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All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

As-run physics analysis for the EPRI-2 experiment for cycle 153B

The purpose of this ECAR is to document the as-run physics analysis and source term in Curies and decay heat rate (Watts) for the shipment of the EPRI-2 experiment located in the Center Flux Trap (CFT). EPRI-2 was irradiated during cycle 153B. The experiment irradiation ended on April 13, 2013. The as-run heat rates and flux/fluence results were calculated using the MCNP ATR full core model. The heat rates and flux are calculated based on an average center lobe power of 30.8 MW. The fluence is based on the operating time of 13.45 days. ORIGEN2 Version 2.2 is used to determine the decay heat rate, gamma spectrum, and source term for the EPRI-2 experiment. The source term analysis was performed as requested by the project staff to demonstrate compliance with shipping requirements in the GE-2000 cask following irradiation as well as provide an estimated source term to support Post-Irradiation Examination as needed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Mobile heat pipe cooled fast reactor system

A mobile heat pipe cooled fast nuclear reactor may be configured for transportation to remote locations and may be able to provide 0.5 to 2 megawatts of power. The mobile heat pipe cooled fast reactor may contain a plurality of heat pipes that are proximate to a plurality of fuel pins inside the reactor. The plurality of heat pipes may extend out of the reactor. The reactor may be configured to be placed in a standard shipping container, and may further be configured to be contained within a cask and attached to a skid for easier transportation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

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

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

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DOE OSTI2021

Nevada National Security Site V3XA Spherical Explosive Confinement Vessel Awareness Package

This report summarizes the plans and activities that will facilitate shipment of the V3XA experimental spheres (a.k.a., Nevada Spheres) from the Nevada National Security Site (NNSS) to the Idaho National Laboratory’s (INL) Advanced Mixed Waste Treatment Project (AMWTP). The spheres require shipment to INL to be segmented, characterized, and repackaged for ultimate disposal at the Waste Isolation Pilot Plant (WIPP). The NNSS does not currently have the facilities or safety basis to perform these activities. The INL has the facilities, the INL M&O contractor owns an 8-120B cask, and AMWTP has a WIPP Waste Acceptance Criteria approved characterization program. The NNSS has shipped other legacy transuranic waste to INL for WIPP certified characterization in the past.

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DOE OSTI2017

Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) 4.x

A new, integrated data and analysis system has been designed to simplify and automate the performance of accurate and efficient evaluations for characterizing the input to the overall waste management system-UNF-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). A relational database has been assembled to provide a standard means by which UNF-ST&DARDS can succinctly store and retrieve M&S parameters for specific SNF analysis. Users can import SNF data into this database for user-specific analysis. A library of various analysis model templates provides the ability to communicat the various set of M&S parameters to the most appropriate M&S application. A process manager facilitates performance of actual, as-loaded, assembly-specific and cask-specific evaluations. Interactive visualization capabilities facilitate data analysis and results interpretation.

Banerjee, Kaushik↗
DOE OSTI2022

Mobile heat pipe cooled fast reactor system

A mobile heat pipe cooled fast nuclear reactor may be configured for transportation to remote locations and may be able to provide 0.5 to 2 megawatts of power. The mobile heat pipe cooled fast reactor may contain a plurality of heat pipes that are proximate to a plurality of fuel pins inside the reactor. The plurality of heat pipes may extend out of the reactor. The reactor may be configured to be placed in a standard shipping container, and may further be configured to be contained within a cask and attached to a skid for easier transportation.

McClure, Patrick Ray↗
DOE OSTI2014

Assembly Level Modeling and Transportation Damage Prediction of Used Nuclear Fuel (UNF) Cladding

Under U.S. Nuclear Regulatory Commission (NRC) regulations, UNF must maintain its integrity in such a way that it can withstand the physical forces of handling and transportation associated with restaging the fuel and moving it to a different location such as an interim storage site, a geologic repository, or a treatment/recycling facility. Hence, understanding mechanical performance under cumulative loading stemming from normal conditions of storage (NCS), transfer from storage container to transport container if needed, and normal conditions of transport (NCT) is necessary as it establishes part of the safety basis via maintaining the fuel confining boundary (geometry), maintaining criticality safety, and is one of the critical components to the preservation of retrievability. An initial demonstration of modeling and simulation capabilities was performed as part of a collaborative effort between Pacific Northwest National Laboratory (PNNL), Sandia National Laboratory (SNL), Idaho National Laboratory (INL), Oak Ridge National Laboratory (ORNL), and the Transportation Technology Center Inc. (TTCI). This demonstration completed preliminary deterministic evaluations of moderate-to-high burnup UNF mechanical performance under NCS and NCT conditions. This paper focuses on work performed by PNNL on the assembly level modeling of a UNF assembly and fuel cladding damage predictions under NCT loads. The Finite Element (FE) model created for a 17 x 17 Pressurized Water Reactor (PWR) optimized fuel assembly (OFA) is discussed along with the communications to and from the fuel rod-level and cask-level models. Additionally, the methodology for determining cumulative damage is discussed including the 3000 mile damage prediction from a 10 second simulation. The damage results are shown for five shock and vibration broadened loading histories. It is found that predicted cladding strains were not large enough to cause structural failure, but cyclic strains roughly projected for the entire route were significant in some cases. The total damage from summation of the worst shock and vibration cases evaluated is approximately 18% of the expected fatigue limit. Therefore it is predicted that the fuel rods do not fail during NCT given the assumptions used throughout this initial demonstration.

used nuclear fuel, UNF, trasportation, damage, fue↗
DOE OSTI2022

The Hanford Lead Canister Collaboration

The Hanford Site, Management of the Cesium and Strontium Capsules (MCSC) Project is preparing to place cesium and strontium material into dry storage using stainless steel canisters that are housed inside vertical concrete casks. Part of the site’s aging management program is to use a spare canister system as a leading indicator canister to provide advance warning for signs of chloride induced stress corrosion cracking (CISCC) or any other kind of canister degradation that might take place over the system’s 300-year design life. The Hanford Lead Canister (HLC) will use electric heaters to simulate the decay heat of cesium and strontium to ensure the HLC simulates the environment and conditions the capsule-loaded systems will experience. In addition to being part of the site’s aging management program, the HLC is being made available for research and development that advances the industry regarding canister mitigation and repair (M&R). The HLC system closely resembles current spent nuclear fuel (SNF) dry storage systems, making the HLC useful for demonstrating technologies or performing research that is relevant to SNF canisters . Several organizations are collaborating to prepare the HLC, along with technical strategy, to perform the needed research and development activities, to include the Pacific Northwest National Laboratory (PNNL), Central Plateau Cleanup Company (CPCCo), and the Electric Power Research Institute (EPRI). In one example of this collaborative research, EPRI is planning to use the HLC to demonstrate in-situ M&R technologies under controlled conditions before the HLC is deployed at Hanford. There is a period of several years when the HLC will be available for canister M&R activity before it is deployed. This paper describes the development of the HLC, its role in the Hanford site’s MCSC Project aging management program, and some of the research activities that are planned for the HLC.

Klymyshyn, Nicholas A.↗
DOE OSTI2013

ASSESSING MATERIAL DEGRADATION EFFECTS ON LONG TERM STORAGE STRUCTURAL ANALYSES

PNNL staff are currently working to conduct analyses, integrate experimental data, and develop the technical basis for extended long term storage and subsequent transportation of used nuclear fuel. Analyses will serve to augment the experimental work by addressing degradation phenomena not readily obtainable through experimentation, such as low temperature creep over very long periods of time. Analyses can also address issues such as concrete degradation that may not require a comprehensive R&D program to resolve, as well as evaluation of engineering design solutions that can be implemented to mitigate the amount and extent of experimental data needed. Most importantly, analyses are necessary to determine how much degradation any structure, system, or component can experience and still maintain its ability to meet its safety functions. This paper focuses on various structural modeling issues that are affected by material degradation in extended long term storage scenarios. Typical storage cask and transportation package designs have a safety basis that includes a number of structural static or dynamic finite element analysis (FEA) models. Some of the features of these standard models need to be reconsidered when the fuel is in an extended long term storage state. In this paper, a number of standard safety basis FEA models are discussed and evaluated for their applicability to long term storage scenarios. A vertical handling drop analysis is used to illustrate a case where additional modeling is needed to define the loads and loading history needed to properly evaluate material degradation in long term storage scenarios.

Klymyshyn, Nicholas A.↗
DOE OSTI2023

Modeling Non-UO2 Fuel With UNF-ST&DARDS

The U.S. Department of Energy’s Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) provides an easy-to-use interface to analyze irradiated UO2 fuel by allowing all analysis to be performed within the software and being able to store and use dozens of fuel assembly, canister, and cask designs [1]. However, performing these same analyses with non-UO2 fuel, such as UN or U3Si2, requires more user intervention in the process. This work uses UN, UN-ZrO2, and U3Si2 fuel to demonstrate how to perform criticality analyses in the current versions of UNF-ST&DARDS and how a non-UO2 fuel will compare to UO2. This work is part of a larger effort that also includes shielding and thermal analyses, but they will not be discussed.

Ivanusa, Pavlo↗
DOE OSTI2021

Experience with Inner Reflector Plug exchange in SNS

The Inner Reflector Plug (IRP) is a central component of the Spallation Neutron Source target monolith, which houses the mercury target and four liter-sized neutron moderator units. It is exposed to high-level radiation fields during routine operation and builds up significant activity. The IRP needs to be replaced due to moderator neutron poison and decoupler burn-out, which is used for shaping neutron pulses. The first IRP exchange took place in March 2018. The old IRP was extracted from the target monolith, providing space for the new one. It was split into three segments, each of which was handled separately. The lowest section of the IRP is the largest segment in size and in activity and is temporarily stored on-site for cool down before conduction post irradiation examination. In support of planning the replacement activities, a wide range of activation and transport analyses were performed. This included calculating isotope inventories and the radiation fields for each segment as it is extracted in storage casks, and the radiation field from the empty IRP pit in the target monolith. While the replacement was taking place, measurements were performed and later on compared to the calculations. During these studies, it was discovered that a significant contributor to the radiation field from the lower IRP segment is from photo-neutrons. Photo-nuclear physics was added to the analyses and calculated results compared well with measured dose rates.

Popova, Irina I.↗
DOE OSTI2017

Improving Decommissioning Success using the Tool of Virtual Reality; Increasing Worker Safety and Project Schedule Stability - 17193

operations, especially underwater, is essential. Equally essential is a precise game plan for disassembly which focuses attention on the skilled remote handling operators who must produce the results. Project success is based on clean, efficient cuts and take-aways, prepared paths for waste placement and containment, programmed return paths; and practice, practice, practice on the part of the operators. Recently, our Partner, WAK Ruackbau- und Entsorgungs- GmbH (Decommissioning and Waste Management Company) was tasked with dismantling and decommissioning the Karlsruhe Reprocessing Plant. The most difficult task in this project is the dismantling of the vitrification plant. Waelischmiller was selected to address this high-consequence project. Together, WAK and Waelischmiller determined that development of a virtual reality suite, where management and operators could create a virtual vitrification layout and subsequently map and rehearse their dismantling plan was essential. This approach, using Virtual Reality (VR) as a tool in streamlining operations and rehearsing difficult moves and cuts, will be a game changer for this and future projects. It will be a first-of-a-kind in Germany. The VR suite is contained in a room at the Karlsruhe facility with a large screen, control cabinet, simulation computer, a Waelischmiller Joyarm and Masterarm. To support the dismantling of the vitrification plant, WAK/HWM will conduct interactive feasibility studies for each dismantling step. All movements; cuts, take-aways and vessel segment insertion into transportation and disposal casks will be established, programed and practiced within virtual test suite. This allows savings in cost-intensive and complex construction of test rigs. Operating personnel can be trained in the VR suite, without taking the risk of damaging expensive remote handling equipment.While the operators are working on the operating element, they stand in front of a wide screen. Instead of a real TELBOT{sup R}, a virtual TELBOT{sup R} is moving on the screen in virtual reality. While practicing, the operators cannot damage the arm and the equipment, which is a big advantage. Moreover an interactive feasibility study can be realized without building a real test rig (mock-up). Consequently significant cost-expenses can be reduced. This paper will describe the virtual suite used by the WAK in Karlsruhe and the advantages of this new and innovative approach, which is unique in decommissioning of nuclear facilities. The viewpoint of each (WAK and HWM), and the decisions and benefits realized along the project path will be detailed in the paper. (authors)

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DOE OSTI2017

Integrated Solutions for Interim Used Fuel Management: From Pool Unloading to Safe Delivery to Recycling or Repository Facilities - 17235

Safe management of used nuclear fuel (UNF) is key to the survival of the nuclear industry. That is why it is more important than ever to have a consent-based, practical, credible, and sustainable roadmap for managing the back-end of the nuclear cycle. Worldwide UNF Light Water Reactors (LWR) inventories are currently around 200,000 tons and are estimated to grow to more than 350,000 tons by 2030, and operating nuclear reactors in the U.S. continue to discharge UNF every 18 to 24 months. International experience has demonstrated that the development of a permanent geological repository can be a multi-decade process, especially in the face of public acceptance challenges. Meanwhile, the solution for permanent disposal of UNF in the U.S. is moving at a glacial pace. As a result, industrial stakeholders have been forced to find practical interim UNF storage solutions in order to continue safely operating their reactors. Among those interim solutions are the cask- or canister-based dry storage systems that have dramatically increased in the last 10 years as they provide flexibility for expansion with a low up-front investment. With interim dry storage in place for more than 30 years, our industry is entering a new phase as we are forced to prepare for a few more decades before geological repositories begin operating. Consequently, interim dry storage has to be extended for 80 years or more. Thus, new considerations have to be taken into account to ensure the safety of this stored used fuel: - Interim storage systems are aging; - High burn-up fuel behavior overtime is uncertain; - Used fuel is losing its self-protection; - The number of shutdown sites without a fuel transfer facility, such as a pool, is increasing; - Some owners of Independent Spent Fuel Storage installation (ISFSI) at shutdown sites are not nuclear operators anymore. With increasing uncertainties, AREVA TN -- based on its long expertise with used fuel transportation and interim storage systems -- is adapting its portfolio to develop integrated and flexible solutions that will remove the burden of interim used fuel management from nuclear operators. AREVA TN is developing comprehensive solutions to reduce risks including the continued supply of proven safe high-performance dry storage systems; on-site services to safely transfer used fuel from pool to pad; the optimization of fuel heat load in the storage pool and pad; innovative aging management services including inspection, monitoring and repair; a realistic solution for consolidated interim storage with a priority to remove Used Nuclear Fuel stored in dry storage systems from shutdown sites; world-class transport and logistics services to safely move the used fuel from nuclear reactor sites to consolidated storage and repository or recycling as applicable, and turnkey management of shutdown ISFSIs. Due to the lack of an available path for transporting used fuel away from reactors, nuclear operators are faced with many challenges and growing uncertainties. Our presentation will describe the practical solutions developed and being developed in order to manage used fuel on behalf of nuclear operators and allows them to continue focusing on the important job of producing carbon-free clean energy. (authors)

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DOE OSTI2017

Optimization Scheme for Establishing Queues for Shipment of UNF from Shutdown Reactor Sites - 17554

Approximately 14, 815 used nuclear fuel assemblies (5,800 MTHM) are stored in 486 dry storage cask systems located at 12 independent spent fuel storage installations (ISFSIs) located at shutdown reactor sites located across the U.S. that have no access to spent fuel pools. A preliminary evaluation was performed by the University of North Carolina at Charlotte (UNCC) in collaboration with AREVA to establish optimization schemes for establishing queues for the removal of this used/spent nuclear fuel (hence forth identified as UNF) and Greater than Class C low level wastes (GTCC) found at these shutdown reactor sites. These queues assumed the major portion of the shipment of the UNF and GTCC would be by rail, with heavy haul truck or barge potentially utilized to move the UNF from the site to rail for sites not served directly by rail. The destination for these rail shipments were a hypothetical consolidated interim storage facility (CSF) located in West Texas. The queues were developed based on the total costs and time to finish all shipments needed to remove all the UNF from these shutdown sites. Results are presented for optimization schemes based on a continuous flow of shipments using 1, 2, and 3 crews dedicated to individual ISFSIs and discrete shipments using 2 crews capable of performing operations at multiple ISFSIs. (authors)

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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)

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DOE OSTI2015

New container development for the international market - 15656

Designs of casks and containers for the disposal of radioactive waste have to meet safety requirements for transport and storage. Since nuclear law is a national issue, the legal specifications often differ significantly. To improve the flexibility of a given basic design, the development of a set of generic design elements and the application of favorable material characteristics are of major importance. As an example for the international approach the design development for an IP-2 container for radioactive non-fuel waste made of Ductile Cast Iron (DCI) for the UK market in the frame of the specific national waste disposal concept is described. The results of drop tests performed during the licensing procedure and of parametric analysis work regarding a wide spectrum of accidental scenarios and impacts show the excellent behavior of the container concept and confirm the related favorable manufacturing procedures. Furthermore, the development work provided a significant set of proposals for the future detailed work. (authors)

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DOE OSTI2018

Panel reports session

This section brings together the reports of the 31 panel sessions listed below: - Avoiding the Human Capital Cliff in the DOE (PS22); - Implementing Technically-Based Cleanup; US DOE Balancing Regulatory and Fiscal Complexity (PS24); - United Kingdom - Remediation of Sellafield Legacy Ponds and Silos (PS25); - Aligning Decision-Making to Deliver Timely Results in Large Clean-up Programs (PS38); - Transition to Government Owned-Contractor Operated Site at Canadian Nuclear Laboratories (PS39); - Emerging Middle East Nuclear States' Status and Plans (PS40); - NPP Panel on Waste Management (PS41); - Robotics - International Innovations in Robotic Decommissioning Technology (PS43); - Fukushima Panel (1/2): Japan Fukushima Daiichi Decommissioning Update (PS54); - Fukushima Panel (2/2): Update on the Revitalization of Fukushima and the Surrounding Area (PS55); - US DOE LLW On-Site Disposal: LLW Disposal Facility Federal Review Group Review Process (PS56); - Challenges in DOE HLW Tank Management: Facility Start-up and Waste Disposition (PS57); - Waste Management Planning for Utility Small Modular Reactors (PS58); - US Nuclear Power Plant Waste Management - LLW Disposal Issues (PS59); - DOE-EM Robotics and Emerging Technologies Road-map (4/8) (PS60); - Graduating Students and New Scientists and Engineers - Wants and Needs - Are Companies Even Listening? (PS63); - U.S. Nuclear Robotics: Yesterday, Today, Tomorrow (5/8) (PS74); - US DOE Mixed Waste: Proposals for Dealing with Problematic Waste Streams and Policy Changes (PS76); - Current and Emerging US NRC LLW Regulatory Topics (PS77); - Agency Perspectives (US and International) on the Use of Risk and Dose Assessment Tools (PS80); - Progress on Deep Repository Programs Around the World (PS81); - US DOE Procurement and Contracting Opportunities (PS82); - Collaboration across Borders to Deliver Cleanup and Decommissioning Challenges (PS93a); - DOE Mission Success through Application of Practices and Solutions (PS94); - Challenges in Cask Certification - International Perspectives (PS96); - Robotics: Leveraging Innovations in Science and Technology to Benefit Safety Performance (PS98); - DOE Panel: DOE EM Featured Office - Portsmouth/Paducah Project Office (PS109); - Robotics: Lessons Learned from Industrial Robotics Deployment that can be Applied to the Nuclear Field (PS110a); - Roundtable: DOE Radioactive Management - Energy Facility Contractors Group - Waste Management Working Group (EFCOG WMWG) (PS118); - Thursday Topical Session 'Management of Used Nuclear Fuel: Will the United States Reengage and Align with International Efforts' (PS124); - Remotely Operated Vehicles for Emergency, Disaster, and Operational Upset Response (PS125)

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DOE OSTI2018

D and D of a Large Waste Treatment Plant in Germany - 18458

With its 40 years of nuclear experience, GNS, the world leading supplier of casks for spent fuel, ILW and HLW, also offers services for conditioning and management of all types of radioactive waste, especially for the German NPP. Since 1985, GNS treats and packages ILW and LLW from the operations and phase-out of German NPP in its Duisburg plant. To enable its permanent safe final disposal, solid wastes are compacted and dried before being shipped to an interim storage facility or - later - the Konrad repository. The total premises cover 20.000 m{sup 2} (215.000 sq ft), of which 9000 m{sup 2} (97.000 sq ft) is controlled area. The Duisburg Plant has three large industrial warehouses (each 25 m x 136 m or 82 ft x 446 ft) with a total capacity of 10.200 m{sup 2} or 110.000 sqft. For the treatment of open contaminated waste, six enclosures were built. These enclosures have a special ventilation system including a specific filtration plant. In one enclosure a high compaction plant was used, in another enclosures a drying facility for drums was in operation. During peak periods, more than 900 tonne of waste were treated annually. One enclosure was mainly used for maintenance, restoration and extension of mobile facilities, which were normally used for campaigns in NPP. Due to the decreasing amount of operational waste during the nuclear phase out in Germany, GNS has decided to close the Duisburg Plant. In the first half-year of 2017, the last waste packages were treated but around 80 20'-containers are still located on site. The D and D started in January 2017. The presentation will show the first experience of decommission processes of the first three enclosures. The following steps including the techniques used will be explained in detail: - Dismantling and removal of the high compaction facility; - Dismantling and removal of the drying facility; - Cleaning and wipe tests of the empty enclosures; - Removal of the stainless steel plates of the floor; - Investigations of the lower floor parts for contamination; - Removal of contaminated underground drainpipes; - Decontamination of the concrete floor; - Dismantling of the ventilation system in three enclosures; - Dismantling of the steel constructions. An important aim during the dismantling is to reduce the amount of nuclear waste and maximize the fraction material for free release. Next to the first three enclosures, in one of the enclosure IV - VI, a 150 tons scrap shear had been installed before the construction of the enclosure building. The scrap shear has to be dismantled and handled by heavy duty lifting equipment. Only afterwards is it possible to start the decommissioning of the last three enclosures. Our goal is to transfer this very large controlled area and the whole plant to the site owner, the Duisburg Port company, after 2 years and 5 months and doing so in a way so that the industrial warehouses can be used again for all possible types of industries. (authors)

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