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At least 847 records · Page 47
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Decommissioning an Active Historical Reactor Facility at the Savannah River Site - 13453

The Savannah River Site (SRS) is an 802 square-kilometer United States Department of Energy (US DOE) nuclear facility located along the Savannah River near Aiken, South Carolina, where Management and Operations are performed by Savannah River Nuclear Solutions (SRNS). In 2004, DOE recognized SRS as structure within the Cold War Historic District of national, state and local significance composed of the first generation of facilities constructed and operated from 1950 through 1989 to produce plutonium and tritium for our nation's defense. DOE agreed to manage the SRS 105-C Reactor Facility as a potentially historic property due to its significance in supporting the U.S. Cold War Mission and for potential for future interpretation. This reactor has five primary areas within it, including a Disassembly Basin (DB) that received irradiated materials from the reactor, cooled them and prepared the components for loading and transport to a Separation Canyon for processing. The 6,317 square meter area was divided into numerous work/storage areas. The walls between the individual basin compartments have narrow vertical openings called 'slots' that permit the transfer of material from one section to another. Data indicated there was over 830 curies of radioactivity associated with the basin sediments and approximately 9.1 M liters of contaminated water, not including a large quantity of activated reactor equipment, scrap metal, and debris on the basin floor. The need for an action was identified in 2010 to reduce risks to personnel in the facility and to eliminate the possible release of contaminants into the environment. The release of DB water could potentially migrate to the aquifer and contaminate groundwater. DOE, its regulators [U. S. Environmental Protection Agency (USEPA)-Region 4 and the South Carolina Department of Health and Environmental Control (SCDHEC)] and the SC Historical Preservation Office (SHPO) agreed/concurred to perform a non-time critical removal action for the In Situ Decommissioning (ISD) of the 105-C Disassembly Basin. ISD consisted of stabilization/isolation of remaining contaminated water, sediment, activated reactor equipment, and scrap metal by filling the DB with underwater non-structural grout to the appropriate (-4.877 meter) grade-level, thence with dry area non-structural grout to the final -10 centimeter level. The roof over the DB was preserved due to its potential historical significance and to prevent the infiltration of precipitation. Forced evaporation was the form of treatment implemented to remove the approximately 9.1 M liters of contaminated basin water. Using specially formulated grouts, irradiated materials and sediment were treated by solidification/isolation thus reducing their mobility, reducing radiation exposure and creating an engineered barrier thereby preventing access to the contaminants. Grouting provided a low permeability barrier to minimize any potential transport of contaminants to the aquifer. Efforts were made to preserve the historical significance of the Reactor in accordance with the National Historic Preservation Act. ISD provides a cost effective means to isolate and contain residual radioactivity from past nuclear operations allowing natural radioactive decay to reduce hazards to manageable levels. This method limits release of radiological contamination to the environment, minimizes radiation exposure to workers, prevents human/animal access to the hazardous substances, and allows for ongoing monitoring of the decommissioned facility. Field construction was initiated in August 2011; evaporator operations commenced January 2012 and ended July 2012 with over 9 M liters of water treated/removed. Over 8,525 cubic meters of grout were placed, completing in August 2012. The project completed with an excellent safety record, on schedule and under budget. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Retrieval System for Calcined Waste for the Idaho Cleanup Project - 12104

This paper describes the conceptual approach to retrieve radioactive calcine waste, hereafter called calcine, from stainless steel storage bins contained within concrete vaults. The retrieval system will allow evacuation of the granular solids (calcine) from the storage bins through the use of stationary vacuum nozzles. The nozzles will use air jets for calcine fluidization and will be able to rotate and direct the fluidization or displacement of the calcine within the bin. Each bin will have a single retrieval system installed prior to operation to prevent worker exposure to the high radiation fields. The addition of an articulated camera arm will allow for operations monitoring and will be equipped with contingency tools to aid in calcine removal. Possible challenges (calcine bridging and rat-holing) associated with calcine retrieval and transport, including potential solutions for bin pressurization, calcine fluidization and waste confinement, are also addressed. The Calcine Disposition Project has the responsibility to retrieve, treat, and package HLW calcine. The calcine retrieval system has been designed to incorporate the functions and technical characteristics as established by the retrieval system functional analysis. By adequately implementing the highest ranking technical characteristics into the design of the retrieval system, the system will be able to satisfy the functional requirements. The retrieval system conceptual design provides the means for removing bulk calcine from the bins of the CSSF vaults. Top-down vacuum retrieval coupled with an articulating camera arm will allow for a robust, contained process capable of evacuating bulk calcine from bins and transporting it to the processing facility. The system is designed to fluidize, vacuum, transport and direct the calcine from its current location to the CSSF roof-top transport lines. An articulating camera arm, deployed through an adjacent access riser, will work in conjunction with the retrieval nozzle to aid in calcine fluidization, remote viewing, clumped calcine breaking and recovery from off-normal conditions. As the design of the retrieval system progresses from conceptual to preliminary, increasing attention will be directed toward detailed design and proof-of- concept testing. (authors)

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DOE OSTI2012

The MARSAME Methodology: Fundamentals, Benefits and Applications - 12135

MARSAME is an acronym for the 'Multi-Agency Radiation Survey and Assessment of Materials and Equipment'. Published in January 2009, MARSAME was a joint effort between the U.S. Department of Energy (DOE), the U.S. Department of Defense, the U.S. Department of Energy, and the U.S. Nuclear Regulatory Commission (NRC) to aid sites in the clearance of materials and equipment (M and E). The MARSAME manual supplements the Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), published in 1997. As cited in the MARSAME, applicable M and E includes metals, concrete, tools, equipment, piping, conduit, and furniture. Also included are dispersible bulk materials such as trash, rubble, roofing materials, and sludge. Solids stored in containers, as well as liquids and gases, represent additional M and E. The MARSAME methodology covers multiple technical areas, including the initial assessment (IA), Measurement Quality Objectives (MQOs), survey approaches and considerations, survey plans, survey implementation, and Data Quality Assessment (DQA). These topics are generally captured under four phases of the data life cycle, which are Planning, Implementation, Assessment, and Decision-Making. Flexibility and a graded approach are inherent components of the MARSAME methodology pertaining to M and E property clearance programs. Because large quantities of M and E potentially affected by radioactivity are present in the United States, owners of the M and E need to identify acceptable disposition options. Thirteen disposition options, broadly defined under both release and interdiction scenarios, are described in MARSAME. Nine disposition options are listed for release; these options are categorized into two for reuse, two for recycle, four for disposal, and one that is essentially 'status quo' (i.e., maintain current radiological controls). Four interdiction options are also cited. To date, applications of the MARSAME approach for M and E property clearance under reuse scenarios have been limited. However, this should change as familiarity increases. Known examples are provided from DOE-contractor and NRC-licensee facilities. If the methodology is appropriately planned and implemented, the benefits of the MARSAME approach include worker and public protection, reduction in the amount of disposed radioactive waste, reuse of materials (resulting in environmental and material sustainability advantages), and cost savings. (authors)

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DOE OSTI2012

The Direct Path To WIPP - 12471

Sandia National Laboratories/New Mexico (SNL/NM), designated as a small quantity site (SQS) by the National TRU Program (NTP), generated contact-handled (CH) and remote-handled (RH) transuranic (TRU) waste primarily from the decontamination and clean-out of glove boxes at the Hot Cell Facility (HCF) at Technical Area (TA) V. All of the waste required repackaging, with the CH TRU waste being repackaged from late 2007 through 2011. Three shipments of CH were completed in October 2011, which de-inventoried SNL/NM's legacy TRU waste. In FY11, RH TRU waste was repackaged at the Auxiliary Hot Cell Facility (AHCF) located in TAV with the support of the Central Characterization Project (CCP). The waste was originally packaged in SNL/NM fabricated casks, cement or lead-lined 55-gallon drums, or 30-gallon drums. The AHCF is a small hot cell, with access only through a roof port which presented challenges for inserting and removing waste from the hot cell. The CCP provided visual examination operators (VEOs) to observe and document each waste item repackaged, removal of prohibited items, and radiological sampling. Dose-to-Curie measurements were calculated by CCP after a radiological report was prepared using scaling factors determined by the analysis of swipe samples. Finally, headspace gas samples were taken and sent to the Advanced Mixed Waste Treatment Project (AMWTP) for analysis. Despite the challenges, the RH waste is on track to be shipped to WIPP in early FY12. The processes used and procedures developed to conduct the repackaging operations, the issues identified and mitigated were challenging but the cooperation between SNL/NM and the Central Characterization Program (CCP) enabled SNL/NM to complete the repackaging and support the characterization and shipment. An inventory list, identification of the campaigns, discussion of the challenges and mitigations, and the final loading of the RH 72-B casks at TA-V for direct shipment to the Waste Isolation Pilot Plant (WIPP) will be discussed. Lessons learned from the RH campaigns are: - Some containers that were originally identified as HC-3 have been re-evaluated and became < HC-3 due to the conservative estimates made by the original generators - Operators at the AHCF were not accustomed to the detail required by the VE operators. However, they worked well together and the repackaging was completed ahead of schedule. - The AK was not always accurate as was demonstrated by the solid waste found in the drum during the first visit by EPA. That waste has since been determined to be low-level. - Two drums originally thought to be RH turned out to be CH and arrangement for RTR had to be made quickly. - Six of the original RH repacked drums became low level. - Lessons learned from the CH campaigns were helpful in avoiding many issues. The RH repackaging effort has been a success due to the expertise of the AHCF operators, supervisor, and manager, the conscientious attention to detail of the CCP VE operators, the experience of the CCP DTC and headspace gas sampling staff, and the guidance and support from CCP and CBFO. Sometimes schedules had to be adjusted, processes updated, and issues discussed, but the communication between CCP and SNL/NM was good. SNL/NM hopes to have the legacy RH TRU waste shipped off-site by early 2012. (authors)

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DOE OSTI2017

Long-Term Stewardship at Hanford - 17023

The Hanford Site's Long-Term Stewardship (LTS) Program is responsible for 220 square miles of the Site's River Corridor, which includes 1,527 waste sites, 6 Manhattan Project Era production reactors that have been placed in interim safe storage, and 46 miles of Columbia River shoreline. More than 24,000 cleanup and historic documents have been identified, indexed, and tagged in the LTS records and document libraries. The LTS program, operated by the Mission Support Alliance, LLC (MSA) manages and provides surveillance and maintenance of facilities and institutional controls and all associated monitoring to ensure continued protectiveness of human health and the environment. Since 2010, through collaborative efforts with DOE and its prime contractors, land in 14 geographic areas and six cocooned reactor facilities were transitioned (mid-contract) from the River Corridor Closure Contractor to the LTS program. This accomplishment relied heavily on Site contractors and DOE working together to address challenges. Other stakeholders were informed of LTS progress through the Hanford LTS web site, presentations, and briefings. Hanford's LTS program is responsible for ensuring the protectiveness of cleanup remedies and the management of institutional controls once cleanup objectives have been achieved. LTS accomplishes this by adhering to post-cleanup requirements specified in the Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA) and the Resource Conservation and Recovery Act of 1976 (RCRA) cleanup decision documents. The LTS program conducted inspection and surveillance and maintenance (S and M) activities of the cocooned 105-F Reactor in October 2014 and the other five reactors in 2015 and 2016. Reactor entries and internal inspections typically are conducted at 5-year intervals to assess the condition of the structures and evaluate potential deterioration of the reactor core, shield walls, and roof. The cocooning process is designed to protect the reactor for 75 years while radioactive decay continues, ultimately making the structures safe for demolition and removal. The results of the recent 5-year inspections allowed DOE to negotiate an extended 10-year inspection cycle, which led to a $5 million avoidance in lifecycle inspection costs. Hanford's LTS program is successfully shifting from a program focused on transitioning land and waste sites to a program focused on data management and S and M activities for those buildings and waste sites within the program. This paper will highlight the accomplishments and collaborative efforts in addressing the challenges faced by Hanford's LTS program. (authors)

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DOE OSTI2017

Consolidated Nuclear Security, LLC, Excess Facility Disposition Program - 17289

Consolidated Nuclear Security, LLC (CNS), the management and operating contractor for the Y-12 National Security Complex (Y-12) in Oak Ridge, Tennessee, and the Pantex Plant (Pantex) in Amarillo, Texas, has developed an integrated strategy to address badly deteriorating excess facilities presenting risks to people, the environment, and the National Nuclear Security Administration (NNSA) mission. The comprehensive, integrated Excess Facilities Disposition Program (EFDP) incorporates NNSA scope for excess facility risk reduction, preparation of facilities to be transferred to the U.S. Department of Energy (DOE) Office of Environmental Management (EM) for demolition, and demolition of non-process contaminated facilities. The Program has been developed using the NNSA Office of Safety, Infrastructure, and Operations (NA-50) Program Management Plan as a guide. The Program has undertaken actions to address the most immediate risks associated with CNS excess facilities. At Y-12, new foam roofs have been installed on Buildings 9201-5 (Alpha 5), 9204-4 (Beta 4), and 9206 to minimize further facility degradation. Tanks and dikes outside of Alpha 5 and Beta 4 are also being disconnected, drained, and filled with concrete. In addition, Beta 4 will be getting a new electrical system as its Manhattan Project-era system will be replaced by a temporary 'construction power' setup that will provide as-needed electricity to the building that is safer, more reliable, and will ease demolition activities in the future. The EFDP is also focusing on the basement of Alpha 5 and is planning to remove the water from the aging facility's flooded basement, which in some areas contains up to 54 inches of water. The EFDP is also planning several efforts to remove excess materials from Alpha 5 and Beta 4, de-inventory the Alpha 5 Mercury House System and numerous transformers, and de-inventory Beta 4 hydraulic pumps and systems. Efforts are also underway to deactivate and decommission non-process-contaminated facilities at Y-12 and at Pantex. The EFDP has prioritized the performance of work to reduce risk while at the same time supporting the preparation of process-contaminated Y-12 facilities for transfer to EM. The Program aligns with the EM goal to have the NNSA scope completed in Alpha 5 and Beta 4 by the end of FY 2020, and Building 9206 completed by the end of FY 2024. These dates and project sequencing will support the Y-12-related DOE EM cleanup initiatives. (authors)

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DOE OSTI2017

Integrated Wash-Aid, Treatment, and Emergency Reuse System (IWATERS) for Strontium Contaminations - 17390

The Integrated Wash-Aid, Treatment, and Emergency Reuse System (IWATERS) is being developed to mitigate the release of radioactivity to the urban environment by disseminating, collecting, and processing decontamination wash water. Previously, IWATERS was designed for the treatment of surfaces contaminated with radioactive cesium. Experiments were performed to investigate the expansion of IWATERS for strontium contamination. Bench-scale tests were performed to investigate various counter-ions and chelators to promote ion exchange of strontium from building surfaces. Desorption results for Cs, Sr, and Eu from concrete aggregate and concrete, brick, and asphalt coupons using low- and high-pressure flow with Ca{sup 2+} and Ba{sup 2+} rich wash water are reported along with those from non-porous surfaces such as glass and vinyl siding. Several sorbents were tested for their ability to specifically separate Sr{sup 2+} from these high ionic strength wash waters. Alkali metal and alkaline earth salts were found to be effective counter-ions for decontamination of Sr{sup 2+}, which may facilitate the use of common sorbents, even if their K{sub d} values are <100 mL/g. In practice, low K{sub d} values require the use of large quantities of sorbent. Computer simulations were run using the GoldSim Contaminant Fate Module software package to optimize the efficiency of IWATERS for strontium and account for the technical limitations of the sorbents. This modeling allows greater flexibility to design the wash water recycling system for various end-user specifications including options from rain barrel designs for treating runoff from roofs and facades to larger options like drainage basins for larger-scale wash down activities. (authors)

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DOE OSTI2015

Long-term stewardship - surveillance and maintenance at Hanford - 15155

In FY 2015, Hanford's Long-Term Stewardship (LTS) Program is tasked to complete the required entries into five cocooned plutonium production reactors (105-C, 105-D, 105-F, 105-H and 105-N), which sit adjacent to the Columbia River. The first reactor entered in October 2014, the cocooned plutonium production reactor, 105-F, signaled the LTS Program's entry into the Surveillance and Maintenance (S and M) era. Reactor entries at Hanford are conducted at 5-year intervals to assess the condition of the structures and evaluate potential deterioration of the reactor core, shield walls and roof. The cocooned structures are designed to protect the reactor core for up to 75 years while radioactive decay continues, ultimately making the structures safe for demolition and removal. The 105-F entry demonstrates that the LTS program at Hanford is shifting from a program focused on transitioning cleaned up land and waste sites to a program focused primarily on S and M activities. The entry process is carefully planned, choreographed and monitored to ensure safety of the workers and minimize the time the reactor structure is opened. The process includes controlled opening of the reactor, air quality monitoring, structural safety inspections, and radiological clearance. Once the structure is cleared for entry, assessments of the structural integrity and radiologic conditions are completed. Evaluation of the interior temperature and flood monitoring equipment is conducted. Facility visits by regulators and DOE personnel close out the process prior to welding the door closed until the next assessment, five years later. This paper will document execution of the entry work package, identify challenges overcome, and provide video documentation as reference material for future workers and the public. The current LTS program has responsibility for approximately 100,000 acres, over 200 waste sites and six cocooned reactors. Just over 5,600 documents have been identified, indexed and tagged for storage in the LTS document library. The program has successfully completed seven consecutive transitions since July 2011, with six more transitions planned in the next 18 months to support the U.S. Department of Energy Richland Operations Office's (DOE-RL) near-term cleanup objectives to significantly reduce the footprint of active cleanup operations for the River Corridor. (authors)

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DOE OSTI2015

Mock-Up activities for the extraction of samples from a salt-stone disposal unit - 15502

Mock-up testing with respect to the core-drilling of simulant salt-stone grout was conducted to support the future retrieval of field-emplaced salt-stone samples from salt-stone Disposal Unit (SDU) Cell 2A. The primary objectives of extracting samples from SDU Cell 2A are to enable measurement of field-emplaced samples with respect to properties that influence long-term performance, such as saturated hydraulic conductivity, and to compare the measured properties for field-emplaced and comparable samples prepared in the laboratory. Initial mock-up activities (Phases 1 and 2) were conducted in Fiscal Years (FY) 2013 and 2014 in order to demonstrate the viability of standard core-drilling techniques that are typically employed for sampling concrete structures. The primary requirement for the core-drilling technique was that it should minimize the potential of physical or chemical sample changes that would artificially degrade key grout properties. Mock-up activities were aimed at optimizing the techniques for core-drilling, and subsequent sample extraction, using a set-up that mimicked, to the extent practical, both the physical characteristics, and operational complexities, of core-drilling salt-stone grout from the roof of the SDU. Samples extracted from the grout monolith were subsequently characterized with respect to saturated hydraulic conductivity and compressive strength, and compared to data for cast samples prepared using the same simulant salt-stone material. The comparison of these properties for the two sample types was expected to highlight potential flaws (e.g., micro-cracking) introduced by the core-drilling process. While the data for saturated hydraulic conductivity were virtually identical (approximately 2 E-09 cm/sec), the compressive strength of the core-drilled samples (approximately 3.5 MPa (500 psi)) was on average half that measured for the cast samples. It was suspected that increased water saturation and surface scoring of the core-drilled samples might account for the disparate compressive Strength data. During Phases 1 and 2 it was determined that cores could be successfully drilled and extracted to a depth of approximately 0.9 meters (3 feet) below the grout surface. Phase 3 (FY 2105) has been initiated and is primarily directed at demonstrating the ability to core-drill through approximately 1.8 meters (6 feet) of overlying grout to retrieve the desired sample from the SDU. The original drill depth of 0.6 - 0.9 meters (2-3 feet) was based on an SDU fill height of 5.6 meters (18.5 feet) and sample location of approximately 4.9 meters (16 feet); recently the allowable fill height in the SDUs has been increased to 6.5 meters (21.25 feet); hence the increased depth of overlying grout above the samples to be retrieved. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Design and Construction of a Loess-Cement Cushion as an Integral Component of an SL-LILW Repository - 16077

The Bulgarian State Enterprise for Radioactive Waste Management (SERAW) is planning the construction of a near-surface repository for short-lived low and intermediate level radioactive waste, the National Disposal Facility (NDF), adjacent to the Kozloduy Nuclear Power Plant (KNPP). The NDF will be constructed as a reinforced concrete vault-type repository system, where prefabricated cubic waste packages (2 m on a side) will be emplaced in large reinforced concrete disposal cells. Waste emplacement operations will be conducted in three phases over a total of 60 years of operations. After the last waste packages have been disposed, an engineered multilayer cover will be installed over the site to further isolate the waste from the accessible environment. The Radiana Site where the NDF will be constructed is a sloping area located between the second and sixth loess terraces above the River Danube with elevations ranging from approximately +40 m to elevation +95 m. The unsaturated zone at the location selected for waste disposal ranges from 13 m to 16 m in thickness. After construction, the unsaturated zone will consist of a 5 m thick, engineered loess-cement cushion installed over naturally occurring Pliocene clays and sandy clayey sediments of the Brusarci Formation with a thickness above the top of groundwater ranging from 8 m to 11 m. The loess-cement cushion is a key component of the 'engineered fill' barrier as identified in the Intermediate Safety Assessment Report (ISAR) for the NDF. This barrier consists of engineered fill both below the disposal cells, i.e., the loess-cement cushion, and above the disposal cells, i.e. the final multilayer cover to be installed at closure of the facility. The loess-cement cushion component of this barrier will act to provide a geo-technically stable foundation for the disposal system. In addition, the barrier will raise the base of the disposal cells and increases the unsaturated zone thickness above the top of the groundwater table, as well as to serve as a chemical barrier that will act to sorb radionuclides thereby retarding their migration to the underlying groundwater. In addition to its assigned barrier functions the loess cement cushion also houses key embedded components required as part of the overall NDF safety concept, including systems necessary for operational and post-closure performance monitoring as well as the support foundation structures required for the mobile roofs, which are designed to handle and protect the waste during emplacement operations. The post-closure monitoring period, i.e., the period of institutional control, is given by as 300 years. While similar although less massive cushions have been previously constructed as foundations for various structures at the KNPP, the loess-cement cushion planned for construction at the NDF including the embedded structures will represent a first of its kind construction in terms of both its physical dimensioning as well as its requirements for long-term performance. This paper describes the design of the cushion, its functions and its construction as planned for the NDF as an integral component of the site safety concept. (authors)

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DOE OSTI2016

Design of Removable Shield Modules for Nuclear Plant Applications Using Monte Carlo Modeling - 16264

A new modular shielding system has been designed and developed for nuclear plant applications as a replacement for traditional concrete blocks and plugs. The shield comprises interlocking modules that are custom designed using CAD and Monte Carlo modelling methods. The key benefit of these new modules is that they can be quickly removed and re-installed to allow temporary access to high radiation environments e.g. for maintenance and repair of plant equipment at nuclear power plants or waste handling facilities. The traditional method for construction of these removable concrete blocks and plugs was to cast them in place. Within the aging fleet of commercial reactors, some plants are experiencing concrete degradation of these shield blocks and plugs as well as failure of lifting points. Additionally, the existing plugs are difficult to remove and re-install. The new modular system comprises a rigid steel enclosure filled with lead that can be installed in a fraction of the normal time. This approach can accelerate outage times leading to reduced dose uptake to operators and provide significant long term cost savings. Use of Monte Carlo methods enables the shielding modules to be designed efficiently for neutron and gamma ray attenuation with the minimum necessary thickness of lead and steel. Furthermore, the modelling ensures the design is not compromised by streaming pathways, for example at the overlaps between modules or through internal structural components. Shielding modules were supplied to the Brunswick Nuclear Power station (BNP), operated by Duke Energy, located in southeast United States. These modules have been installed to replace concrete shield plugs that are located on the roof above a Feed Water Heaters (FWH) bay. While specific in application, this project demonstrates that the shield module design concept could be expanded to address vertical block walls routinely found in personnel and equipment access hatches. This approach is applicable to a wide variety of new and existing nuclear facilities and will lead to improved safety and reduced lifetime costs all the way through to future decommissioning and dismantling operations. (authors)

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DOE OSTI2016

Managing Low-Level Waste at Dounreay - 16269

Dounreay Site Restoration Limited (DSRL) is the site licence company responsible for the clean-up and demolition of Britain's former centre of fast reactor research and development. The Dounreay site requires facilities to dispose of up to 175,000 cubic metres of low level waste (LLW) that has been and will continue to be generated during the decommissioning and restoration of the site. In May 2014, construction of two new vaults for the disposal of LLW at Dounreay was completed, including the necessary roads, infrastructures and services, followed by an encapsulation plant in September 2014. The vaults are located immediately to the northeast of the Dounreay licensed site, on NDA-owned land. The facilities are for disposal of the solid LLW generated at Dounreay and the adjacent Vulcan site only. In 2004, following consultation with stakeholders and members of the public, Dounreay issued the 'Best Practicable Environmental Option (BPEO) Report' on the management of Dounreay LLW. This report assessed potential management options for LLW against a range of criteria, including technical, environmental, cost, health and safety issues. The study was conducted in line with best practice and involved consultation with a range of people, including the local community and organisations with an interest in radioactive waste management. The recommendations and conclusions from the study provided the basis for the 'Dounreay Solid LLW Overall Strategy', which was published in March 2005. A fundamental component of this strategy was the development of new below-ground disposal vaults for LLW at Dounreay. Siting the vaults on land at Dounreay avoids any need to transport the LLW away from Dounreay on public roads. Disposal at Dounreay therefore satisfies the proximity principle of managing the waste at source. Disposal at Dounreay is also consistent with UK Government Policy on LLW management, the Scottish Government's Radioactive Waste Management Policy and the Nuclear Decommissioning Authority's (NDA's) strategy on management of LLW from the UK nuclear industry. A planning application was initially lodged with the Highland Council in June 2006. The period for the determination of this application was extended to allow the Scottish Environment Protection Agency (SEPA) sufficient time to assess fully the supporting Environmental Safety Case (ESC). The planning application was re-activated in May 2008, and planning permission was granted by Highland Council in April 2009. The planning permission included 26 conditions on DSRL, covering a wide range of environmental issues such as construction noise/dust, traffic, visual impact and also a condition to develop a community fund aimed at socio-economic development of the area around the Dounreay site. After undertaking the necessary detailed design work, construction of the phase one vaults commenced in November 2011 by principal contractor, Graham Construction. Work was undertaken on both vaults simultaneously, excavating around 243,000 m{sup 3} of rock. When excavation was complete, steel reinforcement was placed and concrete poured to construct the floor slabs. This was followed by the construction of the reinforced walls and then the installation of the steel support structure and materials that form the roof. Mechanical and electrical installation then followed, in parallel with the construction of the vault apron, ramps and access roads. A pumping system and control building was also constructed to remove groundwater from around the vaults to allow the vaults to be kept dry while they are being filled with waste. A layer of excavated material, approximately 3 m thick, was placed across the existing ground surface between the vaults and the coast, to ensure the water table lies well below the new ground surface after closure of the vaults, and that any future trace contaminated groundwater flows out to the sea rather than to the soil zone. This layer is referred to in project documents as the 'enhanced geosphere'. After disposals are completed and the vaults are closed, this layer will be joined to the capping over the vaults so that the ensemble merges visually with the surrounding landscape. The original vegetation and soil mix will be returned as far as is practicable. An encapsulation plant has been constructed to fill the LLW containers with grout prior to disposal in the vault, similar to the process used at the national Low-Level Waste Repository facility (LLWR), near Drigg in Cumbria. The plant has the capacity to grout up to four containers per day. The grout fills voids in the container and helps provide chemical conditions in the vault that limit the release of radionuclides in the far future, after the vault has been penetrated by groundwater. On April 24 2015, active commissioning began at the Encapsulation Plant and LLW Vaults. Half-height ISO containers packed with low level waste are currently being filled with grout and disposed of in the LLW Vault. Also packaged Demolition Low Level Waste (DLLW) is being processed and disposed of in the Demolition Low Level Waste Vault. To date over 2500 packages of DLLW and 70 containers LLW have been disposed of. Completion of construction and the start of operating the Phase One vaults and encapsulation plant are fundamental to the successful and efficient decommissioning of the Dounreay site, upon which many millions of pounds of investment and hundreds of jobs rely.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Keep It Smart and Simple - Henriette - The Advanced Designed Drum Gripper for Remote Handling Tasks - 19327

A power manipulator has many applications in the handling, treatment and packaging of radioactive waste. In high radiation environments, where humans cannot work safely, a power manipulator is essential to perform difficult and heavy tasks. For more than 40 years, Waelischmiller has provided power manipulators to the nuclear market. Specifically in a high radiation nuclear environment, electrical driven manipulators are preferred. The basic design of a power manipulator is divided into the following parts: Bridge, carriage, telescopic mast and manipulator arm. All parts are customizable in length, width and capacity. The standard power manipulator version is equipped with a manipulator arm. For the nuclear waste disposal industry, there is a special need for a drum gripper. Drums are used for packing low, medium or high activated material and for storage. After packing, the drums must be handled remotely. For both tasks, a power manipulator with a drum gripper is needed. For this project KTE (Kerntechnische Entsorgung Karlsruhe GmbH) required a total refurbishment of equipment in a cell used for handling, sorting and packaging radioactive waste drums. The power manipulator was obsolete and provided low availability and through put. This cell was an essential asset in their decommissioning process and was impacting KTE's ability to meet their volume objectives. Our challenge was to develop a new system with few details from the original cell construction. The new manipulator must meet all capabilities of the old system and in addition have the ability to achieve 180 degree tilting of the drums. An additional challenge was the requirement to use the existing rails in the cell which were very close to the roof structure which required a special design. Our project, named internally Henriette, is the first of its kind. The drum gripper is based on the power manipulator A1000 series. The biggest challenges in this project were to adapt the standard components of the power manipulator to the non-standard type of gripper to have 180 deg. tilting of the drum and a handling capacity of 1200 kg in the whole working range. The new manipulator fits to the old rails and is mountable in a hot cell with restricted access and can operate with the other A1000 manipulator which is located in the cell. The result is a system with no external wiring, maintenance-free and easy to decontaminate. Other features include a 5 t auxiliary hoist on additional and parallel carriages, emergency drives for bridge and carriage, emergency drive for telescope turn movement and emergency lifting unit for telescope. A tilt and pan holder for camera on the telescope and external mechanical emergency drives on the gripper were also essential. This paper will address the complexity, challenges and engineering ingenuity required to replace very old and obsolete equipment with a state-of-the-art system within very limited physical space and within a high radiation environment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Thermal Analysis for SRS Saltstone Disposal Unit Facility

Saltstone Disposal Units (SDU) are long term disposal units for low activity grout produced from the solidification of decontaminated non-hazardous salt waste at SRS. The unit is a cylindrical disposal vault of 150 ft. in diameter and 23 ft. in height. The total volume the SDU can hold is 3 million gallons of grout. The hydration process of the grout material produces heat from an exothermic chemical reaction. If the grout reaches temperatures higher than the maximum safety temperature of 95 deg. C, cracks can start forming leading to contamination of the environment. The purpose of this study is to develop the optimal baseline domain boundary conditions and to perform a sensitivity analysis with respect to the baseline modeling conditions. The following assumptions were made for the modeling calculations: - The system geometry is axisymmetric; - The system was at steady state; - Ambient temperature was set to 100 deg. F; - Air was assumed to be an incompressible ideal gas; - Radiation was neglected for a conservative estimation; - Natural convection for the SDU vapor space is assumed to be turbulent because Rayleigh number based on typical operating conditions is higher than 10{sup 8}; - Water tables were not taken into consideration. The following sensitivity analyses were performed to find the optimal domain boundary: - Mesh size sensitivity; - Soil depth sensitivity - Soil depth varied from 30 ft to 250 ft with a soil thickness of 90 ft; - Soil thickness sensitivity - Soil thickness varied from 30 ft to 270 ft with a soil depth of 180 ft; - Soil conductivity sensitivity - Soil thermal conductivity varied from 0.25 W/m-K to 1 W/m-K; - Ambient Temperature Sensitivity - Ambient temperature varied from 70 deg. F to 110 deg. F. For the thermal calculations, the temperature of the bottom of the soil was set to 17.4 deg. C. The heat flux on the roof of the building and surface of the earth is 163.2 W/m{sup 2}. The soil sides are set to be adiabatic. The sensitivity analysis shows that the mesh size has no effect on the numerical results. This is mainly due to a small convection zone and large conduction zones in the modeling domain. The optimal domain boundary was found to be a soil thickness of 90 ft and a soil depth of 150 ft. The different sensitivity analysis performed here shows the optimal boundary domain to be used for thermal calculations is a soil depth of 150 ft beneath the SDU and a soil thickness of 90 ft. Anything larger would produce similar solutions but at a longer computational time to solve. The maximum temperature was more sensitive to soil properties as shown here. While the ambient temperature affected the max temperature linearly. Most of the heat transfer takes place downward vertically between the SDU and the soil beneath it. Using the optimal domain boundary determined by the sensitivity analyses, the maximum thermal loading to keep the maximum grout temperature below 95 deg. C is 1.646 W/m{sup 3}.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

VENTSAR V3.0: A Python GUI for Estimating Contaminant Concentrations on or Near Buildings Due to Building Effects and Plume Rise and For Calculating Inhalation and Plume Shine Doses

VENTSAR: Originated as VENTAX (Smith and Weber 1983) on the IBM Mainframe at the Savannah River Site (SRS) as a Fortran program. Updated to VENTSAR XL V1.0 (Simpkins 1997) as a spreadsheet version using macros created in Microsoft Excel. Later updated to VENTSAR XL V2.0 (Dixon 2018) as a spreadsheet executing a modified version of the original VENTAX Fortran code. Estimates contaminant concentrations on or near a building from a release at a nearby location. Calculates concentrations for a given meteorological exceedance probability or for a given stability and wind speed combination. Can model a single building with or without a penthouse on top or a ground location from either a stack or ground release. Plume rise can be considered. Contaminant releases can be chemical or radioactive with downwind concentrations determined at user-specified distances. Wind passing over and around buildings creates a complicated dispersion pattern. Air-intake vents may be located on building roofs or near the ground downwind of a release source. An estimation of pollutant concentrations on or near a structure is important in determining expected pollutant levels. Meteorological data are selected based on a specific area of the site. Fortran-based VENTAX able to make fast, complex mathematical calculations. Not user-friendly VENTSAR XL V1.0 no longer supported due to obsolete Excel macros. VENTSAR XL V2.0 an interim solution using an Excel spreadsheet to interface with the modified VENTAX Fortran code. Requires user to have Microsoft Excel. Not an intuitive interface for someone unfamiliar with VENTSAR. Does not perform dose calculations. VENTSAR V3.0 goal to combine the strengths of previous versions of VENTSAR into a single, powerful yet user-friendly program with a Graphical User Interface (GUI). Not dependent upon a specific program or plug-ins. Self-contained package to be used on any computer running Microsoft Windows. User-friendly, intuitive GUI created in Python for parameter input. Executes same modified VENTAX Fortran code as VENTSAR XL V2.0. Outputs formatted text file of completed calculations for easy review and dissemination. Performs inhalation and plume shine dose calculations for up to 11 user-selected nuclides from a nuclide dose factor library of almost 500 nuclides. 12 test cases were created for verification of V1.0 and V2.0. Same test cases run in V3.0 to verify correct performance. V3.0 produced similar results to V1.0. Confirmed GUI did not alter calculations in any way. Comparisons of the test case results confirm the V3.0 GUI does not influence the VENTSAR calculations. Simply passes same input parameters to VENTAX Fortran code for execution. Provides end user an easy-to-use, intuitive tool to quickly make building effect and plume rise calculations, as well as, inhalation and plume shine dose calculations. No experience with or working knowledge of Fortran, Excel, command line, or macros required. Self-contained package allows VENTSAR be deployed to any Windows computer without the need for specific programs or plug-ins.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

History of Ultrasonic Inspection of High-Level Waste Tanks at SRS - 20220

The Liquid Waste Disposition Projects (LWDP) Structural Authority and Inspection Engineering Support Group has teamed with the Materials Science and Technology Non-destructive Evaluation Group of Savannah River National Laboratory (SRNL) to implement a comprehensive In-Service Inspection (ISI) program for the High-Level Waste (HLW) tanks in the sites tank farms. The purpose of this program is to demonstrate the structural and leak integrity of the tanks to maintain the function of waste containment throughout the desired service life. Four types of waste tanks were constructed at Savannah River Site (SRS) from 1951 thru 1981 to store radioactive nuclear waste from the separations process. The Type I, II and III are carbon steel cylindrical tanks with varying degrees of access to the side walls and very limited access to the tank bottom. The tanks contain 750,000 to 1.3 million gallons of waste and are buried or backfilled with soil. The tanks are 75'-85' in diameter and 24'-33' from tank bottom to the tank roof. Inspection of the HLW tanks is made difficult by radiation, contamination, and small access openings. Through the years these challenges have been overcome by progressive improvements to programs and equipment resulting in the sophisticated wall crawler Ultrasonic Testing (UT) technology of today. A formal inspection program was developed and put in service in 1972. UT thickness measurements were first made in 1967 and 1969 of selected waste tanks using an analog-type instrument to measure the thickness of the carbon steel walls. Routine inspections for thinning were performed from 1972-1985 and then discontinued because no thinning had been observed and this technique could not detect pitting or cracking. UT inspections resumed in 1994 using a sophisticated robotic wall crawler with UT technology capable of detecting thinning, cracking and pitting. The current program focuses on the newer Type III tanks that are used primarily for storage. The program was upgraded in 2003 to include selected regions of the tank wall, secondary wall and annulus floor on a 7 to 10-year frequency. The facility is currently in its third cycle of tank inspections. Thus far, there have been no reportable indications of service induced degradation. This result provides assurance that the chemistry control program has been effective at mitigating corrosion and that the tanks remain structurally sound even after more than 40 years of service. This paper will present the UT technology first used at SRS and the evolution to the current systems and their capabilities. Additionally, plans for equipment upgrades that will improve the extent of inspection will be discussed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Atmospheric {sup 41}Ar Plume Tracking of the UF Training Reactor Using a Large-Volume NaI Detector

The ability to track and monitor gaseous radioactive materials is important in the event of a radioactive release such as from a nuclear power plant accident, weapons testing, or other unintended release scenarios. The International Monitoring System (IMS) is currently in place to monitor for weapons testing and has been used to monitor nuclear power plant accidents. However, only 80 radionuclide stations for the IMS exist in the world and are stationary, limiting the resolution of any radioactive plumes detected. Furthermore, the IMS radionuclide stations are designed to be sensitive specifically to atmospheric radio-xenon and no other radionuclides. Historically, monitoring radioactive releases has been limited to the detection of radio-xenon and radio-krypton, both arising as fission products from nuclear fuel. After the accident at Three Mile Island Unit 2, scientists were interested in finding the dose the public received from released radioactive material. This was achieved by analyzing radiation monitors in the exhaust stack and filters within the reactor building. furthermore, a series of 20 thermoluminescent dosimeters (TLD's), 8 air sampling station positioned around Three Mile Island, and a helicopter survey which used a NaI(Tl) detector array provided data on radionuclides, specifically {sup 137}Cs, that escaped the confines of the reactor building. The TLD's and air sampling stations were stationary, and therefore limited the resolution of the radioactive release plume. The helicopter survey using the NaI detectors was limited by the technology available at the time. The NaI detector array output was compressed from 1024 channels to 256 channels, limiting the energy resolution of the spectral data. The helicopter used a microwave ranging system to establish position. Since this survey was taken in October of 1982, advancements in technology allow spectral data to be saved in its full 1024 channel format and the advent of GPS can give a more accurate account of the detector position. An updated mobile system for tracking a radioactive release would help to alleviate these issues as well as gain the ability to track a radioactive release in real-time. We investigate the feasibility of mobile tracking through study and measurement of the {sup 41}Ar plume released from the University of Florida Training Reactor (UFTR) research reactor. We use a large-volume NaI(Tl) detector placed on the roof near the reactor. This detector may also be placed and moved on a cart to survey the area surrounding the reactor. We compare the spectral data taken when the reactor is running at full power to the background to analyze the effectiveness of the detector at identifying atmospheric {sup 41}Ar. Finally we discuss the implications of these measurements with regards to validation of plume models and the capability to track releases over a localized area from the source. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Simulation of the Prompt Dose Environment in the NIF during High Yield Shots

The National Ignition Facility (NIF) at Lawrence Livermore National Laboratory is the world's largest and most energetic laser system for inertial confinement fusion. NIF is a 192-laser beam facility, capable of producing 1.8 MJ, 500 TW of ultraviolet light. The main laser systems are installed in two laser bays. Each laser bay delivers 96 beams into one of two switchyards. Sets of turning mirrors in each Switchyard (SY) redirect the beams into the Target Bay (TB). Additional sets of optics in the TB will align and focus the beams onto the target in the center of the Target Chamber (TC). The 192 beams enter the TC through 48 indirect-drive beam ports. Laser beams are then focused to enter a small metal can (hohlraum) held at the center of the TC creating thermal x-rays that heat the surface of the fusion capsule contained within. The Target Bay has a cylindrical design with an inner radius of 15.24 m, 1.83-m-thick concrete walls and 1.37-m-thick concrete roof. A total of 239 and 121 penetrations are present in the Target Bay and switchyard walls, respectively. Figure 1 shows the facility's layout. There are seven floor levels within the TB at elevations of -33'9'', -21'9'', -3'6'', 17'6'', 29'6'', 40'0'', and 50'6'' with respect to the ground level. The TC is positioned in the middle of the TB with its center at elevation of 23'0'' from the ground level. The NIF will routinely utilize deuterium-tritium (D-T) targets with up to 20 MJ yield per shot, and maximum annual yield of 1200 MJ. This summary focuses on analyzing the expected prompt dose environment during the NIF's high yield phase of operations. Previous publications evaluated the expected doses associated with lower yield phases of operation. During shots, the Target Bay, Switchyards, and Laser Bays are interlocked and unoccupied. On the other hand, Personnel can be present inside the Operations Support Building (OSB), the control room, and the visitor center during the shots. The radiation shielding has been optimized to meet the NIF radiological design goals and the Department of Energy (DOE) limits. The goal is to limit the maximum prompt dose in any potentially occupied area to less than 50 {mu}Sv per shot as well as limiting the annual dose to individuals in the occupied areas to less than 1 mSv.

61 RADIATION PROTECTION AND DOSIMETRY↗