obtain their final acceptance of the installation. More than 3 800 drums of mixed powdered and bead resins have been processed by the reference Hot Compaction process, achieving a Volume Reduction Factor (VRF) of 2.5. The equipment has been proven to be a reliable technology with low operation and maintenance costs. Tractebel Engineering has managed the construction of a new application of this process in Belgium at Tihange NPP. Several developments were required to adapt the reference process and equipment to PWR spent ion-exchange bead resins and Belgian radioactive waste acceptance criteria. The chosen method of conditioning (draining, drying and compaction of the spent resins followed by grouting of the pellets in a 400 litres drum) immobilizes the spent resins under the form of a solid, compact, stable, and non dispersible block free of interstitial water. The various series of inactive tests which were conducted at Tihange NPP, helped among others to determine the best design of the compactable drum and lid and to set the value of critical parameters such as vapour temperature at the end of drying, speed, force and duration of compaction. In an environment of very limited space for interim storage and in the absence of an operational final repository site, or in the case of high final disposal costs, the process exhibits the following key advantages: - Achieving a Volume Reduction Factor (VRF) close to 1 (overpack included) for the interim storage instead of increased volumes observed with other currently available processes; - Achieving a water free end product; - Creating a flexible waste product for interim storage (pellet), which can be retrieved and routed into alternative types of package later, if not initially grouted; - Using well proven standard technologies like drying and compaction; - Flexible use of the system components for the supercompaction of other operational solid waste streams when not conducting resins conditioning campaigns. (authors)

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DOE OSTI2012

Joint EM-NE-International Study of Glass Behavior over Geologic Time Scales - 12303

Vitrification has been chosen as the best demonstrated available technology for waste immobilization worldwide. To date, the contributions of physical and chemical processes controlling the long-term glass dissolution rate in geologic disposal remain uncertain, leading to a lack of international consensus on a rate law for glass corrosion. Existing rate laws have overcome uncertainty through conservatism, but a thorough mechanistic understanding of waste form durability in geologic environments would improve public and regulator confidence. If it is possible to take credit for the true durability of the waste form in repository system evaluations, then it is possible to design the repository with less conservatism with concomitant cost savings. To gain a fundamental understanding of the dissolution rate law, six nations have joined together to formulate a joint plan for collaborative research into the mechanisms controlling the long-term corrosion of glass. This report highlights the technical program plan behind the US portion of this effort, with an emphasis on the current understanding (and limitations) of several mechanistic theories for glass corrosion. Some recent results are presented to provide an example of the ongoing research. Atom probe tomography has been used to provide a high-resolution analysis of elemental concentration gradients present at the hydrated glass / pristine glass interface in SON68 after 25.75 years of corrosion in a simulated granitic groundwater at 90 deg. C. The most valuable result of these initial studies is the success of the technique. Characterization by APT had never been previously demonstrated for glass corrosion layers. The resolution of APT is a powerful addition to the tools with which we can investigate the mechanisms dominating glass corrosion. Some other key results of this study include the observation that the elemental interfacial width between the hydrated glass and pristine glass appears to be much sharper (∼2 nm for B, Na and Al) than had been previously measured using nanoSIMS (∼240 nm). It is not clear whether the APT analysis and nanoSIMS characterizations were possibly performed on topographically unique regions, or whether nanoSIMS overestimated the elemental width. However, the APT data seems very convincing that the elemental width can be much sharper than was previously thought. This result calls into question some of the assumptions made for the diffusion-control models of glass dissolution, since such a sharp profile would not match the diffusion coefficients used to date. Other results, such as the observation of apparently layered concentration profiles, show that gel evolution is not as simple as is currently assumed in nearly every model. This task is a good example of the collaborative nature of the I-TEAM effort. Based on experimental needs and differences in expertise, scientists from DOE and CEA worked together to change the level of understanding in the field. These types of interactions are nearly ubiquitous among the tasks in the technical program plan. With the excellence of the team in place and the willingness of the participants to work together for a common understanding, the stated goal of consensus on the mechanistic basis for radionuclide release from glass is well within reach. (authors)

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DOE OSTI2012

Nye County Nevada Perspectives on the State of the Yucca Mountain Project - 12388

Responding to the Department of Energy decision to try to withdraw the Yucca Mountain license application and the Administration actions to close down the Yucca Mountain project, Nye County undertook a number of activities to articulate its support for continuing the Yucca Mountain project. The activities included responding to inquiries from federal agencies, including investigations undertaken by the Government Accountability Office addressing other potential uses for the Yucca Mountain site, responding to a Draft Environmental Impact Statement on the possible use of Yucca Mountain for disposal of Greater than Class C wastes, testifying in hearings, and interacting with the President's Blue Ribbon Commission on America's Nuclear Future. The paper summarizes Nye County's position on the Yucca Mountain repository, Nye County's perspectives on the various activities that were developed and considered by the Government Accountability Office, Nye County's concerns with the use of the Nevada National Security Site for Disposal of Greater than Class C Low-Level Radioactive Wastes, testimony of Nye County officials expressing local community support for the Yucca Mountain project, and Nye County's perspectives on recommendations provided by the Blue Ribbon Commission to move the nation's high-level radioactive waste disposal programs forward without consideration of the role Yucca Mountain could have served in those recommendations. Nye County believes that every effort should be made to, at a minimum, fund the Nuclear Regulatory Commission to complete the license application review. Then, if Congress does decide to change the Nuclear Waste Policy Act, there will be valuable information available to support new policy development. This administration contends that Congressional language associated with the FY2010 and FY2011 appropriations and authorization process is sufficient evidence of its intent to terminate the Yucca Mountain repository program. The appropriation process needs to be explicit that, absent explicit language to the contrary, the Nuclear Waste Policy Act stands. It also should include language that requires the Department of Energy to preserve all necessary records until the Nuclear Waste Policy Act is amended or rescinded by specific Congressional action. (authors)

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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)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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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