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At least 973 records · Page 54

Evaluation of the MARVEL Reactor Inhalation Dose Consequences

A dose consequence analysis was performed to support the preliminary Microreactor Applications Research Validation and Evaluation Project (MARVEL) design effort. The purpose of the MARVEL project is to develop a nuclear microreactor applications test bed at Idaho National Laboratory (INL) to perform research and development on various operational features of microreactors to ultimately improve integration of microreactors to end-user applications. The MARVEL concept is a 100 kilowatt (kW) thermal (kWth) and approximately 20-kW electric (kWe) generating microreactor. The MARVEL system will be located in the INL TREAT facility in the north high-bay equipment pit. The analysis considers the bounding release fractions of the MARVEL reactor concept and documents the dose consequences to the public and collocated worker.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Long-term Prediction of Spent Fuel Storage Canisters - Issues and Approaches - 18112

The considerations involved in aging management of stainless steel canisters used in dry cask storage of spent fuel are discussed in this paper in a probabilistic framework. The two important modes of degradation in the long-term performance of the casks are localized corrosion and stress corrosion cracking initiating from the external surface. These degradation modes are determined by the evolution of the electrochemical behavior and the strain rate of the stainless steel. Laboratory data suggest that Type 304L/316L stainless steels should suffer from localized corrosion and SCC under salt deposits. However, field inspection of the canisters has not indicated extensive pitting or stress corrosion cracking. This discrepancy can be understood in terms of the joint probabilities of various factors that must come together to cause these failure modes. A Bayesian network approach is suggested as the framework for modeling the probability of failure of canisters and long-term management of aging. The gaps in knowledge to conduct a Bayesian network analysis are discussed. (authors)

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Waste Encapsulation and Storage Facility Stabilization and Ventilation Project-18121

Department of Energy contractor CH2M HILL Plateau Remediation Company (CH2M) is responsible for the safe and compliant operation of the Waste Encapsulation and Storage Facility (WESF), a nuclear materials storage facility at Hanford containing 1,936 highly radioactive capsules containing cesium and strontium. The capsules are currently stored under approximately 4 m (13 ft) of water. The water shields workers from radiation and helps cool the capsules. WESF is a large, concrete structure nominally 48 m (160 ft) long, 30 m (98 ft) wide and 12 m (40 ft) high with the pool cells being approximately 4 m (13 ft) below grade level. It is a robust facility with walls ranging from 0.3 m to 1 m (1 ft to 3 ft) thick. While the capsules are currently in a safe configuration, WESF is an aging facility, and DOE and CH2M are modifying the facility to ensure continued safe and compliant operations and to support the eventual transfer of capsules to dry storage. Dry storage would eliminate the possibility of a release of radioactive material in the unlikely event of a major earthquake that might result in loss of pool storage water and overheating and breach of one or more capsules. The project was divided into two major deliverables: 1) Ventilation: replacement of the existing ventilation system which included a below-grade high efficiency particulate air (HEPA) filter pit with an above grade HEPA filter skid and new fans and ducting. 2) Stabilization: grouting deactivated facility hot cells and other void spaces to the maximum extent practical to reduce material at risk from legacy contamination. The aging ventilation system relied on HEPA filters that exceeded their operational life. Due to the high levels of radionuclide contamination, replacing the filters was impractical and highly hazardous. CH2M installed a new ventilation system to provide safe and compliant ventilation for future operations. The project used an existing, unused ventilation skid that was designed for another project. Repurposing the unused skid presented additional technical challenges to modify the system for the project specific requirements. The new system is above ground, easier to maintain and is built to current standards. New connections inside the facility were installed to allow the new ventilation to operate effectively. The stabilization project is unique in that core drilling and grouting was performed in an operating Hazard Category 2 nuclear facility, which brought forth safety basis and environmental regulatory challenges. The hot cells were isolated from all existing support utilities including electrical, process chemicals, manipulator ports, instrument air, etc. to seal the hot cells and prevent highly contaminated supernate or grout from leaking into adjacent galleries. A high flow grout was used to maximize the distance between injection ports. (authors)

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Subsidence Risk Mitigation in LLW Final Covers: A Case History - 18286

Typical waste packaging and disposal practices at low-level radioactive waste (LLW) disposal facilities, while adequate for safe transportation and facility operations, introduce significant subsidence risk for the facilities' final cover systems. Subsidence is often caused by the collapse of voids in the waste packaging and surrounding cover soils that eventually propagate to the ground surface. This subsidence could result in the loss of final cover system integrity, stormwater ponding, concentrated flow of surface water and/or gas, increased leachate generation, transport of contaminants from the landfill, and potential adverse chemical reactions within the waste mass. For new LLW facilities, steps can be taken to minimize the subsidence potential of newly placed waste. However, there is a significant inventory of existing disposal facilities (trenches, pits, shafts, vaults, etc.), tank farms, and historic, subsurface burial grounds that require final covers - all of which present real subsidence risks. Final covers for these existing LLW disposal sites can be engineered to mitigate these risks. A case history of a 60-acre LLW final cover system design and construction project illustrates an innovative approach to this problem that offers superior risk mitigation compared to typical final cover design practice. This approach combines engineering simulations of the projected final cover mechanical performance, specific probabilistic performance targets, and targeted cover system improvements to manage the subsidence risk in the final design. The result is a successful final cover project that meets the needs of the community, state and federal regulators, and facility owner. (authors)

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Real-time Erosion-Corrosion Detection in Waste Transfer Pipelines using Guided Wave Ultrasonic Sensors - 18368

A Fitness-for-Service (FFS) program for the Waste Transfer System has been implemented by the United States Department of Energy Hanford Site Tank Farm to examine structural parameters of the waste transfer systems in order to develop erosion/corrosion rates for relevant system components. The FFS information is acquired from opportunistic evaluations of pipelines that have been removed from service. The objective is to deliver more realistic estimates to the remaining useful life of the components and incorporate those estimates into future design/testing plans across the DOE complex. To aid to this objective, previous efforts at Hanford included the installation of sensors on some of the POR 104 components, to provide real time pipe wall thickness measurements. Due to certain limitations, alternative approaches for remote permanently mounted pipe wall ultrasonic thickness measurement systems have been investigated by FIU. Present research work is a supportive effort by FIU-ARC for DOE and WRPS, which investigates the option for remote-permanently mount guided wave ultrasonic transducer (UT) systems for HLW pipe wall thickness measurements. Specific applications include straight sections, elbows and other fittings used in jumper pits, evaporators, and valve boxes. The proposed method will provide information that will assist engineers with understanding the failure potential of HLW transfer components due to corrosion and erosion. This information can assist in determining if and when lines need to be removed, saving time and resources on the unneeded excavation of transfer lines. This information will also assist engineers with designing new transfer systems by establishing more detailed/accurate guidelines governing the life expectancy of the transfer system and its components. In the previous work, FIU has investigated the use of various ultrasonic systems that are either commercially available or used previously at Hanford and selected the most promising systems for further evaluation. The down-selected system is the guided wave UT sensor system from Permasense which is a UK based manufacturer. It is an integrated wireless sensor network system which used a patented technology for the acoustic wave propagation and has proven applications in the oil and gas industries. Scope of the present research includes the initial verification and validation of the Permasense Guided Wave sensor system as a potential erosion/corrosion detection system for carbon steel pipelines under static conditions. In future, the real-time erosion/corrosion detection will be investigated by measuring the thinning of the pipe sections by circulating simulants and continuous monitoring. This is achieved by passing abrasive and caustic solutions through an in-house designed pipe loop system. Results obtained provide realistic wear rates on 2 inch and 3 inch carbon steel pipes with straight sections and elbows using the Permasense UT sensor system. Thus, the present research delivers solutions for sensor evaluations, conducts bench scale testing followed by data acquisition and analysis for corrosion and erosion assessment. (authors)

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Effects of Chemical Wastes on Concrete through Spectroscopic and Computational Models - 18405

Liquid mercury was used as a coolant for early fast reactors, as catalyst in the dissolution of uranium alloy fuels, lithium isotope separation and other processes. Cesium is produced during nuclear fission reactions within the reactor. The interaction between chemical contaminants and solid structures can cause structural damages (e.g. dislocations) and consequently modify the material's mechanical properties. Another problem associated with the interaction of chemical contaminants and metal structures is corrosion pitting. Mercury intrusion porosimetry (MIP) has been utilized to investigate the pore structure of the cement material. The overall goal of this project is to determine the physiochemical interactions of Hg and Cs to steel and concrete using spectroscopic analysis: - Evaluate the physicochemical interactions of Hg and Cs to steel and concrete using spectroscopic techniques under standard conditions; - Evaluate the impact of environmental conditions (temp., salt content, pH, and time) on the physicochemical interactions of Hg and Cs to steel and concrete using spectroscopic techniques. Conclusion and Future Work: Adsorption process depends on the surface morphology. Surface morphology analysis demonstrated that concrete samples had a heterogenous surface. Adsorption of Mercury and Cesium are more favorable onto heterogeneous regions such as edges, grooves, and wrinkles. The mechanism for cation removal by concrete includes surface complexation and ion exchange. Overall adsorption of Mercury was more favorable than Cesium onto concrete. Continuing investigations will be focusing on application considerations such as multicomponent interactions and desorption studies.

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Aging Management Considerations for Dry Cask Storage Canisters - 18472

Dry cask storage systems (DCSSs) are used to store used nuclear fuel (UNF) at independent spent fuel storage installations (ISFSIs) prior to final geologic storage or reprocessing. As storage durations increase beyond the twenty-year period commonly implemented for licensing the storage sites, it is important to consider aging related degradation issues for DCSS components. A large portion of the DCSS designs used throughout the world employ a ventilated concrete overpack with a welded stainless steel canister that confines the UNF. This paper explores key issues related to aging management for these canister components, including: (1) canister environment, (2) inspection limitations, and (3) aging management program guidance and requirements. While the austenitic stainless steel used for the canisters are generally resistant to most forms of environmental attack, it is possible for conditions to arise that lead to pitting and even stress corrosion cracking of the canister systems. A large number of studies related to canister degradation have been performed by industry groups and government researchers, both in the US and worldwide. These advances in the collective knowledge provide insight into risk factors for DCSS. Such information is beneficial to improve ISFSI siting decisions and ensure successful relicensing of existing ISFSIs. This paper highlights relevant research and also discusses applicability to degradation of other equipment and containers used to process and isolate UNF and other nuclear waste streams from the public and the environment. Inspection of the canister surface is an important tool to investigate and evaluate the potential for degradation. Existing DCSS designs typically do not offer ready access to the canister surface in order to perform the inspection. Further complicating inspection efforts are the surface conditions present on the sheltered canisters after extended storage times, including atmospheric debris such as dust and pollen. Given ongoing development efforts, inspection capabilities and associated limitations should be assessed at multiple points in time over the course of an aging management program to incorporate new advances in the field. This paper explores the limitations of currently available capabilities in inspection and delivery, and provides insight into the gaps remaining to be filled to permit an effective inspection program. Recommendations and requirements related to programs for aging management of canisters have been published or are in progress from a variety of groups, including industry groups, regulatory authorities, and Codes and Standards. An effective aging management program should consider the technical bases of these various inputs and maintain compliance with requirements. Given multiple sources, decisions will be required when considering conflicting guidance, and this paper will include discussion of the relative merits of the various approaches. (author)

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Development of Remote Dry Cutting Technology for Reactor Vessel Segmentation - 19176

Decommissioning of nuclear power plants is becoming a large market due to the many reactors reaching their operational lifetime limit in the near future. One of the first activities in a decommissioning project is to segment the reactor internals and the reactor vessel. These activities are also one of the more complex ones since they deal with most of the radioactivity inventory that is left in the plant after the spent fuel is gone. They are almost always on the critical path, so an improvement in execution time will be favorable for the entire decommissioning program. Westinghouse has been successful for more than 30 years in more than 40 segmentation projects globally. The traditional technology that has been used in these projects has up until now been underwater mechanical cutting of both the reactor internals and the reactor vessel. Underwater mechanical cutting is proven and very reliable, but it is slow and since there are major advantages to the decommissioning program with improvements in schedule performance, Westinghouse is developing alternative technologies both as part of current offerings but also as entire new concepts. For reactor vessel cutting, Westinghouse has developed a new concept involving thermal cutting in a dry environment. The benefit of dry cutting is the speed of the cutting and the subsequent schedule and cost improvements. A dry environment also means that the pools can be drained which means that a critical activity can be moved forward in the schedule and that there will need to be less monitoring and controlling with no pumps, cooling and recirculation system that need to be operated. Going from underwater mechanical to dry thermal cutting reduces the on-site schedule by more than 50%. The dry cutting concept development considers both PWR and BWR reactors. Some of the challenges of dry thermal cutting involve the creation of radioactive off-gases and how to ventilate these, the extreme heat generation, remote handling of the thermal equipment, cutting station arrangements and how to remove the insulation around the reactor vessel. Depending on the concept the reactor vessel might need to be lifted from the pit in one piece, which also poses some challenges due to the weight, volume and radioactivity of the vessel. (authors)

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Pilot Test Remediation of a Fractured Bedrock Aquifer at the Maywood Superfund Site, Maywood, New Jersey - 19217

The U.S. Army Corps of Engineers (USACE) and Cabrera Services are conducting a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Pilot Test treatment of a fractured bedrock aquifer at the Formerly Utilized Sites Remedial Action Program (FUSRAP) Maywood Superfund Site (FMSS) located in Maywood, New Jersey (NJ), USA. The Maywood Chemical Company Superfund Site in Bergen County, NJ is listed on the United States Environmental Protection Agency (USEPA) Superfund National Priorities List (NPL). The National Superfund Comprehensive Environmental Response, Compensation, and Liability Information System (CERCLIS) identification number is NJD980529762. The Maywood Chemical Company Superfund Site (hereafter referred to as the FMSS) consists of 92 designated properties in the communities of Maywood, Lodi and Rochelle Park, NJ. These are collectively known as FMSS 'Vicinity Properties' and include property owned by the federal government (the Maywood Interim Storage Site, or MISS); the Stepan Company (former location of the Maywood Chemical Works, or MCW); 62 residential properties; three properties owned by the state or federal government; four municipal properties; and 21 commercial properties. The Stepan Company property includes contaminated buildings and three remediated U.S. Nuclear Regulatory Commission (NRC) licensed burial pits; the MISS also includes a contaminated building. Contamination on the FMSS is being addressed under three separate actions coordinated by the lead regulator USEPA Region 2. The USACE is addressing thorium (Th) and other wastes at the site defined as 'FUSRAP waste' in the Soils and Buildings and Groundwater Records of Decision (ROD) for the FMSS. Stepan Company is addressing other chemical wastes (non-FUSRAP waste) at the FMSS. The Groundwater ROD [1] was signed into agreement by USACE Division Commander in June 2012 and USEPA Region 2 Administrator in July 2012 and presents the selected remedial actions for Operable Unit (OU) 2 groundwater at the MISS and adjacent properties. The selected remedial action is removal of non-radiologically contaminated soils on the MISS containing arsenic, lithium, and benzene with concentrations above cleanup levels, and monitored natural attenuation (MNA) of arsenic, lithium, and benzene in groundwater. In situ treatment of arsenic in the overburden aquifer using oxidation reduction (redox) alteration will also be performed if needed to meet the groundwater cleanup levels after the source soil is removed. The need for in situ treatment will be determined during the Groundwater ROD review at least three to four years after the contaminated source soil removal is completed. The third action addressing non-FUSRAP chemical wastes will be the responsibility of the Stepan Company as a primary responsible party in this action. (authors)

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Development of a Preliminary Performance Assessment to Support Single-Shell Tank Waste Retrieval - 19446

A preliminary Performance Assessment (PA) of Single-Shell Tank (SST) Waste Management Area A-AX (WMA A-AX) located at DoE's Hanford Site in southeastern Washington is being conducted to satisfy the requirements of the Hanford Federal Facility Agreement and Consent Order (HFFACO), as well as other Federal and State requirements of closure plans and permits. The WMA A-AX PA effort is assessing the fate, transport, and impacts of radionuclides and hazardous chemicals within residual wastes left in tanks, ancillary equipment, and facilities in their assumed closed configuration, and the subsequent risks to humans into the far future. WMA A-AX includes the 241-A Tank Farm (A Farm), the 241-AX Tank Farm (AX Farm), and the ancillary equipment associated with each Farm. There are six underground SSTs in A Farm, and four underground SSTs in AX Farm. Each of these SSTs have a storage capacity of 1,000,000 gallons. A complex waste transfer system of ancillary equipment supported the transfer and storage of waste within WMA A-AX SSTs. This ancillary equipment includes transfer lines, catch tanks, diversion boxes, vaults, valve pits, and other miscellaneous structures. This paper emphasizes the development of the PA effort focused on radiological impacts to meet the requirements for a closure authorization under DOE Order 435.1. Lessons learned from retrieval and development of a recently completed PA of the nearby Waste Management Area C (WMA C) led to the decision to develop a preliminary PA for WMA A-AX prior to initiation of retrieval efforts so that a site-specific risk management capability could be available to support anticipated retrieval decisions. The technical approach for the PA analysis includes developing both a process-level model using site-specific hydrogeological data, as well as a system-level model. The system model is being developed to facilitate uncertainty and sensitivity analyses being conducted in the PA effort. Preliminary results of the DOE Order 435.1 PA are evaluating potential impacts to the groundwater and air pathways to determine radiation exposure once radionuclides are released to the environment from residual waste after closure of the tank farm. To meet other analysis requirements in DOE Order 435.1, the PA effort must also evaluate potential radiological impacts from inadvertent intruder scenarios into the closed WMA A-AX facility. The results of the PA evaluation will be compared to performance objectives and measures in the DOE Order 435.1 to support retrieval and eventual closure decisions at WMA A-AX. (authors)

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Fluor Idaho Technical Investigation-ARP V Lid Ejection Reactants and Mechanism - 19521

In April 2018, four 208.2-L (55-gal) drums containing transuranic waste at the Idaho National Laboratory Site became over-pressurized, resulting in lid ejection and expulsion approximately 50% of the contents. The parent drums were part of a retrieval effort and were exhumed from a portion of Subsurface Disposal Area Pits 11 and 12 beginning in 1974 and ending in 1978. Drums were later retrieved from the cargo containers, characterized using real-time radiography and nondestructive assay, and opened. The contents were visually inspected and repackaged into new 'daughter' drums within the Accelerated Retrieval Project V facility. Lid ejection occurred 7-12 hours after repackaging. The lid ejections were accompanied by significant heat generation. Empirical evidence of heat generation was observed in each reacted drum. The heat generated was sufficient to warp the high-density polyethylene liner (< 122 to 137 deg. C) in three of the drums and char the drum paint (∼400 deg. C) on one drum. To determine the potential reactant(s), a comprehensive chemical analysis was performed by two independent laboratories, Southwest Research Institute and Savannah River National Laboratory, on the reacted drum contents, ejected drum material, and related drums. A comprehensive chemical compatibility analysis was performed, and the most likely reactant(s) were identified. The oxidation of depleted uranium was the source of the heat generation. The presence of depleted uranium in the daughter drums and its chemical behavior (reactive properties and lack of visual indicators during packaging) are consistent with the event. The heat generated was sufficient to accelerate secondary chemical reactions leading to significant gas formation sufficient to generate pressures well in excess of the pressures needed to eject the lid and contents. These reactions fall into two categories: (1) simple volatilization of organics and water present in the drum and (2) hydrolysis of beryllium carbide, which generated significant quantities of methane gas. (authors)

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Development of Robotic Crawlers for Inspection of Radioactive Waste Tanks and Transfer Lines - 19573

The leaks in the primary liner of Double Shell Tank (DST) 241-AY-102 have motivated a number of research efforts to develop inspection tools that can obtain information regarding the integrity of the Hanford DSTs. Recently, there have been additional concerns regarding the DST secondary liners as well various pipelines within the waste transfer system. Over the past few years, Florida International University (FIU) has developed a pneumatic pipe crawler that can navigate through 3- and 4-inch diameter pipes and provide visual feedback while traveling through pipelines in excess of 100 feet and multiple 90 degree elbows. Recent updates to the crawler include a new sensor module that can provide information regarding environmental conditions within the pipeline and a mapping of the interior surface of the line. The sensor module is a carousel type device that can rotate 360 degrees and is placed near the front of the 3- and 4- inch crawler. The module can house a number of sensors including temperature, pressure, radiation, Lidar, optical camera and an inertial measurement unit (IMU). The intent of the module is to improve the functionality of the crawler by providing environmental conditions within the pipe and internal surface topography maps that can assist in understanding potential wear, pitting or thinning of the pipe. In addition to the 3- and 4-inch crawler, FIU is developing a 6-inch crawler that utilizes the same principles as the smaller unit. The objective of this tool is to navigate through the 6-inch drain lines leading to the drain channels within the foundation underneath the secondary liner of the DSTs. This crawler will house a miniature rover that will be deployed at the end of the drain pipe, where it connects to the foundation. The miniature rover will operate upside down, via magnets, underneath the secondary liner within the drain slots and provide visual information regarding the condition of the secondary liner. Thinning of the secondary liner has been noted from the annulus of the some of the DSTs, but it is unclear if the thinning is isolated to locations near the annulus. (authors)

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Pilot Test Remediation of a Fractured Bedrock Aquifer at the Maywood Superfund Site, Maywood, New Jersey - 20073

The U.S. Army Corps of Engineers (USACE) and Cabrera Services are conducting a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Pilot Test treatment of a fractured bedrock aquifer at the Formerly Utilized Sites Remedial Action Program (FUSRAP) Maywood Superfund Site (FMSS) located in Maywood, New Jersey (NJ), USA. The Maywood Chemical Company Superfund Site in Bergen County, NJ is listed on the United States Environmental Protection Agency (USEPA) Superfund National Priorities List (NPL). The National Superfund Comprehensive Environmental Response, Compensation, and Liability Information System (CERCLIS) identification number is NJD980529762. The Maywood Chemical Company Superfund Site (hereafter referred to as the FMSS) consists of 92 designated properties in the communities of Maywood, Lodi and Rochelle Park, NJ. These are collectively known as FMSS 'Vicinity Properties' and include property owned by the federal government (the Maywood Interim Storage Site, or MISS); the Stepan Company (former location of the Maywood Chemical Works, or MCW); 62 residential properties; three properties owned by the state or federal government; four municipal properties; and 21 commercial properties. The Stepan Company property includes contaminated buildings and three remediated U.S. Nuclear Regulatory Commission (NRC) licensed burial pits; the MISS also includes a contaminated building. Contamination on the FMSS is being addressed under three separate actions coordinated by the lead regulator USEPA Region 2. The USACE is addressing thorium (Th) and other wastes at the site defined as 'FUSRAP waste' in the Soils and Buildings and Groundwater Records of Decision (ROD) for the FMSS. Stepan Company is addressing other chemical wastes (non-FUSRAP waste) at the FMSS. The Groundwater ROD [1] was signed into agreement by USACE Division Commander in June 2012 and USEPA Region 2 Administrator in July 2012 and presents the selected remedial actions for Operable Unit (OU) 2 groundwater at the MISS and adjacent properties. The selected remedial action is removal of non-radiologically contaminated soils on the MISS containing arsenic, lithium, and benzene with concentrations above cleanup levels, and monitored natural attenuation (MNA) of arsenic, lithium, and benzene in groundwater. In situ treatment of arsenic in the overburden aquifer using oxidation reduction (redox) alteration will also be performed if needed to meet the groundwater cleanup levels after the source soil is removed. The need for in situ treatment will be determined during the Groundwater ROD review at least three to four years after the contaminated source soil removal is completed. The third action addressing non-FUSRAP chemical wastes will be the responsibility of the Stepan Company as a primary responsible party in this action. (authors)

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3-D Radiological Data Acquisition, Visualization and Modeling - 20211

The U.S. Army Corps of Engineers (USACE) was tasked to investigate and remediate low activity radiological contamination from research and production of the nation's first nuclear weapons at the former DuPont Chambers Works Formerly Utilized Sites Remedial Action Program (FUSRAP) site (DuPont). The DuPont site had several buildings used for the Manhattan project that were demolished in the 1940's and 1950's apparently using heavy earthmoving equipment. Some of the contaminated rubble from the demolition appears to have been spread out by this equipment resulting in somewhat random scattering of radiologically contaminated soil and debris along with aqueous spills. Traditional investigative methods such as soil borings, test pits and 2-dimensional gamma walkovers were only partially successful in delineating the radiological contamination at the site. It was feared that even 'chasing' the contamination during remediation would miss contamination if the demolition resulted in discontinuous trails of radiologically contaminated soils. In evaluating the data generated over the interceding decades, the USACE determined that a better method to collect and process the remedial action radiological data was needed to enable the project team to optimize predictive planning and meet documentation expectations. The purpose of this paper is to provide an overview of the effort and progress to combine and organize radiological survey methods into a highly flexible sampling, modeling, and decision analysis approach that emphasizes the quality of decision-making during remediation. This innovative system blends multiple tools to develop a methodology that can extend MARRSIM [1] into the subsurface and provide tools that can be applied to other sites. (authors)

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Investigation of Non-Nuclear Plowshare and Vela Uniform Sites - 20447

The Plowshare Program was initiated by the Atomic Energy Commission, a predecessor agency to the US Department of Energy (DOE), to investigate peaceful uses of nuclear devices. 'Vela Uniform,' part of the broader Project Vela, was designed to use nuclear and other explosions to increase the capability of the United States to detect nuclear tests through seismicity. There were underground nuclear tests conducted off the Nevada National Security Site (NNSS) as part of the Plowshare Program (e.g., the Gnome-Coach, New Mexico, Site) and the Vela Uniform Program (e.g., Project Shoal, Nevada, Site) at which the DOE Office of Legacy Management (LM) has been conducting long-term surveillance and maintenance since fiscal year (FY) 2006. There are many additional locations, particularly for the Plowshare Program, that were considered for nuclear tests or used for non-nuclear tests off the NNSS. Some were only named on paper, but at 30 sites, field work such as drilling test wells and conducting tests using conventional explosives was done. A Plowshare nuclear explosion had been proposed at Project Chariot but was canceled in 1962. In 1992, the discovery of soils contaminated with radionuclides from tracer experiments led to a cleanup by the DOE Office of Environmental Management (EM) of the Project Chariot, Alaska, Site, which was completed in 1994. Following the cleanup of the Project Chariot site, EM was concerned that there might be environmental liabilities at other Plowshare or Vela Uniform sites where field investigations were done, even if the nuclear test was never conducted. The EM Program at the Nevada Site Office began examining records and visiting selected locations that were evaluated for Plowshare and Vela Uniform non-nuclear tests. In total, 161 Plowshare and 14 Vela Uniform sites or proposed projects were identified. The Desert Research Institute (DRI) created summaries of the work carried out at these sites and developed a collection of unclassified documents for them. Preliminary investigations had been conducted at these sites by both LM and EM since the early 2000's. In 2016, in preparation for accepting responsibility for the sites, LM started a program to determine if there were any environmental or safety liabilities associated with them. In 2019, LM formally accepted responsibility for these sites off the NNSS where non-nuclear explosive tests were conducted or where underground nuclear tests were proposed as part of the Plowshare and Vela Uniform programs but were never conducted. LM responsibilities will include maintaining institutional knowledge of the sites, responding to public inquiries, and addressing any concerns at each site. To minimize risk associated with site management, LM has archived all historical records compiled by DRI and has performed limited investigations at sites identified by the prior studies as having a higher potential liability. Preserving records related to site operations reduces the risk of doing unnecessary site investigations in the future caused by the loss of institutional knowledge. During the investigations, some maintenance needs were identified and will be addressed by the end of FY 2021. At the end of FY 2021, investigations, maintenance, and archiving activities will be complete. Remnants of test activities found at sites include craters, mud pits, debris, and open boreholes. Open boreholes are being plugged and abandoned in accordance with state regulatory standards. Other issues are being addressed in accordance with land owner requests and preferences as well as state regulations. Archiving activities include the development of project summaries identifying key operational details and current condition to allow site managers to be easily briefed. The LM experience at non-nuclear Plowshare and Vela Uniform sites applies to other programs where there is the potential for liability associated with the non-nuclear aspects of nuclear programs. This may include support activities or early phases of planned nuclear projects. Preserving institutional knowledge enables land managers to adequately address any concerns associated with operations and lowers risk. (authors)

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Progress on Advancing the Robotic Air-slot Volumetric Inspection System (RAVIS) for Hanford Under-tank Inspection - 20519

The 27 in-service double-shell tank (DST) systems at Hanford provide interim storage for nearly 56 million gallons of nuclear and hazardous chemical legacy waste. The safe and reliable storage of the waste is critical to the site's extended cleanup mission. While in service, the leak integrity of the DSTs must be demonstrated to satisfy the Washington Administrative Code, inform decisions on tank management, and gauge fitness for continued service beyond originally intended design lives. The carbon steel plates and welds that comprise DST primary tank sidewalls and lower knuckles are routinely 'sampled' for volumetric non-destructive testing using ultrasonic inspection technology to determine whether corrosion pitting, cracking, or general corrosion are occurring that could compromise tank leak integrity. However, the carbon steel plates and welds that comprise the bottoms of primary tanks have not been thoroughly inspected since the DSTs were constructed (i.e., since 1968-1986) due to limited physical access to the exterior tank bottoms. The Robotic Air-slot Volumetric Inspection System, or 'RAVIS,' is being engineered to overcome the challenge of tank bottom in-service inspection. The system will be deployed under primary tanks within the small cooling channels/air-slots in the refractory pads upon which the tanks rest and will use ultrasonic guided wave inspection technology to volumetrically inspect physically inaccessible regions that make up ∼90% of a tank bottom, as well as regions that are physically accessible from the air-slots but comprise only ∼10% of a tank bottom. This paper describes the air-slot sensor and robotic deployment system and focuses on the testing approach and methods that are being applied to support (1) design feedback and performance acceptance (requirement verification) and (2) performance repeatability/reliability assessments (qualification) to ensure the RAVIS is prepared for successful inservice inspections of primary tank bottoms. Test progress made against the test plan, the results of 2019 testing, and the future outlook are included. (authors)

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Panel Session 46: Final Closure of Unlined LLW Disposal Facilities Worldwide (R3.5)

This panel discussed the on-going and emerging issues with the design and construction of final cover system of unlined disposal facilities (e.g., material disposal areas (MDAs), subsurface disposal areas (SDAs), burial grounds, and similar historical waste trenches, pits and vaults). The discussion focused on key technical and regulatory issues as well as lessons learned from years of engineering and construction in US and non-US locations. Panelists with presentations: Closure of Unlined LLW Disposal Sites (Craig Benson); Closure of LLW Facilities (Paul Pointon); Final Closure of Unlined LLW Disposal Facilities - Lessons Learned (Kevin Foye)

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Augmented Monitoring and Condition Assessment Program for SNF Wet Storage Life Extension - 20489

Approximately 27 MTHM of spent nuclear fuel (SNF) owned and managed by the U.S. Department of Energy, Office of Environmental Management is stored in the L Basin at the Savannah River Site (SRS). This 'DOE SNF' is comprised of approximately 12,000 aluminum-clad, aluminum-based fuel assemblies (∼7 MTHM), and approximately 2000 non-aluminum fuel assemblies (∼20 MTHM). A program is in progress to perform Non-Destructive Examination (NDE) of the fuel and their storage containers to characterize the materials' condition, and to evaluate the effects of service to enable continued safe wet storage of the SNF. The Augmented Monitoring and Condition Assessment Program (AMCAP), a two-part program to develop and implement remote underwater Non-Destructive Examination of the aluminum SNF, and of the containers of the non-aluminum SNF, respectively, is aimed at the characterization and evaluation of corrosion degradation of aluminum fuel and container materials. The predominant design of the aluminum SNF (ASNF) stored in L Basin is the Materials Test Reactor (MTR) equivalent design, a plate fuel design. These SNF are stored in 5'' diameter tubes or 5'' x 5'' squares (called bundles) that are nominally 12' or up to 14' long. The SNF is de-bundled and inspected using a custom-designed MTR Fuel Inspection Table. The inspection table provides for indexed fuel positioning for a video camera examination with controlled lighting. A total of 10 of the SNF originating from foreign research and test reactors were selected for examination based on burnup, enrichment, and prior damage caused by service/storage history. A special inspection campaign of these 10 assemblies is in progress. The observed corrosion damage included minor to moderate attack from general corrosion, pitting, crevice, end grain and galvanic corrosion. Example results from the completed inspections are shown and discussed. The focus is a comparison of the as-received condition versus the as-found current storage condition that will serve to validate Water Chemistry and Corrosion Monitoring Programs. The non-aluminum SNF (NASNF) stored in L Basin are of diverse design that includes various geometries with claddings of stainless steel, Zircaloy, and Hastelloy. The fuel core materials include uranium alloys, oxides/mixed oxides, and carbides. These SNF materials, originating from early experimental and test reactors, are in various physical forms including single fuel elements and cut pieces. The fuel is stored in L Basin in various configurations including in direct bundled storage in aluminum tubes and in isolation cans that are in the bundles or in a larger over-size storage container (OSC). Concern with inside-out corrosion and the potential loss of configuration control and ability to handle the storage containers prompted the development of remote NDE methods that include visual and UT technologies to assess the condition of the containers. Candidate stored materials were selected for a special inspection campaign; the development of the NDE methods for the inspection for galvanic, crevice and sediment-induced corrosion (inside-out) are discussed. A summary of the SNF storage in L Basin at the SRS, and an overview of the AMCAP to enable continued safe storage of DOE SNF are described. (authors)

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