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DOE OSTI2020

Spent Nuclear Fuel Exploratory Roadmaps - 20054

The U.S. Department of Energy manages nearly 2,500 metric tons of heavy metal (MTHM) of spent nuclear fuel (SNF) resulting from several decades of research, testing, and production reactors [DOE 1995]. This SNF is managed at the DOE Hanford Site in Washington State (Hanford), Idaho National Laboratory (INL), and Savannah River Site in South Carolina (SRS). From 1995 to 2004, the Department of Energy (DOE) made several key programmatic decisions, supported by appropriate documentation in accordance with the National Environmental Policy Act (NEPA) for the management of SNF [DOE 1995]. These decisions have provided an overarching framework for SNF management, as well as site-specific and SNF-related management decisions for the past two decades. In the years since these decisions were made, with the notable exception of the successful drying and packaging of the production reactor fuel at Hanford, a majority of the decisions have been largely unimplemented. Also, since these decisions were made, a number of changes that bear on considerations relative to the path forward have occurred. A SNF Exploratory Road map activity, identified reasonable alternate pathways for DoE's inventory of SNF. Three fundamental pathways were identified in this activity for the long-term management and disposition of DOE SNF: Direct Disposal Pathway, Existing Processing Pathway and Alternate Processing Pathway. Further actions would be required (e.g., evaluation of NEPA analysis, technology maturity evaluations, refinement of cost and risks, benefits and advantages relative to other alternate paths) to define and inform many key decisions that will result in the selection of the disposition pathways that are most beneficial to the US Government in dispositioning DoE's SNF inventory. DOE is currently managing all SNF safely; however, the age of DOE-owned SNF and facilities for storing and processing SNF, coupled with the uncertainty of the storage duration, necessitates decisions and actions to ensure that the infrastructure will be in place to ensure continued safe and effective long-term management and eventual disposal. The identification of reasonable alternate pathways took these considerations into account in an effort to proactively manage conditions that could challenge the safety of storing and managing DOE SNF over the time periods now contemplated and also to ensure that flexibility is preserved to ensure that the DOE SNF remains compatible with final disposition pathways when they become available. (authors)

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DOE OSTI2020

Development of the Mobile Systems for Conditioning of Disused Sealed Radioactive Sources in Serbia - 20105

Sealed radioactive sources (SRS) are being used worldwide in the field of medicine, agriculture, industry and research. They can be found in mobile as well as stationary devices. SRS contains radioactive material that is (a) permanently sealed in a capsule or (b) closely bounded and in a solid form. The capsule or material of an SRS should be strong enough to maintain leak tightness under the conditions of use and purpose for which the source was designed, also in case of accidents. In this case only emitted radiation is utilized. Firstly, the hazard from external radiation has to be considered, but the possibility of contamination due to fracture of the capsule should not be disregarded. The radioactive sources are composed of the radiating isotope contained in the filling medium, the single or double isotope holder that partially or totally surrounds the filling medium, the outer cover that contains the parts mentioned above and the capsule closed airtightly by welding or using some other method. The capsule must be tested for leakage periodically. If the SRS is no longer needed (e.g. replaced by a different technique) or it becomes useless for the intended application (e.g. the activity becomes too weak, the equipment containing the source works poorly or becomes obsolete, the source is damaged or leaking) it is considered disused. Disused sealed radioactive sources (DSRS) are typically conditioned and disposed if a facility is available. If the disposal option is not available, conditioned DSRS should be stored under proper conditions. In some cases, the radionuclide(s) in DSRS can be recovered/recycled or the DSRS can be repurposed for other applications. Conditioning of DSRS ensures containment of the radioactive material, provides confinement for leaking sources, provides sufficient radiation shielding, reduces storage/disposal volume by allowing consolidation of multiple sources into a single storage/disposal container, facilitates transport operations and contributes to safety and security as well. Typically, conditioning technologies are deployed either as permanently installed stationary systems in centralized or mobile on-site waste processing facilities, or in a mobile configuration. Centralized stationary facilities provide a single processing location for multiple users that requires transport of the waste to the facility. On the other hand, mobile systems may be provided for the selection and application of the optimum technology for a specific waste stream (such as DSRS) by bringing the process to the point where the waste is generated. In addition, mobile systems could offer additional flexibility by sharing equipment among multiple waste generating sites for processing campaigns that vary in duration, from very short periods to several years. The term 'mobile processing system' refers to any radioactive waste processing system or component which is designed to be transportable and which is not considered permanently installed. Two mobile system for conditioning of disused sealed radioactive sources are developed in the Public Company Nuclear Facilities of Serbia. Development of these mobile systems was supported by SRB9005 national project via Technical Cooperation of the IAEA. The first mobile system, built inside the 20 feet ISO container, will be used for conditioning of DSRS category 3 to 5. The second mobile system, built inside the 7 m long vehicle (Iveco Daily Van), will be used for dismantling of ionizing smoke detectors mostly with Am-241 sources. Designs of the mobile systems were defined in cooperation with two companies from Belgium (Belgoprocess and Leniko) and a Croatian company Ekoteh as well as with the support of the IAEA experts. The generic safety assessment and operational procedures for the mobile systems are developed. Based on safety assessment the acceptance criteria and operational limits and conditions are established. Operational procedures include: (a) equipment and material requirements, (b) assembling procedure of the mobile unit, (c) procedure for acceptance of devices for dismantling and conditioning, (d) dismantling procedure for devices to recover the DSRS, (e) characterization of DSRS, (f) encapsulation procedure of DSRS, (g) disassembling procedure of the mobile unit, and (h) keeping records, identification and traceability. In addition, radiation safety, health safety, security and emergency preparedness plans are prepared. The generic safety assessment and operational procedures could be updated with site specific requirements, DSRS inventory, and different needs for future customers. Developed mobile systems could be used in all situations when it is feasible to perform conditioning of DSRS on the spot in the county and worldwide. Development of these mobile units was just the first step to create the Reference Center for Radioactive Waste Treatment and Disused Radioactive Sources Conditioning for Small Facilities which can become a regional training center in the future, and/or as a tool for comprehensive national search and secure programmes. In the next phases development of e-learning platforms and blended learning packages as well as application for the IAEA Qualified Technical Centre (QTC) for the management of DSRS is foreseen. (authors)

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DOE OSTI2020

Update to the Performance Assessment for the Savannah River Site Saltstone Disposal Facility - 20124

In 2019, Savannah River Remediation developed a revision to the performance assessment (PA) on behalf of the U.S. Department of Energy (DOE) Savannah River Operations Office (SR) for the near-surface disposal of low-level waste at the Savannah River Site (SRS) Saltstone Disposal Facility (SDF). Soluble waste from SRS Tank Farms undergoes salt processing to remove cesium and other high activity constituents. The low-activity decontaminated salt solution (DSS) is then immobilized by mixing it into a cementitious waste form known as saltstone. After mixing, the saltstone is poured into leak-tight concrete vaults, known as saltstone disposal units (SDUs), where the waste form cures. By the time of facility closure, the SDF is expected to consist of 15 SDUs with a combined capacity of 1.06 E+09 L (280 Mgal) of cured saltstone. The facility operates under a Disposal Authorization Statement (DAS) from DOE and a permit from the South Carolina Department of Health and Environmental Control (SCDHEC). Since the start of operations in 1990, the SDF has received almost 6.7 E+07 L (18 Mgal) of DSS, resulting in the safe disposal of 2.7 E+16 Bq (7.3 E+05 Ci) of activity. Due to the radioactive decay of short-lived contaminants, the total remaining activity in the disposed waste is estimated to be approximately 1.4 E+16 Bq (3.9 E+05 Ci), as of September 2018. The Disposal Authorization Statement requires a demonstration that the system of engineered and natural features of the disposal facility will limit releases from the facility and be protective of human health and the environment for at least the next 1,000 years. The long-term performance of the facility was evaluated under the requirements of the DoE's Radioactive Waste Management Manual (US DOE Manual 435.1-1). Simulations were performed to demonstrate that the disposal facility would meet performance objectives specified in the manual. The evaluation was based on numerical models that simulate the releases of contaminants from the saltstone waste form. Contaminants were transported through groundwater and air pathways to points of assessment to evaluate compliance. In addition, the potential consequences of an inadvertent human intrusion (IHI) were also evaluated. In accordance with the guidance and recommendations in US DoE's technical standard for DAS, deterministic and probabilistic analyses were performed to demonstrate the SDF system, which includes the engineered cover system, the SDUs, the waste form, and the natural features of the site, provides a reasonable expectation that saltstone disposal will meet performance objectives. Post-closure doses to future members of the public (MOP) were evaluated; dose estimates from the air pathway were well below the 1.0 E-04 Sv/yr (10 mrem/yr) performance objective within 1,000 years, and the calculated doses to the MOP from all exposure pathways, including the groundwater and air pathways, were well below the 2.5 E-04 Sv/yr (25 mrem/yr) performance objective within 1,000 years. Doses following an assumed intrusion event within the facility boundaries were well below the IHI performance objectives, where the acute IHI dose was below the acute IHI dose performance objective of 5.0 E-03 Sv (500 mrem) and the chronic IHI dose was below the chronic IHI dose performance objective of 1.0 E-03 Sv/yr (100 mrem/yr) within 1,000 years. (author)

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DOE OSTI2020

Issue Resolution During the Development of the Performance Assessment for the Savannah River Site Saltstone Disposal Facility - 20125

In 2019, Savannah River Remediation developed a revision to the performance assessment (PA) on behalf of the U.S. Department of Energy (DOE) Savannah River Operations Office (SR) for the near-surface disposal of low-level waste at the Savannah River Site (SRS) Saltstone Disposal Facility (SDF). Soluble waste from SRS Tank Farms undergoes salt processing to remove cesium and other high-activity constituents. The low-activity decontaminated salt solution (DSS) is then immobilized by mixing it into a cementitious waste form known as saltstone. After mixing, the saltstone is poured into leak-tight concrete vaults, known as saltstone disposal units (SDUs), where the waste form cures. By the time of facility closure, the SDF is expected to consist of 15 SDUs with a combined capacity of 1.06 E+09 L (280 Mgal) of cured saltstone. The facility operates under a Disposal Authorization Statement from DOE and a permit from the South Carolina Department of Health and Environmental Control (SCDHEC). Since the start of operations in 1990, the SDF has received almost 6.7 E+07 L (18 Mgal) of DSS, resulting in the safe disposal of 2.7 E+16 Bq (7.3 E+05 Ci) of activity. Due to the radioactive decay of short-lived contaminants, the total remaining activity in the disposed waste is estimated to be approximately 1.4 E+16 Bq (3.9 E+05 Ci), as of September 2018. The Disposal Authorization Statement requires a demonstration that the system of engineered and natural features of the disposal facility will limit releases from the facility and be protective of human health and the environment for at least the next 1,000 years. The long-term performance of the facility was evaluated under the requirements of the DoE's Radioactive Waste Management Manual (US DOE Manual 435.1-1). Simulations were performed to demonstrate that the disposal facility would meet performance objectives specified in the manual. The evaluation was based on numerical models that simulate the releases of contaminants from the saltstone waste form. Contaminants were transported through groundwater and air pathways to points of assessment to evaluate compliance (i.e., 100 m from the SDUs). In addition, the potential consequences of an inadvertent human intrusion (IHI) were also evaluated. A number of issues were overcome during the development of these simulations. These issues were identified as part of internal technical reviews. Simulations are developed by people and people make mistakes, so the internal technical review process is a vital step in PA development. Specific examples of resolved issues include a unit-conversion error, an inappropriate definition for a model boundary condition, a model time-stepping issue, and an error in the calculation for the buildup of contaminants in soil. Actions taken to address these issues resulted in an improved product with a better supported technical basis and more defensible results. The identification and correction of these issues are discussed. By understanding these issues, model developers and technical reviewers working on PAs in the future may avoid repeating these types of mistakes. Transparency with respect to these mistakes builds trust between waste management sites, regulators, and stakeholders. (authors)

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DOE OSTI2020

Implementation of the Glycolate Destruction Process with Sodium Permanganate in the Defense Waste Processing Facility (DWPF) - 20141

The Savannah River Site (SRS) seeks to replace formic acid with glycolic acid as a chemical reductant in the Defense Waste Processing Facility (DWPF), where borosilicate glass is mixed with high-level radioactive waste, melted at high temperatures, then poured into stainless steel canisters for safe storage. Prior research and development have demonstrated the feasibility and advantages of the new nitric-glycolic acid flowsheet over the current nitric-formic acid flowsheet to chemically adjust the radioactive sludge slurry waste in the Sludge Receipt and Adjustment Tank (SRAT) prior to vitrification. Use of glycolic acid reduces hydrogen and ammonia generation during the reduction process, thus reducing flammability hazards and purge requirements in the DWPF processing vessels. In addition, glycolic acid will support the new Salt Waste Processing Facility (SWPF) by allowing for receipt of higher volumes of waste due to reduction of flammable gases, increased boil-up rates in the vessels, and easing the processing of future sludge batches containing high levels of mercury. However, during DWPF operations, trace amounts of glycolate are anticipated to be entrained in the recycle stream and sent to the Concentration, Storage, and Transfer Facilities (CSTF) via the Recycle Collection Tank (RCT). The RCT receives condensate from the Slurry Mix Evaporator Condensate Tank (SMECT) and Off-gas Condensate Tank (OGCT) that may contain unreacted glycolate. A literature review found the potential for hydrogen generation due to thermolysis of glycolate in caustic CSTF conditions. Savannah River National Laboratory (SRNL) conducted several tests on simulant and radioactive waste from various tanks at SRS and confirmed that glycolate added to waste at concentrations of approximately 1000 mg/L and higher generates hydrogen through thermolysis. There are potential operating scenarios in the future where glycolate concentration could exceed 1000 mg/L in the 2H Evaporator Drop Tank, the highest accumulation point for DWPF organics received via recycle. These findings presented a need to develop another process to mitigate the impact of glycolic acid to the CSTF. Several DWPF compatible options were evaluated for their ability to destroy glycolate in the DWPF recycle stream under various processing conditions, thereby mitigating its introduction to the CSTF. Downselection testing identified sodium permanganate as the most promising and simplest option to oxidize glycolate. It has been demonstrated to be effective in various RCT simulant compositions, operating temperatures, oxidant addition rates, solution pH, and initial glycolate concentrations. A feasibility study identified existing tanks within the facility that could be used for receiving and storing the commercially available sodium permanganate, as well as feeding it to the RCT for processing. The use of existing equipment as well as the ability to purchase the pre-mixed chemical reduces implementation and operational complexity. The downstream impacts of the sodium permanganate reaction were also evaluated and demonstrated to have minimal impact on other chemical species present in the RCT and CSTF, as well as on the materials of construction of the RCT and CSTF vessels and associated piping and instrumentation. Future work is in progress to validate the glycolate destruction process with testing in actual radioactive RCT waste before implementation in DWPF. (authors)

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DOE OSTI2020

Use of Unmanned Aerial Systems for Post-closure Waste Site Maintenance and Surveillance - 20143

The P- and R-Area Reactor buildings, located at SRS near Aiken, SC, were in-situ decommissioned by grouting below-grade portions of the buildings and demolishing some above-grade structures to grade level. Other reactor building structures were left above-grade and sealed to prevent human or animal access. Because the above grade-structures are expected to continue in their present state for hundreds of years, the condition of the building roofs is critical to mitigate rainwater intrusion. For this reason, building roof areas were strengthened with high strength concrete to ensure the long-term integrity of the roof structure. Periodic inspections of the roof structures are required to ensure that the roofs are functioning properly and to identify damage areas. The traditional method of inspecting closed reactor buildings requires the use of a helicopter and photographer to generate photographs and videos for review. This method proved sufficient but lacked the resolution and clarity that is required for a thorough inspection of the building structure. Fortunately, the Savannah River National Laboratory (SRNL) has established a Small Unmanned Aircraft System (sUAS) program using commercially available and custom-built remote-controlled aircraft. EC and ACP contacted the SRNL program to evaluate whether the sUAS technology could be employed for periodic reactor building inspections. The sUAS can fly within a few meters of the reactor buildings and hover, allowing for a more thorough aerial inspection. The initial sUAS inspection of the P-Area Reactor building was completed in February 2018 and at the R-Area Reactor building in August 2018 and were successful in providing higher resolution photos and videos. The sUAS inspections also revealed that vegetation had begun to grow on the roofs that could potentially damage the structure integrity. SRNL partnered with Virginia Polytechnic Institute and State University to build a custom heavy lift sUAS that could be equipped with herbicides to remotely treat the vegetative growth. Herbicides were successfully applied to the R-Area Reactor building roof using the custom sUAS in September 2018. The use of unmanned aircraft systems at SRS to perform aerial inspections and herbicide treatment in otherwise inaccessible areas has proven to be an efficient and cost-effective technology that provides high value, increases knowledge of facility conditions, and provides for early detection of damage. Periodic inspections of the reactor building roofs using sUAS technology are performed safely, efficiently, and at a significant cost and schedule savings while reducing emissions, noise, and fossil fuel use. The use of sUAS equipment for building inspections and herbicide application at SRS is unique within the DOE complex. The goal of SRS is to apply this remote technology to other waste unit operations and maintenance activities in the future. (authors)

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DOE OSTI2020

Reducing Cost of Chlorinated Volatile Organic Compound Remediation by Transitioning from Active to Passive Soil Vapor Extraction - 20157

Areas of high chlorinated volatile organic compound (cVOC) contamination at the Savannah River Site (SRS) have been undergoing remediation via soil vapor extraction, sometimes coupled with thermal treatments to enhance extraction rates. These active systems are effective in removing large amounts of contaminant mass from the subsurface and mitigating the impacts to groundwater. However, as extraction rates decline, costs must be evaluated with respect to the benefit of continued active operation. A decision framework for identifying conditions when a transition to a more passive remediation is appropriate has been developed with state and federal regulatory agencies. Two remediation areas have recently been transitioned from active soil vapor extraction (ASVE) to passive soil vapor extraction (PSVE) at the SRS. Performance evaluation goals including plume stabilization, overall mass removal trends, environmental sustainability and costs were considered in transitioning from active remediation to passive technologies at both sites. At the Dynamic Underground Stripping (DUS) project at the M-Area Settling Basin, ASVE was combined with steam injection to extract cVOCs during active operations. Steam injection occurred from September 2005 to September 2009. DUS utilized 63 steam injection wells, 34 active vapor extraction wells, and 3 active soil vapor extraction units (SVEUs). Two active SVEUs had 60 horsepower blowers; the third had a 25-horsepower blower. Mass removal was closely tracked during DUS operations; over 181,437 kilograms (400,000 pounds) of cVOCs were removed while active steam injection occurred. After steaming was stopped, ASVE continued. The 34 active wells were evaluated in 2012. The ASVE wells were grouped into categories of high, medium, and low extraction rates. High producing wells remained connected to a single active SVEU. Low producing wells were abandoned, and the medium producing wells were transitioned to PSVE (Microblowers{sup TM}). Microblowers{sup TM} utilize a dedicated blower per well and are solar powered. This passive technology provides energy, maintenance, and operation costs savings while still providing an efficient reduction in cVOC migration to groundwater. In 2018, the remaining ASVE wells were evaluated again. The purpose of this testing was to identify which wells removed the most mass. An optimal well configuration was determined. The criteria to discontinue ASVE was removal of less than 18 kilograms (40 pounds) per week of cVOCs. After 3 months of shutdown (rebound conditions), 2.7 kilograms (5.9 pounds) of cVOCs per week were being removed. Data from the rebound test justified ending ASVE and transitioning wells with higher extraction rates to PSVE. Wells that had depleted the cVOC mass within their zone of influence were abandoned. Performance data from existing PSVE wells justified ending PSVE at wells with depleted extraction rates. Currently the system has 16 PSVE wells operating. Another ASVE system was being used to treat cVOC contaminated soil at the A-Area Miscellaneous Rubble Pile (AMRP) at SRS. System operation began in 2004, with 7 ASVE wells connected to a 60- horsepower blower. Mass removal rates and contaminant concentrations remained consistently low over the ASVE lifespan at AMRP. This indicated that mass removal was diffusion limited. With this data, the 7 ASVE wells were transitioned to PSVE in 2017. Twelve pressure monitoring points were also transitioned to PSVE wells. AMRP currently has 19 PSVE wells operating. Both transitions from active to passive remediation had concurrence from the United States Environmental Protection Agency and the South Carolina Department of Health and Environmental Control. These transitions ensure that only the necessary amount of energy is being exerted to remediate the environment. (authors)

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DOE OSTI2020

Novel Foaming Solutions for High Level Waste (HLW) Processing - 20195

Foaming occurs during treatment of high level waste (HLW) at Savannah River Site's Defense Waste Processing Facility (DWPF) in the Chemical Process Cell (CPC) due to high gas generation from process steam and chemical offgas products [1]. The presence of amphiphilic particles in the waste slurry stabilizes the foam [2], [3]. Efficient processing of HLW requires foam control, as foam-overs lead to lower productivity and potential radioactive contamination of condensate streams. DWPF currently employs Siovation Antifoam 747, a surfactant produced by Momentive Performance Materials, as an antifoaming agent during waste treatment. Despite its ability to control foam generation, processing issues have arisen from its use [4]. During DWPF chemical processing, a chemical antifoam agent must be effective at temperatures up to 103 deg. C between pH 3-13. Antifoam 747 is effective at pH 7 but degrades as pH deviates [5]. Antifoam 747 is 90% by weight Momentive{sup TM} L77 and 10% by weight Momentive{sup TM} Y-17580. These ingredients are trisiloxane products, which are especially vulnerable to attack in acidic and caustic environments. Since Antifoam 747 readily degrades, periodic addition is required. Several processing concerns related to Antifoam 747 have affected DWPF operation. Its overuse warranted concern of potential flammable melter offgas under upset conditions due to the generation of carbon dioxide and hydrogen. As a result, a Potential Inadequacy in the Safety Analysis (PISA) was declared [6]. Antifoam contribution has been included in the melter offgas flammability strategy and the quantity of antifoam is monitored to ensure limits are not exceeded. Testing at Savannah River National Laboratory (SRNL) determined that propanal, tri-methylsilanol, and hexamethyldisiloxane are generated as Antifoam 747 degrades, leading to a second PISA [7]. The formation of these byproducts poses additional flammability risks. Furthermore, the methyl and ethyl groups of these compounds may be responsible for the formation of the methyl and ethyl mercury discovered in DWPF [8], [9]. The formation of organomercury has specifically been called out by DOE as a need for increased research. The effectiveness of non-chemical methods for foam control and the efficiency of alternative antifoam agents were evaluated as part of an effort to improve HLW treatment operations at DWPF and to reduce, or eliminate, the flammability hazards associated with Antifoam 747 currently in use. Foaming of high-level waste (HLW) slurries is an issue at the Defense Waste Processing Facility (DWPF) Chemical Process Cell (CPC) which is currently mitigated with a chemical antifoam agent. The effectiveness of non-chemical methods for foam control and the efficiency of alternative antifoams were examined to improve HLW treatment at DWPF and to eliminate the flammability hazards associated with Siovation Antifoam 747 currently in use. Non-chemical foam control methods were deemed unviable for chemical processing at DWPF, as the spray/mist technique requires too much water and space limitations make installation and implementation of headspace agitators unfeasible. The use of ultrasonic energy thickened and further stabilized the foam. Two suitable alternative defoaming agents, Momentive{sup TM} Y- 17112 and Evonik Surfynol{sup R} MD-20, were identified. Both defoamers control foam over a pH range of 4- 13 at boiling, with no flammable degradation products, and only minimal quantities (100-200 mg/kg) are required under glycolic acid conditions. In order to validate the effectiveness of these defoaming agents, SRNL recommends irradiation testing and laboratory scale SRAT/SME experiments with actual radioactive waste. (authors)

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DOE OSTI2020

Advancement of Commercial Intumescent Expanding Foams for Deactivation and Decommissioning in the Nuclear Sector - 20198

Florida International University (FIU), in collaboration with The Department of Energy's Office of Environmental Management (DOE-EM), Savannah River National Laboratory (SRNL), and sites across the Savannah River complex, have identified an operational requirement for a fixative technology that is intended to immobilize and/or isolate residual contamination within a 3-dimensional space. Fixation of radiological contamination can reduce worker risk and mitigate potentially hazardous conditions, however nearly every marketed contamination fixative has been found to be flammable; a significant concern in radiological facilities. Coupled with this, industry fixatives are normally used as a thin coating which can present problems when attempting to stabilize irregular geometry or areas that are difficult to access whilst ensuring full coverage. The technical evaluation and advancement of commercial-off-the-shelf (COTS) polyurethane foams has yielded a down-selected candidate that shows potential in meeting the requirements to support deactivation and decommissioning activities. Several performance criteria have been established and tested to progress the technology readiness level towards an active field demonstration (TRL-7). Such criteria include: mechanical failure limits, adhesive and cohesive properties, thermal/fire resilience, determining thermal behavior, ability to immobilize contamination, and a means of non-destructive evaluation of applications. The test scenario examined was targeted towards an application for decommissioning nuclear pipework, in which the down-selected polyurethane foam would act as a barrier to segregate pipework and mitigate the potential for release during cutting, packaging, and storage operations. Testing carried out at SRNL included: mechanical evaluation of tensile, compressive, and adhesion strength by dynamic mechanical analysis (DMA), as well as thermogravimetric analysis (TGA). FIU examined the foam's fixative properties by utilizing phosphorescent europium-dysprosium doped strontium aluminate powder to investigating the extent to which contamination can be immobilized. FIU has also exploited previous successes in the field of intumescent technologies to assess the down-selected foam's tolerance to an extreme fire scenario, while maintaining the ability to effectively mitigate a contamination release. Parallel to this, extensive thermal investigations were carried out to determine the upper boundary of anticipated heat generation during the curing process as heat generation has the potential to compromise rubber parts of contaminated enclosures. These investigations subsequently yielded a promising method for a non-destructive application evaluation by means of infrared thermography. Utilizing the high sensitivity of modern IR cameras, coupled with the heat generated during the curing process of the polyurethane foam, FIU has been exploring the concept of monitoring the external pipe surface for indications of an irregular or abnormal application, thus informing operational decision making. The testing carried out utilized several current 'best fit' ASTM standards, which serve as helpful guidelines for testing, however, a precise definition of the operational parameters and requirements is still necessary. With continued collaboration with SRNL, FIU aims refine said definitions and develop new standards by which this, and other decommissioning technologies, can be accredited by relevant standards based testing. (authors)

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DOE OSTI2020

Investigation of Polyurethane and Fire-Retardant Foams as a Radiological Contamination Fixatives - 20202

Fixation of radiological contamination in decommissioning and deactivation can reduce worker risk, and mitigate potentially hazardous conditions; however, nearly every marketed contamination fixative has been found to be flammable, a significant concern in radiological facilities. Coupled with this, industry fixatives are normally used as a thin coating which can present problems when attempting to stabilize irregular void spaces or areas that are difficult to access and ensure full coverage. To this end, Savannah River National Laboratory (SRNL), in conjunction with Florida International University (FIU), have begun investigating the applicability of polyurethane based foaming materials, which will expand to fill a given asymmetric volume and provide fire retardancy to ensure no release of contamination in the event of a facility fire. Commercial foams undergo exothermic reactions resulting in gaseous release that subsequently causes the curing foam material to expand. Heat generation during curing can potentially compromise rubber parts of an enclosure, a significant concern in nuclear facilities. To mitigate large body heat generation, SRNL has also examined multiple layering of the material to ensure compatibility of the curing material with a previously cured body of the same material. Environmental studies were also completed in order to determine the influence on polyurethane based foams curing process. These studies showed that nearly half of the materials tested delaminated from a cured body of the same material, a property not discussed in commercially available documentation on the products. Of note was the compatibility of the foaming material with the substrate, depending on the type material that the foam was cured in/on, significant delamination could occur. Mechanical testing via tensile pull tester, dynamic mechanical analysis (DMA), and thermogravimetric analysis (TGA) was also performed. (authors)

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DOE OSTI2020

Extraction of Ln(III) and An(III) from Alkaline Solutions by Sulfonamide Ligands and Analogs - 20216

As part of our effort to develop advanced separations involving combined extraction of cesium, strontium, and actinides from High-Level Waste (HLW) at the Savannah River Site, we are now reporting highly efficient extraction of Sm(III) from aqueous media of various alkalinity by dichloromethane solutions of several ligands, including tri-sulfonamides, o-mono-sulfonamido-phenols and o-phenylenediamine- derived disulfonamides. Distribution experiments and DFT calculations provide insights on the complexation of Samarium vs Americium by these ligands. Further optimization of these ligands for extraction into highly lipophilic solvents and application to HLW simulants are underway. (authors)

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DOE OSTI2020

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

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

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

The Use of Silver Chloride Injection in Remediation of Iodine-129 by In Situ Capture as Silver Iodide at the F-Area Seepage Basin - 20225

The Savannah River Site (SRS) produced tritium, plutonium, and special nuclear materials for national defense, medicine, and the space programs. As part of operations, the F-Area Seepage Basins operated until 1988 for the disposition of deionized acidic wastewater from the F Separations Facility. The wastewater contained dilute nitric acid and low concentrations of non-radioactive metals, and radionuclides, with the major isotopes being Cs-137, Sr-90, U-235, U-238, Pu-239, Tc-99, I-129, and tritium. The seepage basins were closed in 1988 and backfilled and capped by 1991. The groundwater emanating from beneath the closed and capped seepage basins is acidic and contains elevated levels of both chemical and radiological contaminants. Releases from the groundwater plumes sourced from the F-Area Seepage Basins have impacted the water quality of Fourmile Branch, which is a small tributary to the Savannah River, a regional water source. A large pump-and-treat system was constructed in 1997 and operated until 2003 in an attempt to capture the releases to Fourmile Branch. The system in F Area and a similar system in H Area were expensive (∼$1.3 M/month) to operate and produced large quantities of radioactive waste, with concentrations of I- 129 too high to be disposed of at the SRS. In 2004, SRS replaced pump-and-treat with a funnel and gate system that along with operation of a base injection system at the gates reduces the flux of contaminants to the wetlands adjacent to Fourmile Branch. The alkaline solution injected into the aquifer neutralizes the acidic plume and immobilizes many of the cationic constituents. However, base injection is not effective in managing the release of iodine-129, an anionic contaminant. To address iodine-129 SRS and the Savannah River National Laboratory developed an in situ technology that uses ultra-fine ground silver chloride (AgCl) as an injectable capture medium for the sequestration of iodine-129. The AgCl amendment has a very small particle size and is designed to be injected into the contaminated aquifer to capture iodine-129. Dissolved iodine-129 forms a stable and highly insoluble solid (silver iodide) upon contact with AgCl. Laboratory studies, a field scale pilot test (2009), and three deployments (2011, 2015 and 2019) of AgCl have been successfully performed at the F-Area Seepage Basins. (authors)

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

Impact of UV-light and pH on the Fate of Tc, I, and U in Wetlands at Savannah River Site - 20230

The Savannah River Site (SRS) is one of the major nuclear facilities owned by the U.S Department of Energy. During the Cold War, these facilities produced large amounts of radioactive and hazardous waste. On site, three unlined seepage basins in the F-Area received approximately 1.8 billion gallons of low-level waste containing nitric acid, radionuclides, and dissolved metals due to plutonium and tritium production operations. The acidic nature of the waste created a source of relatively mobile radionuclides below the basins. Radionuclides previously disposed of within the F-Area, including uranium isotopes (U), technetium-99 (Tc-99), and iodine-129 (I-129), are moving with groundwater towards Four Mile Branch Wetland, where they are subsequently upwelling and interacting with natural organic matter (NOM). Many environmental factors including redox conditions, porewater composition, pH, light, and temperature affect the degradation of organic matter as well as interactions with Tc-99, U, and I-129. In particular, the high concentrations of nitrate from the acidic waste may increase the formation of reactive oxygen species (ROS) that impact both degradation of NOM and behavior of contaminants. In the presence of sunlight, I-129, Tc-99, and U speciation and complexation behavior may be affected by ROS and NOM degradation products in addition to the factors generally considered in subsurface systems in the absence of light. This research aims to determine whether the interactions between radionuclides, NOM, and nitrate affect the fate of I-129, Tc-99, and U and which processes are controlling their behavior. Experiments were conducted at variable pH in the presence of NaNO{sub 3} in order to determine the impact of light and pH on NOM degradation and to evaluate the impact on the fate of contaminants of concern. Soil samples high in NOM collected from two sites in the Southeastern United States (Savannah River Site and the Everglades) were studied. Batch experiments were conducted with NOM and the following aqueous contaminants, U, I-129, Tc-99 with exposure to ultraviolet (UV) light in an environmental chamber. This study was conducted in a sterile environment in order to exclude the potential for microbial degradation of organic matter. Results indicated photodegradation of NOM and significant interaction of radionuclides with NOM. (authors)

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

Structural Effects of Postulated Hydrogen Explosions in Process Piping and Vessels - 20274

At Savannah River Site (SRS), preparations to startup new liquid waste operations and recent advances in understanding thermolytic hydrogen generation have led to several recent studies on the structural effects of postulated hydrogen explosions. This paper reports on seven representative piping and vessel cases that are evaluated for deflagrations, detonations, and deflagration-to-detonation transition (DDT) assuming the initial mixture is stoichiometric hydrogen and air. The evaluations are based on the approach used by the Department of Defense for analyzing structures subjected to accidental explosions and include inelasticity, dynamic structural amplification, and material strain-rate effects. Evaluation criteria consider the potential for repeated explosions causing fatigue but also consider that some degree of permanent plastic deformation is acceptable as long as rupture is prevented. Specific conclusions depend on the assumed pre-ignition pressure. If the pre-ignition pressure is atmospheric and if the components are designed to a code from the American Society of Mechanical Engineers, the components evaluated are expected to plastically deform but not rupture under the explosion scenarios considered. (authors)

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

Effect of Evaporation on Vapor Space Flammability in DWPF Process Vessels - 20278

The Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) processes liquid high-level waste (HLW) and encapsulates it in glass within stainless-steel containers suitable for safe, long-term storage and disposal. Radiolytic and chemical decomposition of the liquid waste generates hydrogen and other flammable gases creating a hazard. The rate of flammable gas generation increases by an order of magnitude as temperature increases from sub-boiling to boiling. DWPF controls this hazard in process tanks by ensuring a minimum purge flow through each tank's head space. The minimum purge flow rate is established assuming dry air and neglecting evaporation. However, this study determined that steam from boiling in a representative process tank is sufficient to dilute flammable gases and, thereby, preclude flammability. For sub-boiling, water vapor is an insufficient diluent considering hydrogen's lower flammable limit (LFL) in air. However, the upper and lower flammable limits approach each other in a hydrogen-air-water-vapor environment as the water vapor concentration increases. This effect may preclude flammable conditions if water vapor concentration is about 60 volume percent or higher. A saturated, constant-pressure air-water system would achieve 60 volume percent water vapor at 86 deg. C. (authors)

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

Ion-Exchange Modeling of Crystalline Silicotitanate for Cesium Removal - 20283

The Tank Closure Cesium Removal (TCCR) system is a Savannah River Site (SRS) demonstration 'at-tank' process designed to remove {sup 137}Cs from the high-level aqueous tank waste so that the decontaminated solution can be disposed as low-level waste. Cesium is removed by ion exchange (IX) columns using engineered IONSIV{sup TM} R9120-B form of the Crystalline Silicotitanate (CST) media. The TCCR system is deployed at Tanks 10 and Tank 11 in the SRS H Tank Farm. Water is added to the salt-cake in Tank 10 H to dissolve it. The dissolved salt solution waste is pumped out of Tank 10H (feed tank), through filters and IX columns. The decontaminated salt solution is transferred to Tank 11 (receipt tank), and on to Tank 50H for final disposal in the Saltstone Production Facility. The current TCCR can accommodate lead-lag (two-column) or lead-lag-guard (three-column) configurations to optimize media utilization and achieve the target decontamination. To assist the TCCR operations, a parametric study was conducted to evaluate the impact of different parameters (e.g., column configurations (single column, two columns or three columns in series), waste characteristics, operating temperature, process flow rate, CST average particle size) on the IX column performance including CST bed utilization. The initial results indicate that the IX column performance is improved at slower process flow rate, at lower operating temperature, and with smaller CST average particle size. Multi-column configurations are recommended, because the single-column configuration does not utilize CST bed effectively. This paper demonstrates the versatility of the ion exchange modeling to evaluate the effects of CST characteristics and operational parameters on IX column performances. (authors)

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

Utilizing Commercial Submersible Mixer Pumps for Sludge Removal in Savannah River Site's Tank 26 - 20289

The Savannah River Site (SRS) Liquid Waste System (LWS) safely manages, stores, treats, and dispositions liquid radioactive waste. The LWS consists of 51 underground waste storage tanks (eight of which are operationally closed and filled with grout), waste evaporators, treatment facilities, and solidification facilities, known as the Defense Waste Processing Facility (DWPF) and Saltstone Production Facility (SPF). One of the waste storage tanks, Tank 26, was placed into service in 1980 as an F Tank Farm (FTF) Evaporator Feed Tank. From 1980 to 2013, Tank 26 received F Canyon receipts and dilute supernate that was then transferred to the 242-16F FTF Evaporator for volume reduction. In 2013, the steam tube bundle failed in the FTF Evaporator, and the decision was made not to repair/resume evaporator operation. During operation of Tank 26, solids built up to a level of 2.03 m in the tank with a volume of 1062.9 kL. Due to the evaporator failure and subsequent shutdown, the solids in Tank 26 were selected to feed Sludge Batch 10, which is collected and prepped prior to being sent to DWPF for final disposition. To accomplish solids removal, Commercial Submersible Mixer Pumps (CSMPs) were selected to slurry the solids in preparation to be sent to Tank 51. Four CSMPs were installed in Tank 26 with each one installed in a separate quadrant of Tank 26. This work would mark the first deployment of CSMPs in an SRS waste tank. The CSMPs were developed in response to operational issues from previous mixing pumps and budgetary constraints. The CSMPs use the concept of modifying commercially available equipment for nuclear waste applications. The CSMPs consist of a 230-horsepower submersible mixing pump (manufactured by GPM, Inc.) joined to a mast fabricated at SRS. For this application, Savannah River Remediation (SRR) design services was challenged to provide a mast design that required no steel superstructure support system on top of Tank 26. Also, SRR design services provided a simplified Tank 26 riser interface that did not require extensive machine parts to adjust height of the suction screen of the CSMP within the tank. Other design improvements include no requirements for tank top radiation monitors and no requirements for flushing of the CSMPs during startup/shutdown operations. The CSMPs were operated at an initial height of 2.29 m above the Tank 26 bottom for 10 days. After sludge sounding, the CSMPs were lowered to a height of 1.52 m with no issues due to proper work planning and the simplified riser interface. With the CSMPs lowered to a level of 1.52 meters above the tank bottom, the CSMPs were operated for 12 days and another sounding was performed. The sounding level came back matching the disturbance depth results of the first pump run. Chemistry samples were analyzed while the tank was left to settle. The analysis revealed enough weight percent solids to provide good feed to Sludge Batch 10. Also, the results revealed the supernate within the mixture would provide good salt solution feed to Salt Waste Processing Facility (SWPF) Salt Batch 3. So, the plan was modified to let the disturbed solids settle and decant the salt solution for addition to SWPF Salt Batch 3. Then water will be added back to Tank 26, and the CSMPs run in order to wash the solids to decrease settling time prior to being sent to Tank 51 for Sludge Batch 10. Utilizing CSMPs for waste removal in an SRS Tank has provided a cost-effective means for further waste removal efforts. In addition, the CSMPs are easier to operate by utilizing a robust and simplified design. The CSMPs performed quite well with no process shutdown or delays during operation. As a result, CSMPs are integral to the future of removing radioactive waste from storage tanks at SRS. (authors)

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