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DOE OSTI2016

One System Overview - 16579

The U.S. Department of Energy (DOE), Office of River Protection (ORP) is responsible for managing and completing the River Protection Project mission, which is to 'safeguard the nuclear waste stored in Hanford's 177 underground tanks, and to manage the waste safely and responsibly until it can be treated in the Waste Treatment and Immobilization Plant for final disposition'. The Office of River Protection's work scope is currently performed through two primary contracts: - The Tank Operations Contract, held by Washington River Protection Solutions LLC, (WRPS) includes the construction, operation, and maintenance activities necessary to safely store, retrieve, and transfer tank wastes; provide supplemental tank waste treatment; provide storage and/or disposal of primary and secondary waste streams; design, construction and commissioning of the Low-Activity Waste Pretreatment System; and support ORP to integrate Hanford Tank Waste Treatment and Immobilization Plant (WTP) completion and startup with Tank Farms operations. - The WTP Contract, held by Bechtel National, Inc. (BNI), includes the design, construction, and commissioning of a Pretreatment Facility, two vitrification facilities (one for high-level waste and one for low-activity waste), a dedicated analytical laboratory, and supporting facilities to treat radioactive tank wastes by immobilizing them into a stable glass waste form. One System was established as a means to improve the integration between the tank farms and the WTP project to support its safe and successful startup. One System's scope and purpose focus on mission integration consistent with DOE's plans to treat tank waste in phases, starting with direct feed of the low-activity waste stream to the WTP's Low-Activity Waste (LAW) Vitrification Facility (referred to as DFLAW). The One System organization now has four major focus areas: - Flowsheet Integration? Integrated River Protection Project (RPP) flowsheet development and management; developing a gaps, risks, and opportunities management plan and technical roadmap; and waste feed qualification. - Mission Analysis and Planning? Strategic planning, system planning, and modeling integration. - WTP Startup, Commissioning, and Operations Integration? Interface Control Document (ICD) management, coordinating facilities startup-to-commissioning-to-operations, contract directives and standards identification, and DFLAW program integration. - Project Integration and Controls? Integrated schedule, metrics/dashboards monitoring, and One System risk register. ORP established the DFLAW Program Office in February 2015 as one of its key performance goals. ORP staff, along with contractor managers, have identified roles and responsibilities for the each of the four major focus areas. One System has established an Executive Council and Governance Board to drive program progress and make decisions on DFLAW. This approach allows appropriate government and contractor senior management involvement in the DFLAW Program. Monthly critical path analysis on the integrated DFLAW schedule is being performed. Interface control documents specific to DFLAW have been developed, and system planning and modeling tools are being used to help define the DFLAW flowsheet. (authors)

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DOE OSTI2016

The Evolution of Chemical Waste Treatment in the Nuclear Industry Simple Solutions for Complex Problems - 16694

The U.S. nuclear waste treatment industry was born in the 1980s and early 1990s with the end of the Cold War, the peak of commercial nuclear power, and after implementation of federal regulations that drove new disposal requirements for radioactive waste generators. New regulations resulted in soaring costs for radioactive waste disposal, and thus an industry was born that focused on recycling, volume reduction, and waste form improvement. New treatment facilities and technologies were started that primarily dealt with low-level radioactive wastes; however, there was little to no treatment capacity for wastes commingled with hazardous chemicals. Large volumes of 'mixed wastes', of which many became known as 'orphan wastes', were being stored on U.S. Department of Energy (DOE) sites with no path for disposal and thus posed a risk to workers, the public, and the environment. Perma-fix Environmental Services, Inc. (Perma-Fix), which started in the hazardous waste industry in 1990 and had developed facilities and technologies to treat hazardous wastes, evolved in the late 90's to meet this mixed waste market opportunity. By adapting our technologies and collaborating with waste generators to understand their needs, we successfully developed treatment and disposition solutions for some of the most challenging mixed wastes in the U.S. By establishing a pathway for treatment and disposal of orphan wastes, Perma-Fix played an important role in the safe treatment and disposition of orphan wastes, which facilitated closure of DOE sites such as Rocky Flats, Mound, and Fernald. We found that understanding chemistry and having experience in managing the hazardous constituents of mixed wastes allowed us to apply solutions that were not overly complex and expensive. Further, by adapting proven hazardous waste treatment solutions to wastes commingled with contaminants such as polychlorinated biphenyls (PCBs), mercury, and organics we were able to safely manage even more problematic mixed wastes. During the 1980's, commercial low-level radioactive waste treatment and disposal solutions were being developed. At the same time, the federal government began dealing with increasing stakeholder demands for permanent solutions of more complex radioactive wastes such as high-level waste (HLW) and transuranic (TRU) wastes. HLW and TRU waste disposal became a contentious political issue between the federal government, regulatory agencies, and the public. As a result, DOE was mandated to develop treatment and disposal methodologies to address these wastes. DOE's Waste Isolation Pilot Plant (WIPP) facility solved most issues with disposition of TRU wastes; however, for HLW (for example tank wastes at Hanford, Savannah River, and Idaho) the complexity of the wastes and the fact that government entities were developing political solutions for waste treatment resulted in a variety of complex and expensive solutions. Decades later we still do not have a permanent solution for HLW. Whether some simple commercial approaches are possible, what regulatory changes should occur, and what the implications may be if we do not change our current path, will be discussed. After two decades, we see the low-level nuclear waste industry maturing and replicating the path of the hazardous waste industry. When new hazardous waste federal regulations were enacted in the 1970's and 1980's, the hazardous waste industry began to boom with new company start-ups and construction of numerous treatment and disposal facilities. Initially, this was a good business to be in, managing dangerous wastes in an environmentally safe and responsible manner, and it was profitable. However, over-capacity and fierce competition drove down pricing, and companies failed. This ultimately led to a major consolidation within the hazardous waste industry. Similarly and more recently in the nuclear industry, over-capacity and competition has driven down profitability, companies have failed, and there is a major consolidation occurring within sectors of the industry. For consolidation to be successful and healthy for either industry, a leader must emerge to set the direction for the others and establish a sustainable path that is acceptable for waste generators, stakeholders, and investors. This was the case for the hazardous waste industry. This paper will examine the parallels between the hazardous and nuclear waste industries as they have matured and evolved, and what discuss the long term implications of consolidation within the nuclear industry. (authors)

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DOE OSTI2018

The DISTINCTIVE University Consortium: An Overview - 18102

The UK government remains committed to nuclear energy having an important role to play in delivering a secure, low carbon and affordable energy future. Successful delivery of their strategy recognises the need for underpinning research, in addition to the provision of a skilled workforce and international collaboration as key enablers. One crucial area is the decommissioning and clean-up of all existing civil nuclear licensed sites and the implementation of geological disposal of highly active wastes, with the Nuclear Decommissioning Authority responsible for the delivery of policy aims with respect to legacy wastes. A central component of public acceptance of this strategy is a demonstrated ability to safely manage and dispose of waste, with acceptance linked to having scientifically robust routes for its safe clean-up and disposal. Similarly, the UK Engineering and Physical Sciences Research Council (EPSRC) strategy is to maintain investment in nuclear fission, with the development of approaches for implementing safe, cost-effective clean-up, decommissioning and waste disposal of existing and future nuclear sites and facilities a strategic focus. The EPSRC-supported DISTINCTIVE (Decommissioning, Immobilisation and Storage Solutions for Nuclear Waste Inventories) consortium is carrying out research that addresses the broad area of nuclear waste and decommissioning, bringing together academic researchers from 11 of the UK's leading research intensive universities and key industry partners. When allied to the strong links of the partners with the industrial and international R and D community, the project provides an excellent route for building a greater capacity for R and D within the UK, underpinning the strategic needs noted. The consortium also addresses the skills agenda in training a significant number of researchers with high level skills of direct relevance to nuclear waste and decommissioning. This paper describes the aims and objectives of the work being undertaken, and the technical work underway within each of the research theme areas: AGR, Magnox and exotic spent fuels; PuO{sub 2} and fuel residues; legacy ponds and silo wastes; and structural integrity. Projects using radioactive materials are also considered. As the project enters its final year, consideration is given to the impacts of the project for both academic and non-academic beneficiaries. (authors)

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DOE OSTI2018

Impacts of Crust Properties on the Waste Feed Delivery of Hanford Tanks with Flammability Concerns - 18161

Mitigation of flammability concerns within Hanford waste tanks is essential to Safe Storage and Waste Feed Delivery sub-functions of the River Protection Project Mission. The tanks that have a higher propensity for the generation, retention, and release of flammable gas are considered Waste Group A tanks. Group A tanks include 241-SY-103, 241-AN-103, 241-AN-104, 241-AN-105, and 241-AW-101. Successful mitigation of flammability concerns in these tanks is necessary, as they are critical to the completion of the River Protection Project Mission as indicated in the Integrated Waste Feed Delivery Plan. The benchmark for examination of tank flammability issues is 241-SY-101. Studies of Tank 241-SY-101 gas release events showed that the mechanism of gas retention and release was highly dependent on the volume and configuration of a floating crust layer. Beginning in 1993, 241-SY-101 underwent water additions and mixer pump operations resulting in dissolution of the crust. These efforts mitigated the flammability concerns of the tank. A similar mitigation approach was recommended for Tank 241-SY103 and is recommended for the other Waste Group A tanks. However, this recommended mitigation will be more or less effective based on the expected thickness of the crust present in each of the Group A tanks. Considerations of flammable gas generation and release mechanisms were essential in the development of a successful mitigation strategy for Tank 241-SY-101, and will be essential in the development of future mitigation strategies. Because crust properties are difficult to measure, tank surface level trends and current temperature profiles are evaluated to understand crust properties. 241-AN-103 and 241-AW-101 are the only two tanks with a net increase in surface level, indicating a higher generation and retention of flammable gas. Additionally, these are the two tanks with the thickest crusts, which causes both a higher retention of gas and a reduced evaporation rate. Given the current temperature profile, crust thickness for 241-AW-101 was re-evaluated to be 145 cm (57 in) ± 30 cm (12 in) which is different from its historic value. This gives 241-AW-101 the thickest crust, second only to 241-SY-101 prior to mitigation. The crust thickness for 241-AN-103, however, retains its historic value of 89 cm (35 in) ± 16 cm (6 in). These crust properties suggest modifications to the mitigation strategy are needed for each tank. For example, 241-AN-103 and 241-AW-101 will require additional water to dissolve the crust layer before flammability concerns are mitigated and further operations could commence. Implementation of the modified mitigation strategies will mitigate flammability issues and help fulfill the Safe Storage and Waste Feed Delivery sub-functions of the River Protection Project Mission safely and efficiently. (authors)

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DOE OSTI2018

Unique Facilities, Innovative Treatment Processes, and a Talented Workforce Allow Fluor Idaho to Successfully Disposition Transuranic Waste - 18388

Fluor Idaho is utilizing state-of-the-art equipment, treatment technologies, and employee expertise at the Department of Energy's (DOE) Idaho Site to treat contact-handled transuranic (CH-TRU) waste, as well as extremely challenging remote-handled transuranic (RH-TRU) waste forms, to prepare the material for shipment and permanent disposal at the Department's Waste Isolation Pilot Plant (WIPP). Nowhere else in the DOE Complex - perhaps the world - will one find the compilation of tools and an experienced workforce with an unparalleled history of safely retrieving, characterizing, treating, repackaging, certifying, and shipping TRU waste. Since 2003, the Advanced Mixed Waste Treatment Project (AMWTP) has used both conventional and unique retrieval concepts, high-tech characterization equipment - including real-time radiography and radioassay - robotics, and automated treatment processes to retrieve and treat 65,000 cubic meters of above-ground CH-TRU waste. Additionally, engineered controls and specialized personal protective equipment have allowed employees to make thousands of entries into high-radiation areas without a skin or clothing contamination incident - evidence of the employees' commitment to safety. AMWTP has encountered and found solutions for many challenging waste forms, including polychlorinated biphenyl-contaminated wastes, TRU waste containers with up to 1,000 fissile gram equivalent plutonium-239, oversized TRU boxes, and partially absorbed organic sludge waste. By utilizing additional facilities and equipment at the Accelerated Retrieval Project V and VII, Fluor Idaho has developed new processing lines and capabilities to ensure worker safety and milestone completion. These two soft-sided buildings, located at a facility adjacent to AMWTP, were originally constructed to exhume Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) TRU wastes from a section of a 97-acre landfill. After a total of 1,150 cubic meters of targeted waste were removed, personnel, working in collaboration with state regulators, successfully implemented a unique regulatory strategy that allowed these CERCLA facilities to be modified and permitted to treat Resource Conservation and Recovery Act (RCRA) wastes from the AMWTP - a savings of tens of millions of dollars over the cost of constructing new buildings. Adding to the AMWTP's technological uniqueness is its ability to treat wastes from other facilities in the DOE Complex. Two prior decision documents and an agreement with the state of Idaho have allowed AMWTP to treat wastes from 15 other DOE sites. It is the only project in the complex to have done so. At another project located next to AMWTP, targeted waste exhumation within the Subsurface Disposal Area is approximately two years ahead of schedule. Crews have removed more than 8,600 cubic meters of CH-TRU and hazardous wastes from eight exhumation areas that equal a combined footprint of 4.75 acres. Less than one acre remains to be exhumed to satisfy the 2008 Record of Decision waste exhumation end state between the DOE, Environmental Protection Agency, and state of Idaho. Construction of the final building where targeted waste exhumation will continue was completed during the summer of 2017. Exhumation will be completed within the next few years. Following exhumation completion, an evapotranspiration barrier (or cap) will be constructed over the entire landfill. Fluor Idaho continues to make progress on a more challenging project involving an inventory of RH-TRU waste that also contains the reactive metal sodium. Crews are utilizing hot cells, once used for spent nuclear fuel reprocessing and high-level waste treatment at the Idaho Nuclear Technology and Engineering Center, to render this challenging waste form safe for disposal in a national geologic waste repository. The treatment methods being employed on this RH-TRU waste are unique in the DOE Complex. Minute concentrations of water are sprayed onto the sodium-contaminated material in a process called 'spritzing' to chemically change the sodium from a reactive metal to a non-reactive salt. In other cases, sodium-contaminated waste is dipped in water to initiate a reaction in a process called 'immersion.' Lastly, for more challenging waste shapes that could not be successfully treated with the two prior methods, crews lower the sodium-contaminated RH-TRU waste into a one-of-a-kind distillation unit that uses high temperature and vacuum to volatilize the sodium and recapture it in a condenser unit for disposal. The RH-TRU waste treated with this technology is then safe for permanent disposal. Fluor Idaho is proud to operate DOE's CH- and RH-TRU waste treatment facilities, as well as use innovative processes and its talented workforce to complete this important work-scope. Such capabilities allow the contractor to meet the commitments made by the DOE to the state of Idaho. (authors)

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DOE OSTI2018

Robotic Crane System for the Deconstruction of Chernobyl Unit 4 - 18583

The Chernobyl Nuclear Power Plant (ChNPP) Unit 4 accident occurred on April 26, 1986 leaving a severely damaged reactor building and a highly radioactive area. To minimize the long-term radiological impact to personnel and the environment, Unit 4 was immediately enclosed by a structure named the Object Shelter (OS). To transform the OS into an ecologically safe system, the Shelter Implementation Plan (SIP) was developed in 1997. The most significant SIP tasks was the design and construction of the New Safe Confinement (NSC). The NSC is a 36,000 tonne arch enclosure that includes a suite of process equipment to address the ruins of Unit 4 while assuring the safety of workers, public, and the environment. The NSC has a span of 257 meters, length of 164 meters, and height of 110 meters and is designed for a 100 years life. The New Safe Confinement was slid from its erection area to directly over Unit 4 (a distance of 327 meters) in November 2016. One of the major mechanical handling systems installed in the NSC is the Main Cranes System (MCS), designed and fabricated by PaR Systems (USA). The planned deconstruction activities require the handling of shielded waste disposal casks containing fuel as well as transporting extremely large structural elements -remotely controlled using video surveillance equipment. The lifting activities required the cranes to be tele-robotically controlled to prevent inadvertent collisions, expedite repetitive and high accuracy functions, protect the equipment and structures within the NSC as well as increase the overall effectiveness of the MCS. In addition, due to the elevated radiation levels, remote deconstruction activities will need to be performed. The MCS has design provisions included for a tele-robotically controlled manipulator and a suite of tooling to be used for multiple activities including concrete coring, cutting, debris collection and other heavy demolition activities. (authors)

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DOE OSTI2019

Waste is our Product. Developing, Implementing and Optimising Waste Management at Dounreay - 19039

The UK's Fast Breeder Reactor Research programme was centred at Dounreay, in the very north of Scotland, from 1955. Since 1994, the site has been focused on decommissioning, and represents one of the most complex, technically challenging and hazardous nuclear clean-up projects in Europe. Since 2012, Dounreay Site Restoration Ltd (DSRL) has been contracted by the NDA as the Site License Company to deliver the defined scope of work, under a Target Cost contract, to decommission the site to achieve the Interim End State by 2030-2033, whilst managing the transition of its people. All parties, including DSRL's Parent Body Organisation Cavendish Dounreay Partnership Ltd (a consortium of Cavendish Nuclear, AECOM, and Jacobs), recognise that there are significant technical and regulatory challenges to achieve this, as well as safely managing the transition for our workforce on what is effectively a 'closure' contract. The product of decommissioning is waste; over the course of the programme, a full range of higher activity and lower level radioactive wastes will be generated, and must be safely managed. This paper will address: - The Policy issues relating to radioactive waste management at Dounreay, including particular issues relating to implementation of Scottish policy, including how Waste Management is Optimised. - An overview of the decommissioning programme and the delivery of an Interim End State; - The type, range and nature of radioactive wastes requiring safe management at Dounreay; - Higher Activity Wastes (HAW): - Setting higher activity waste strategy; defining products and interim storage requirements; - Technical, regulatory and stakeholder issues for HAW strategy; - Lower Activity Wastes: - Setting Low Level Waste (LLW) strategy; defining a disposal product and development and status for Dounreay's LLW Disposal Vaults; - Implementing LLW strategy; - Technical, regulatory and stakeholder issues for LLW strategy; - Optimising Waste Management -Optimising waste from generation (from decommissioning workfaces) through to disposal -why, and how; - The Waste Informed Decommissioning model as a methodology and approach; - Impact and importance of Waste Management on a target cost contract; - Establishment of a Waste Optimisation Programme to drive ownership, delivery and culture change; - Management responsibilities and informing organisational design; - Sharing and learning from Best Practice; - The Challenge - how to ensure workforce and teams are engaged in delivering optimisation requirements; - Conclusions. (authors)

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DOE OSTI2019

Molten Salt Reactor Experiment (MSRE) Lay-Up - 19187

UCOR, an AECOM-led partnership with Jacobs, manages the cleanup of the 2,200-acre East Tennessee Technology Park for its client, the U.S. Department of Energy (DOE), in Oak Ridge, Tennessee. The site was contaminated with radioactive, hazardous and industrial wastes generated by more than 40 years of national defense and energy missions. 2019 is the 50. anniversary of the shutdown of the Molten Salt Reactor Experiment (MSRE). MSRE ran a brief four years in the 1960's but earned an enduring legacy as an innovative reactor technology concept which has enjoyed a renewal of global interest in the past several years. In 2018, it was given elevated status among advanced reactor programs within DoE's Nuclear Energy office. The MSRE was a graphite-moderated, liquid fueled reactor that operated from June 1965 through December 1969. Although it was shut down nearly 50 years ago, certain systems within the reactor have continued to operate to maintain the facility in a safe and stable condition until it can be demolished as part of the comprehensive cleanup of the Oak Ridge Reservation. These systems are composed of a complex network of pipes, valves, gauges and associated equipment that control air filtration and pressure buildup resulting from fluorine gas generated through radiolysis of the fluoride salt which once fueled the reactor. In 2016, an MSRE Engineering Evaluation Team was chartered to assess current facility conditions and develop recommendations to maintain the facility in a safe state until deactivation and demolition commences. Plans are now underway to 'lay up' the MSRE facility until deactivation and demolition can occur. The proposed MSRE Facility Layup will minimize maintenance costs, reduce personnel risk of injury and exposure, provide reliable electric service to the facility and require zero personnel occupancy at MSRE. In the proposed layup approach a new source of electrical power to required loads will be routed through the facility, and new gas processing and ventilation systems will be installed that will be optimized for safe processing of the constituent hazardous gasses that exist in the legacy vessels at MSRE. (authors)

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DOE OSTI2020

Analysis of Exposure during Transportation of Waste to the Nevada National Security Site (NNSS) - 20445

Safety is a priority for the U.S. Department of Energy (DOE) Environmental Management (EM) Nevada Program. This commitment to safety is integrated into requirements and work procedures at DOE sites across the U.S. that generate low-level radioactive waste, mixed low-level radioactive waste, and classified waste that is transported to the Nevada National Security Site (NNSS) for permanent disposal. Generators of this waste are responsible for ensuring it is packaged and transported in a safe and compliant manner. These generators and their contracted commercial carriers must comply with U.S. Department of Transportation (DOT) regulations; the Nevada National Security Site Waste Acceptance Criteria (NNSSWAC); and other applicable federal, state, and local regulations and requirements. This includes packaging and shipping the waste to meet DOT regulatory radiation dose limits to ensure public safety during transportation. NNSS waste generators have a solid safety record for complying with these stringent regulations as demonstrated by the fact that no shipments exceeding DOT dose rate limits were received at the NNSS Area 5 Radioactive Waste Management Complex (RWMC) during FY 2014 through FY 2018. A 2005 study conducted by the Desert Research Institute validated shipments to the NNSS comply with DOT radiation dose limits. In response to recent Nevada stakeholder requests for an updated study, an analysis was conducted to verify and provide evidence that the public continues to be safe from exposure to radioactive waste shipment to the NNSS. Radiation exposure from radioactive waste shipments in transit to the NNSS was evaluated. Three studies were performed: 1. Confirm compliance of waste shipments to the DOT radiation limits using dose rate surveys taken at the generator site and at the disposal site. 2. Estimate the dose to the public from waste shipments using the dosimetry results for disposal site personnel by comparing time of exposure for various individuals/scenarios. 3. Estimate the truck driver dose using the receipt dose rate surveys and time of exposure, and from dosimetry results. Study 1 confirmed compliance with the DOT dose rate limits for a sample of waste shipments (123 of 1,038) received during fiscal year (FY) 2017. Study 2 showed that the average annual radiation equivalent dose for disposal site personnel during FY 2017 was approximately 6.7% [0.415 mSv (41.5 millirem)] of the annual average dose that a person living in the U.S. receives from all sources [6.2 mSv (620 mrem)]. Maximum dose to a member of the public, a person resting/sleeping in an adjacent truck at a rest area to a waste shipment, was 0.0166 mSv (1.66 mrem), which is less than the dose received from a person taking a round trip from the east coast to the west coast of the continental U.S. on an airplane [0.07 mSv (7 mrem) per round trip]. Study 3 showed that the truck driver who made the most shipments of radioactive waste to the NNSS, who drove the most hours (estimated time of exposure 2,520 hours), during FY 2017 had a total dose of 0.15 mSv (15 mrem), which was obtained from their personal dosimeter. The annual dose recorded for disposal site personnel, the dose estimates for the public, and the truck driver dose were all a small percentage of the average radiation dose a person receives from all sources (e.g., background and medical radiation). At these levels, risk should only be described qualitatively. Radiation exposure from radioactive waste shipments to the NNSS were shown to be compliant with the DOT dose rate limits, safe, and very low (i.e., within background levels). (authors)

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DOE OSTI2020

When Every Drum is a Win - Tackling a Thirty-Five Thousand Drum Legacy - 20064

Cameco Corporation's (Cameco) Port Hope Conversion Facility (PHCF) was previously owned by the federal Crown Corporation, Eldorado Nuclear, which held a significant inventory of legacy waste material at the time Cameco was formed in 1988. As a result, Cameco was granted an allocation of 150,000 cubic metres of space in the Long-Term Waste Management Facility (LTWMF) located in the Municipality of Port Hope (MPH), which opened to the receipt of Cameco material in June 2018. Cameco is currently undertaking a major site cleanup and renewal of its Port Hope Conversion Facility (PHCF) which is known as the Vision in Motion (VIM) project. Over its operating history, Eldorado accumulated an inventory of over 35,000 drums of accumulated waste that was primarily stored at two offside warehouse locations. In 2017 and 2018, the project focused on repackaging two well characterized legacy wastes (magnesium fluoride slag and depleted uranium titanium oxide) from one of the two offsite locations. At the conclusion of this work, approximately 15,000 drums of these wastes had been repackaged and disposed of at the LTWMF. There remained approximately 1000 drums from this location that either could not be repackaged safely or were different waste types and required further verification and/or processing. The buildings at this location were scheduled for demolition in early 2019, which triggered the development of a process for triaging legacy drums in August 2018, which allowed for preparation and shipment of approximately half of these drums to the LTWMF by March 2019, with the remainder moved to the second offsite warehouse location in February 2019 for further characterization which is ongoing. This paper will discuss key lessons learned as the drummed legacy waste disposal inventory has been reduced to approximately half of the initial inventory. This will include the development of an alternate packaging process; prioritization of characterization activities and selection of techniques where minimal inventory information is available; determination of next step(s) for each drum as it is assessed; key safety considerations; and how to make inroads into an overwhelming task while under public and regulatory scrutiny. In a relatively short period of time, significant progress has been made to organize and gather information about the legacy waste, update the inventory records and determine the most appropriate pathways (i.e. LTWMF disposal, disposal at another appropriate facility, site storage until future processing and/or disposal). Since 2017 the volume of the decades-old legacy waste inventory has decreased significantly. With the majority of the known materials safely disposed of at the LTWMF, every drum removed from the endless rows of 20,000 pyramidally-stacked waste drums with limited history that is safely characterized, prepared and shipped is considered a win. (authors)

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DOE OSTI2020

Contaminant Migration Modeling to Support the In Situ Decommissioning of Hardened Nuclear Facilities - 20215

The decommissioning of hardened nuclear facilities provides a unique challenge in balancing current worker dose with exposure to future receptors and media. In situ decommissioning (ISD) is a cost-effective and safe option for the closure of such facilities but requires a detailed understanding of the threat that radioactive and hazardous constituents remaining in the facilities pose to groundwater and surrounding surface water. The Savannah River Site (SRS) extensively employs contaminant migration models within the decommissioning process to develop site-specific removal, grouting, and monitoring strategies in support of safe and effective final end states of nuclear facilities. This work discusses the use of contaminant migration modeling in the ISD process at SRS, including specific examples from the closure of P-Reactor and R-Reactor, Building 235-F, F-Canyon and F/H Laboratory Complex. Migration models developed during the closure of P-Reactor and R-Reactor were divided into four source areas: the reactor vessel, the process area, the disassembly basin, and the purification wing. Each source area was assigned a specific inventory and migration pathway and was then subject to varying hypothetical removal, capping, and grouting schemes. Modeling indicated that groundwater Maximum Concentration Limits (MCLs) might be exceeded if no action was taken for eleven and ten constituents of concern (COCs) at P-Reactor and R-Reactor, respectively. These exceedances could occur in as few as 200 to 500 years. Alternatively, the migration modeling demonstrated that the selected ISD actions reduced contaminant mobility which allowed for significant radioactive decay and resulted in fewer predicted exceedances of groundwater MCLs (five COCs for P-Reactor; eight COCs for R-Reactor). In response to the modeling results for P-Reactor and R-Reactor, effectiveness monitoring programs were developed to target contaminants, specific to each reactor, that may migrate to groundwater. Contaminant migration modeling also revealed that roof collapse was a large factor in the release of COCs to the environment, giving rise to roof improvements, and an inspection and vegetation control program to ensure roof stability over time. Contaminant migration modeling is also aiding in the closure planning for hardened facilities in F Area. Building 235-F housed the Actinide Billet Line, which produced Np-237 billets for irradiation in SRS reactors, and the Plutonium Fuel Form (PuFF) facility that produced Pu-238 heat sources for the space program. As a result of these missions, areas within Building 235-F contain considerable residual amounts of both Pu-238 and Np-237. Contaminant migration modeling of Building 235-F was originally performed in 2012 to identify the feasibility of ISD and the amount of radioactive material removal required to prevent the exceedance of groundwater MCLs. The original model indicated that a 60% reduction in the PuFF facility Pu-238 inventory could keep groundwater concentrations below standards, while Np-237 did not pose a threat to groundwater. However, updates to the model with an emphasis on source impact pathways revealed that, due to the orientation of the source areas relative to groundwater flow, no amount of reasonable removal of Pu-238 would keep groundwater concentrations below MCLs and that Np-237 could be a large contributor to localized MCL exceedances. With this insight, the refined 2019 model is being used to assist in the development of grouting plans specific to each facility source area, where bentonite may be utilized to slow the migration of Pu-238 and its daughter products from the PuFF facility and a reducing grout may decrease Np-237 transport by ensuring the nuclide remains in the less mobile +IV oxidation state. The beginning phases of contaminant migration modeling are underway for F-Canyon and associated facilities using lessons learned from the Reactors and Building 235-F. A contaminant migration pathway, similar to the pathway used for the reactor vessels, is being developed for the hot and warm canyons. The F-Canyon inventories are being spatially refined to identify specific and localized source zones, comparable to Building 235-F, that may impact groundwater. Contaminant migration modeling provides vital input within the ISD process, from facility investigation to post closure effectiveness monitoring, making it a valuable tool in the development of safe and effective end states for hardened nuclear facilities at SRS. (authors)

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DOE OSTI2020

Achieving the End State for the Pahute Mesa Corrective Action Units at the Nevada National Security Site - 20355

Underground nuclear testing at the Nevada National Security Site (NNSS) ended in 1992. To address the impact of radionuclide contamination from underground nuclear testing on the groundwater resources of Nevada, the U.S. Department of Energy (DOE) Environmental Management (EM) Nevada Program created the Underground Test Area (UGTA) activity to characterize the radionuclides in groundwater, understand the nature and extent of radionuclide migration, and forecast the distribution of radionuclides in groundwater for 1,000 years. The UGTA activity is regulated by the Federal Facility Agreement and Consent Order (FFACO), an agreement negotiated between the Nevada Division of Environmental Protection (NDEP), DOE, and the U.S. Department of Defense (DoD). DOE is close to achieving the end state (closure in place with monitoring and institutional control) for three of the five UGTA corrective action units (CAUs) on the NNSS. A number of challenges remain to achieve the end state of the other two CAUs, both located on Pahute Mesa. Pahute Mesa was the location of 82 underground nuclear tests, less than 10% of the total number of tests on the NNSS. Yet, Pahute Mesa contains slightly more than 60% of the total radionuclides (in curies). Based on groundwater sampling in monitoring wells, two radionuclide plumes have migrated several kilometers (km) in groundwater from selected test cavities and have crossed the boundary of the NNSS (yet remain within the boundaries of Federally controlled land). The path to achieve the end state continues to evolve as more data are collected and a better understanding of radionuclide migration in groundwater is developed. DOE has invested in drilling and sampling more than 50 characterization and monitoring groundwater wells over the past 25 years. The data indicate that the primary radionuclide of concern is tritium as it is about 89% of the total radionuclide inventory (in curies). As well, only tritium has been measured in groundwater outside of cavities at concentrations exceeding the Safe Drinking Water Act (SDWA) standard. For one of the tritium plumes that has migrated across the NNSS boundary, the average rate of migration has been measured as about 45 meters (m) per year with the rate of migration at the leading edge of the plume of about 85 meters per year. At that rate of migration, the tritium plume will decay to safe levels and not reach the publicly accessible environment at concentrations above the SDWA standard. Other radionuclides, at concentrations below the SDWA standards, will be monitored to ensure they remain at safe levels. Consequently, the end state path forward for Pahute Mesa has evolved to take full advantage of the data from the monitoring network to constrain uncertainty in model forecasts and to reduce reliance on probabilistic simulations. The focus of the end state approach relies on developing a monitoring well network that is protective of human health by increasing confidence that no radionuclide plume in the groundwater will migrate undetected to the accessible environment, located about 22 km from the nearest up-gradient underground nuclear test. Using the measured data to remove uncertainty, the evaluation of radionuclide migration from Pahute Mesa is directed toward meeting the goals of the end state. (authors)

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

Sludge Removal: Success and Partnership with T Plant - 20510

U.S. DOE contractor CH2M Hill Plateau Remediation Company (CHPRC) has safely and efficiently managed the removal of highly radioactive sludge from the 105-K West Reactor fuel storage basin (105- KW Basin) near the Columbia River. In mid-September 2019, sludge retrieval was completed after 21 storage containers were safely loaded with highly radioactive sludge and transported approximately 13 miles from the 105-KW Basin to T Plant, located in the 200 West Area on the Central Plateau of the Hanford Site. Sludge removal represents the last major source term reduction necessary before the K West Reactor and fuel storage basin can transition to closure activities in preparation for final deactivation and demolition. The technical complexity of the sludge retrieval process, coupled with the challenging physical and radiological characteristics of the waste, necessitated a methodical and deliberate approach using unique design and operational solutions to safely conduct the work. This challenge was further complicated by the need to plan and conduct the work in two separate facilities, each subject to the controls and requirements of its own nuclear facility safety basis. The project overcame these challenges through creating an integrated team that actively engaged, communicated and coordinated each phase of the project to ensure successful completion. This paper will cover how these key elements of the Sludge Removal Project led to its success: - Integrating project planning and management; - Lessons learned from design and construction; - Preparing for a successful campaign (lessons learned from testing and start-up); - Lessons learned from operating in two separate facilities, transporting sludge containers on the Hanford Site and ensuring compliance with a range of regulatory and safety basis requirements. The establishment of an integrated group spanning two facilities and multiple organizations created a team capable of overcoming the challenges necessary to successfully plan and execute the sludge retrieval mission. Sharing lessons learned from this successful project can enhance the ability of teams across the DOE complex to successfully plan and execute complex projects. (authors)

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

Research and Experiment on the Key Technologies of the Spent Fuel Storage in HTR-PM - 18486

High Temperature Reactor demonstration project of China (HTR-PM) is one of the major national science and technology projects, some key technologies for the design and operation of the demonstration project are formed to research according to the main technology. The spent fuel storage system is one of the main systems of the HTR-PM. In order to solve the technical problems related to the operation of spent fuel, several experiments have been conducted, including: spent fuel loading, getting back, spent fuel canister hoisting, storage well cover hoisting, storage canister drop and so on. In the study process, the test facility has been set up to carry on the operation and the experiment verification to ensure the reliability and safety of the spent fuel storage system. Spent fuel storage system undertakes spent fuel receiving, spent fuel canister transfer, storage, lifting and other functions in HTR-PM, the project research need to solve several major technical problems related to the spent fuel operation, which mainly include: Develop the transportable spent fuel canister for HTR-PM, the canister is corrosion resistance, impact resistance, and is in subcritical condition when loaded with fuel. Develop the spent fuel canister hoister, which could hoist the spent fuel canister safely and reliably. Develop the spent fuel storage well cover hoister, which could hoist the well cover safely and reliably. Study the small probability of the spent fuel canister in extreme condition of canister hoist falling accidents, to ensure the radiation protect safety and residual heat removal safety of the canister in case of the canister drop when lifting. Develop the spent fuel charging device and return device, which could load the spent fuel elements into the spent fuel canister, and could take out the fuel elements from the canister. The key technology and equipment of the spent fuel storage system in HTR-PM would be solved through the research of the project. Project research is very important for the implementation and safe operation of the HTR-PM project. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Recovery from loss of reductant feed to a U/Pu separation contactor, magnox reprocessing plant, Sellafield - 14217

Following a revealed sustained loss of reductant feed to the Primary Separation 4 (PS4) U/Pu separation contactor in the Magnox Reprocessing Plant in 2014 the solvent extraction cycles were safely shutdown. However, initial mass balance calculations indicated that over four times the typical mass of Pu could be present in PS4. Due to the loss of reductant feed, much of this Pu was expected to be in the solvent phase of Tributyl Phosphate and Odourless Kerosene (TBP/OK) where it would be difficult to observe with the neutron monitors located under PS4. Samples were taken of the solvent in PS4 which confirmed the presence of Pu in solvent; normally almost all of the Pu in PS4 is in the aqueous phase. The volume ratio of solvent to aqueous in the PS4 contactor is high. This meant that simply re-dosing PS4 with reductant could not be supported as a recovery option due to the potential to realise high Pu concentrations in the aqueous phase. This paper details the recovery options which were considered and the ultimate decision to recover PS4 using Gadolinium (Gd) poisoning and Gd-dosed reductant feeds. The recovery included: - The need to consider the impact of Gd poisoning on monitoring of the recovery with the PS4 neutron monitors; - Close support from criticality safety personnel on the plant at all times during the recovery; - Production of operator aids and short fit for purpose safety documentation to assist with the recovery. The recovery of PS4 proceeded safely and smoothly. There were no losses of Pu to raffinate streams despite the prolonged shutdown. The plant was then safely rundown to flush out residual Pu and Gd. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Special Considerations for the Removal and Disposal of Micro-Reactor Experiments

Idaho National Laboratory (INL) is preparing to host several microreactor experiments through 2030 and beyond. Two new test beds currently under development will operate as microreactor or nuclear system experiment user facilities. Test bed experiments will be performed in series, with each nuclear experiment installed, operated, and removed before installation of the subsequent experiment. The necessarily brief transition period between experiments introduces unique equipment removal and radioactive waste disposal challenges. This paper evaluates equipment removal and radioactive waste management topics associated with tight sequencing of nuclear experiments and presents some of the solutions and approaches currently planned to meet these special considerations for one such experiment, the Molten Chloride Reactor Experiment (MCRE), slated for operation at INL’s Laboratory for Operation and Testing in the United States (LOTUS). The MCRE project is a collaboration between Southern Company Services, TerraPower, and INL, among others, to provide integral nuclear data that will advance molten salt fast reactor technology. MCRE will be the first critical fast-spectrum circulating fuel system ever operated and the first experiment operated in the LOTUS test bed. The experiment will nominally operate at zero power with planned low power excursions as part of operational planning for MCRE has been to minimize at-power operations to limit fission product formation while still achieving experimental objectives. After the experiment is complete, MCRE will be allowed to radioactively decay for a short period (nominally 90 days) prior to system defueling, flushing, removal, and disposal of all MCRE equipment, readying the test bed for the next nuclear experiment. Equipment removal and radioactive waste disposal have been integral to MCRE project planning since the Cooperative Research and Development Agreement was formalized in 2021. Due to requirements for future use of the test bed, the MCRE system, of necessity, must be removed in a much shorter time frame than typical for historic reactor decommissioning projects at INL. This results in minimal time for radioactive decay, resulting in not only elevated radiation dose rates but also the presence of short-to-medium-lived isotopes not typically encountered in the reactor decommissioning and radioactive waste management space. Additional unique constraints placed on the project include lack of intrinsic remote-operations capabilities in the test bed, space constraints in the test bed once MCRE has been installed, and contamination minimization requirements to return the test bed to as-found conditions to enable future use. This paper discusses planned solutions to these challenges. Approaches for implementing remote or semi-remote technologies in a non-hot cell environment with limited space availability are discussed. The paper also summarizes the systems engineering approach for concept development and design of equipment removal systems, which are being implemented concurrent with the MCRE design phase, providing feedback to system designers to incorporate features enabling efficient and safe equipment removal approaches. The timing of this paper at a relatively early phase of the project is meant to highlight the importance of early planning for nuclear system decommissioning while reactor design is ongoing to allow design feedback on componentry driven from decommissioning system needs. As new, innovative nuclear reactor technologies enter the nuclear market sector, reactor experiments are crucial for providing new integral nuclear data that establish safe operational margins for technology advancement. Microreactor experiments at INL will require safe, effective, and timely decommissioning approaches, which in turn require nuclear systems designed to expedite decommissioning. In addition to the advancement of nuclear reactor technologies, these nuclear experiments provide an opportunity to demonstrate, deploy, and test new removal and disposal capabilities to support the next generation of nuclear reactor technology.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2007

Operations at the JPL OCTL Telescope

The JPL Optical Communications Telescope Laboratory (OCTL) is a 200 sq-m located at 2.2.km altitude in Wrightwood California and houses a state-of-the-art 1-m telescope. The OCTL team is involved in the development of operational strategies for ground-to-space laser beam propagation for future NASA optical communications missions. Strategies include safe beam propagation through navigable air space, line of sight optical attenuation monitoring, adaptive optics, and multi-beam scintillation mitigation. This paper presents the results of recent operations at the OCTL facility including telescope characterization data and laser beam propagation experiments to Earth-orbiting retro-reflecting satellites; experiments that validate the telescope's tracking and blind-pointing performance and safe laser beam transmission procedures for propagating through navigable airspace.

optical communications↗