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DOE OSTI2022

From legacy contamination to watershed systems science: a review of scientific insights and technologies developed through DOE-supported research in water and energy security

Abstract Water resources, including groundwater and prominent rivers worldwide, are under duress because of excessive contaminant and nutrient loads. To help mitigate this problem, the United States Department of Energy (DOE) has supported research since the late 1980s to improve our fundamental knowledge of processes that could be used to help clean up challenging subsurface problems. Problems of interest have included subsurface radioactive waste, heavy metals, and metalloids (e.g. uranium, mercury, arsenic). Research efforts have provided insights into detailed groundwater biogeochemical process coupling and the resulting geochemical exports of metals and nutrients to surrounding environments. Recently, an increased focus has been placed on constraining the exchanges and fates of carbon and nitrogen within and across bedrock to canopy compartments of a watershed and in river–floodplain settings, because of their important role in driving biogeochemical interactions with contaminants and the potential of increased fluxes under changing precipitation regimes, including extreme events. While reviewing the extensive research that has been conducted at DOE’s representative sites and testbeds (such as the Oyster Site in Virginia, Savannah River Site in South Carolina, Oak Ridge Reservation in Tennessee, Hanford in Washington, Nevada National Security Site in Nevada, Riverton in Wyoming, and Rifle and East River in Colorado), this review paper explores the nature and distribution of contaminants in the surface and shallow subsurface (i.e. the critical zone) and their interactions with carbon and nitrogen dynamics. We also describe state-of-the-art, scale-aware characterization approaches and models developed to predict contaminant fate and transport. The models take advantage of DOE leadership-class high-performance computers and are beginning to incorporate artificial intelligence approaches to tackle the extreme diversity of hydro-biogeochemical processes and measurements. Recognizing that the insights and capability developments are potentially transferable to many other sites, we also explore the scientific implications of these advances and recommend future research directions.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2024

Soil Desiccation Treatability Testing at BC Waste Disposal Cribs

During Hanford’s production period, low-level waste products generated from chemical processing of uranium fuel rods were discharged directly to the ground through a system of cribs and trenches located in the 200-BC-1 Operable Unit (OU). The site consists of 6 cribs and 20 trenches that received more than 117,000 m3 of radioactive liquid waste discharged to the soil. These unlined infiltration galleries held volumes of liquid waste while it seeped into the ground, with the understanding that the 100 m (330 ft) thick vadose zone in the area would effectively capture the effluent and prevent groundwater impacts. A conceptual model showing the operation of cribs and trenches is shown in Figure 1. Data show effluent from the 26 cribs and trenches containing about 410 curies of Technetium-99 (Tc-99) is primarily located between 30 m and 70 m (98 ft and 230 ft) depth (Corbin et al., 2005; Ward et al., 2004). Despite no evidence indicating that the contamination has reached the groundwater at BC cribs and trenches, the mobility of Tc-99 had been demonstrated in laboratory tests and was recognized as a threat to groundwater at the site. Using data from numerical models, laboratory analyses, field investigations, and information on historical discharges, the EPA and Ecology identified Tc-99 and U contamination of the vadose zone as a remediation priority. The U.S. DOE was notified by EPA and Ecology regarding risks associated with Tc-99 contamination in a letter requesting development of a strategy for improved methods to understand the nature and extent of vadose zone contamination, specifically Tc-99, and to develop remedial options for addressing such contamination. To develop the appropriate technology for characterizing, remediating, and monitoring the deep vadose zone Tc-99 contamination, the U.S. DOE worked with the EPA and Ecology to create a Treatability Test Plan under a Remedial Investigation/Feasibility Study (RI/FS) for the Hanford 200 Areas. Under this RI/FS, it was determined that a treatability test for soil desiccation should be carried out as it was identified as a promising in-situ treatment technology for mitigating risks posed by Tc-99 contamination to the groundwater table. The BC Cribs and Trenches site was identified as a representative site for Tc-99 and U contamination and selected for the soil desiccation treatability test. In this chapter, we summarize the overlying regulatory framework of RI/FS and treatability tests and illustrate how development and experimentation supported the evaluation of selected remedies. We briefly discuss the RI/FS for the 200 Areas of the Hanford Site and focus on the soil desiccation treatability testing performed at the BC cribs and trenches site under the Deep Vadose Zone Treatability Test Plan for the Hanford Central Plateau (DVZ-TT). The DVZ-TT is one component of the remedial investigation/feasibility study for the Hanford 200 Areas and represents the underlying regulatory framework that drives site operations towards records of decision and site closure.

Mangel, Adam R.↗
DOE OSTI2020

Development of an In-Situ Leaching Technology for Extracting Residual Uranium from Remediated Soil - 20099

Site-specific chemical solutions for extracting uranium were developed and evaluated for treating previously remediated soil at a former thorium mill tailings site, known as the Former Rare Earths Facility (REF), in West Chicago, Illinois. Prior remediation efforts had resulted in residual sand and gravel aggregate that met the soil cleanup standards. The target soil contains residual levels of uranium, which continues to leach into the site groundwater at levels exceeding groundwater protection standards (GWPS). A series of bench-scale tests were conducted to evaluate the efficacy and viability of in-situ leaching (ISL) technology to treat the target soil for the purposes of attaining uranium GWPS at the REF. Specific objectives of the bench-scale tests entailed identification and selection of optimal leaching, quantification of the extent and rate of uranium released as a function of chemical composition, chemical concentrations, chemical residence times, and obtaining critical data to estimate treatment costs. A combination of sodium bicarbonate and sodium carbonate reagent was the most effective leaching solution. In general, leaching solutions containing a higher concentration of sodium carbonate demonstrated superior uranium leaching capabilities. Removal of 50% of the uranium mass from the soil matrix was adequate for achieving compliance with the site-specific GWPS. After six to seven pore volume flushes (PVFs), the uranium leaching effectiveness of all leaching solutions was similar. The amount of initial uranium removed with each subsequent PVF was minimal, suggesting limited solubility of the remaining mass of uranium. The effects of leaching duration were significant and persisted through at least the first five consecutive PVFs. Leaching of uranium also resulted in leaching of several non-uranium constituents, many of which exceeded applicable National Pollutant Discharge Elimination System (NPDES) standards. Given historical operations and waste disposal practices at the site, leaching of non-uranium constituents was expected. Hydrochloric acid was used to stabilize treated soil. Savings of approximately $148 million in direct and indirect remediation costs are expected with an ISL alternative compared to the excavation and landfill disposal remediation option. (authors)

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

Remediation of Temporary Storage Sites in Support of the Port Hope Area Initiative - 20295

The Port Hope Area Initiative is a community-based solution for the long-term management of historic low level radioactive waste (LLRW) resulting from 60 years of uranium and radium processing operations in the Town of Port Hope which is located in Ontario, Canada. The Eldorado refinery, on the north shore of Lake Ontario, began refining radium-226 from pitchblende ore, later transitioning to the refining of uranium. Through the history of the operation, LLRW was deposited throughout the town of Port Hope as a result of fugitive emissions from the plant and/or through the re-use of process residues as building material and backfill. Historical clean-up activities conducted in the late 1970's involved the remediation of approximately 400 properties and the relocation of 100,000 cubic metres of contaminated soil to a disposal facility in Chalk River operated by Atomic Energy of Canada Limited (AECL). Owing to space limitations at that disposal facility, any LLRW identified through construction monitoring since that time has been stored in the community at three temporary storage sites located throughout the town. These include: the Centre Pier mound that contained approximately of 19,800 m{sup 3} of LLRW-impacted soil that originated from the construction of a new water treatment plant; two mounds located at a licensed storage facility containing LLRW obtained from residential clean-up activities (11,000 m{sup 3}); and a small pad adjacent to the municipal sewage treatment plant containing 2200 m{sup 3} of LLRW-containing sludge. With the construction of a new long-term waste management facility (LTWMF) that has been designed to house all of the LLRW identified within Port Hope, the three sites were early candidates for remediation. The clean-up of the three temporary storage sites was a significant milestone for the Port Hope Area Initiative. After a decade of planning and consultation, this work represents the first sites in the municipality to be remediated with the waste being safety removed and transferred to the newly constructed LTWMF. This paper discusses the challenges associated with the clean-up activities for these three sites and the strategies employed to address those challenges. These included weather-related challenges, owing to the seasons over which the work was conducted as well as those associated with working within a closely-knit community. Canadian Nuclear Laboratories (CNL), working on behalf of the federal government, has worked diligently to develop a positive and trusting relationship with the community. Consequently, the successful execution of this project needed to be sensitive to, and respectful of the needs of the community. In addition to the usual Health, Safety and Environment training, project staff received community awareness training that spoke to the history of this community-based initiative and the expected behavior when working within the community. Transportation routes were defined based on safety and the need to minimize disruption to local traffic while haul-times where scheduled around school bus hours to enhance public safety. The successful completion of this first of many remediation projects to be completed under the Port Hope Area Initiative reflected years of careful planning. Nevertheless, there were a number of 'lessons learned' that have been applied on other ongoing projects be completed under the Port Hope Area Initiative. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Irradiation Impact on Uranium Recovery Under Direct Extraction Conditions

Reducing the quantity of high-level radioactive waste is essential for minimizing environmental impact and improving efficiency of using natural resources for nuclear power. The current standard, Plutonium Uranium Solvent EXtraction (PUREX), uses tributyl phosphate (TBP) ligands to extract complexes of uranium and plutonium from a nitric acid (HNO3) phase. Although this method is effective, large volumes of HNO3 and the non-incinerable phosphate ligands increase the amount of hazardous waste produced. Alternative extractants and flowsheets have been proposed that allow for more selective extraction of radioactive metals, reduced nitric acid use, and easier incineration by only containing carbon, hydrogen, oxygen, and nitrogen (CHON). One candidate, N,N-di(2-ethylhexyl)-isobutyramide (DEHiBA) exhibits promising properties for direct extraction. A HNO3 pre-equilibrated DEHiBA phase selectively extracts U(VI), leaving plutonium, transuranics, and fission products behind as precipitate and reducing the volume of radioactive HNO3 produced.

38 - RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCL↗
DOE OSTI2024

Recommendations for an Applicant to Calculate Activity Data for Greenhouse Gases Estimates

In 2009, the U.S. Nuclear Regulatory Commission (NRC) directed the NRC staff to address climate change issues and consider the impacts of the emissions of carbon dioxide (CO 2 ) and other greenhouse gases (GHGs) in its environmental reviews for major licensing actions (NRC 2009b). To implement this direction from the Commission, the staff issued guidance in 2011 and updated guidance in 2014 in Attachment 1 to Interim Staff Guidance COL/ESP-ISG-026 (NRC 2011; NRC 2014). This guidance provides a simpler method than the method described in RG 4.2 Rev. 3, that an applicant can use to meet the plant parameter envelope (PPE) value from the Generic Environmental Impact Statement for Licensing of New Nuclear Reactors (NR GEIS). NRC staff estimated the 97-year lifecycle GHG emissions from a reference 1000 megawatt electrical (MWe) light-water reactor (LWR) for various activities associated with construction, operation (including uranium fuel cycle), and decommissioning of nuclear power plants and presented the results in Appendix H of the NR GEIS. Appendix H of the NR GEIS includes estimates of direct emissions from construction equipment and emergency diesel engines in a nuclear facility and indirect emissions from workforce vehicular traffic, fuel transportation and the uranium fuel cycle. The NR GEIS Section 3.3 extended the estimates in Appendix H for the installation of two 1000 MWe nuclear reactors on the same site. Scaling factors were used to extrapolate the GHG emissions of a reference 1000 MWe reactor to a two-unit nuclear reactor plant (each reactor unit generating 1000 MWe). GHG emission estimates for building, operation, decommissioning and safe storage (SAFSTOR) for a two-unit nuclear reactor plant would be based on the plant’s physical size, and therefore estimates for these source categories were assumed to be twice the value of the reference 1000 MWe reactor. However, GHG emissions from the fuel cycle (including fuel transportation) were scaled upward by a factor of 3, based on plant efficiencies greater than the 80 percent assumption in Appendix H. Table 1 below shows the PPE emissions for two 1000 MWe nuclear reactors as provided in NR GEIS. The total GHG emissions for two 1000 MWe reactors were calculated as 2,534,000 metric tons (MT) of CO 2 equivalent (CO 2 (e)) based on a 97 year GHG life cycle period. The GHG emissions lifetime of 97 years for a reference nuclear reactor includes a 7-year building phase, 40 years of operation, 10 years of active decommissioning, and 40 years of SAFSTOR operations (NRC 2024). Construction equipment and vehicular traffic from workers commute would contribute to the GHG emissions during a 7-year building phase. Uranium fuel cycle, vehicular traffic, fuel and waste transportation, and testing of standby diesel generators would contribute to GHG emissions during the 40-year operations phase. While NRC’s regulations allow up to 60 years of reactor facility decommissioning, Appendix H estimated that most of the GHGs would occur over an estimated 10-year period during which to the licensee would engage in significant demolition and earth-moving activities, as discussed in Supplement 1 to NUREG-0586 (NRC 2002). Vehicular traffic by the workforce during a 40-year SAFSTOR period would additionally contribute GHG emissions. The carbon footprint for a 40-year SAFSTOR period was separately analyzed from the decommissioning activities as provided in Table YYYY-2 of the staff issued guidance in 2011 (NRC 2011).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Characterization and CST Batch Contact Equilibrium Testing of Modified Tank 9H Process Supernate Samples in Support of TCCR

The Tank Closure Cesium Removal (TCCR) system uses ion exchange columns filled with crystalline silicotitanate (CST) media to process radioactive waste solutions for the removal of ¹³⁷Cs. TCCR currently focuses on dissolving Savannah River Site (SRS) radioactive tank waste (primarily sodium saltcake solids) within Tank 10H followed by at-tank ion exchange column treatment. Two supernate batches from Tank 10H have been processed through the TCCR unit and processing of a third batch is expected soon. After processing of this supernate batch, plans are to replace the CST columns and process dissolved salt solution from Tank 9H through Tank 10H and then through the new CST columns installed in the TCCR unit. The new columns are expected to contain either a media similar to an archived CST batch (IE-911) or the R9120-B CST media used in the current TCCR columns (the two materials are fundamentally the same; just different specifications, product names, and preconditioning steps). Samples of Tank 9H dissolved saltcake were received at the Savannah River National Laboratory (SRNL) and characterized. The Tank 9H supernate contained a high sodium concentration (~9.6 M Na⁺) and will require dilution to near 6 M [Na⁺] prior to processing through the TCCR unit. The cesium concentration in Tank 9H is currently significantly higher than was observed with Tank 10H. Three dilutions of the Tank 9H supernate were conducted to mimic possible dilutions that could be conducted in the tank farm prior to TCCR processing using inhibited water, sodium hydroxide, and sodium nitrate solutions. Dilution #1 was prepared by diluting the Tank 9H supernate by a factor of 1.6 with inhibited water, Dilution #2 was prepared by diluting the Tank 9H supernate by a factor of 3.7 with a mixed sodium hydroxide/sodium nitrate diluent, and Dilution #3 was prepared by also diluting by a factor of 3.7 but with sodium hydroxide only. All three dilutions have similar Na⁺ concentrations (~6 M), but Dilutions #2 and #3 contain significantly less cesium. Major and key minor components of the diluted Tank 9H supernate samples are provided in Table ES-1. Batch contact equilibrium tests were conducted with the Tank 9H dilutions and the two different CST media batches being considered for use in the new TCCR columns. Results are summarized in Table ES-2 and compared to ZAM model predictions. The highest cesium distribution coefficient and percent removal were observed with Dilution #3. IE-911 CST (an archived CST media batch) was more effective at removing cesium than the more recently prepared R9120-B CST media, though both media samples removed >88% of the cesium and the differences may not be statistically different considering the overall uncertainty. Based on the results, maximum cesium loadings were calculated for each CST media type and Tank 9H dilution. In general, maximum cesium loadings from dilutions of this supernate batch are quite high (approaching 0.1 mmol total Cs⁺/g CST for Dilution #1). The highest calculated maximum ¹³⁷Cs loading for the Tank 9H dilutions using the ZAM model with input of the tank compositions and batch contact results was 207 Ci/kg CST (Dilution #1 with IE-911 CST). In all cases, higher maximum cesium loading values were predicted for IE-911 CST versus R9120-B, although the differences varied considerably between the dilutions. The maximum loading value for IE-911 CST with Tank 9H Dilution #1 was only 7% higher than the maximum loading for R9120-B. The maximum loading value for IE-911 CST with Dilution #3 was 24% higher than the maximum loading for R9120-B. Dilution #2 was intermediate between these values. Note that these maximum loading values are the theoretical calculated values, and actual operating conditions will cause differences. A CST binder dilution (correction) factor near 0.7 (relative to pure powder CST) was required for each batch contact test with IE-911 engineered CST using the three Tank 9H dilutions. Correction factors calculated for R9120-B CST ranged from 0.56 to 0.66 for the three dilutions. Following the batch contact tests, the CST samples were isolated from the Tank 9H solution, washed, dried, and digested in acid following established procedures. The analysis results are provided in Table ES-3. Total cesium loading values determined by CST digestion were similar to the calculated loading values based on solution analysis (Table ES-2) for all samples. In addition, the CST was observed to load calcium (R9120-B sample only), iron, strontium, lead, uranium, and plutonium after contact with the waste supernate, as has been observed previously.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Development of the technological process for the IGR reactor's highly-enriched irradiated uranium-graphite fuel immobilization

The immobilization of irradiated highly enriched uranium (HEU) fuel is a critical component of nuclear waste management and non-proliferation efforts. In Kazakhstan, at National Nuclear Center of the Republic of Kazakhstan special attention is given to managing legacy HEU fuel from research reactors. One such case involves the IGR research reactor, whose first core containing irradiated HEU uranium-graphite fuel was operated from 1961 to 1966 and removed following reactor modernization. This fuel now requires a reliable and secure immobilization strategy. Here, this paper presents the development of a technological process for immobilizing this fuel to reduce its enrichment to below 20 % in terms of 235U content. The proposed method involves down-blending irradiated HEU fuel with depleted uranium, followed by encapsulation in a Portland cement matrix. Full-scale experiments were conducted to assess the uniformity of uranium distribution within the matrix. The results confirm the effectiveness of this approach, ensuring reliable immobilization of fuel in accordance with international requirements, including IAEA standards and Kazakhstan's regulatory framework. These findings contribute to the broader effort of adapting immobilization strategies for the safe management of spent fuel from research reactors.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗
DOE OSTI2024

Design of Hypothetical Processes for the Production of 131 I and 99 Mo from Activation Targets

For over six decades, medical isotope production has been a high-priority focus of many research reactors across the globe. The majority of these isotopes were produced using highly-enriched uranium (HEU) or low enriched uranium (LEU) – delivering millions of doses of diagnostic and therapeutic isotopes. As a consequence of this production, however, six decades of isotope production has resulted in massive quantities of spent uranium material worldwide with no known disposition pathway creating growing proliferation concerns. Supported by the National Nuclear Security Administration’s (NNSA) Material Management and Minimization (M3) program, there has been increased focus in the production of high-priority isotopes without special nuclear materials or without uranium altogether. Isotope production via activation can potentially fulfill regional isotope demands – particularly in under-developed regions without access to isotope supply chains. The benefits of this approach would be a reduction in uranium proliferation risks, less special nuclear material wastes, and reduced risk of supply disruption in the likely event that major isotope producers will again go off-line as has happened in recent years due to a number of factors.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

Uranyl oxalate species in high ionic strength environments: stability constants for aqueous and solid uranyl oxalate complexes

Uranyl ion, UO 2 2+ , and its aqueous complexes with organic and inorganic ligands can be the dominant species for uranium transport on the Earth surface or in a nuclear waste disposal system if an oxidizing condition is present. As an important biodegradation product, oxalate, C 2 O 4 2– , is ubiquitous in natural environments and is known for its ability to complex with the uranyl ion. Oxalate can also form solid phases with uranyl ion in certain environments thus limiting uranium migration. Therefore, the determination of stability constants for aqueous and solid uranyl oxalate complexes is important not only to the understanding of uranium mobility in natural environments, but also to the performance assessment of nuclear waste disposal. In this work, we developed a thermodynamic model for the UO 2 2+ –Na + –H + –Cl – –ClO 4 – –C 2 O 4 2– –NO 3 – –H 2 O system to ionic strength up to ~11 mol•kg –1 . We constrained the stability constants for UO 2 C 2 O 4 (aq) and UO 2 (C 2 O 4 ) 2 2– at infinite dilution based on our evaluation of the literature data over a wide range of ionic strengths up to ~11 mol•kg –1 . We also obtained the solubility constants at infinite dilution for solid uranyl oxalates, UO 2 C 2 O 4 •3H 2 O, based on the solubility data over a wide range of ionic strengths. The developed model will enable for the accurate stability assessment of oxalate complexes affecting uranium mobility under a wide range of conditions including those in deep geological repositories.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Deliquescence of eutectic LiCl-KCl diluted with NaCl for interim waste salt storage

Molten eutectic LiCl-KCl salt is a widely used electrolyte for electrorefining uranium from spent nuclear fuel. Due to the hygroscopic nature of this salt, careful handling and storage of solidified waste electrorefiner (ER) salt is required to avoid deliquescence and corrosion of container materials. This study investigated a potential processing path for reducing the deliquescence through dilution with NaCl. The hydration behavior of LiCl-KCl salts diluted with NaCl was evaluated in terms of mass gain due to mass of water absorption, visual evidence of deliquescence, and visual evidence of corrosion to stainless steel containers in a humid air environment. Pure eutectic LiCl-KCl exhibited a 50 mass% increase due to water absorption and showed signs of deliquescence after 24 hours. 89 mass% NaCl was required in order to prevent deliquescence. Dilution with 89% NaCl was also found to protect corrosion to stainless steel crucibles. Corrosion, thus, appears to require deliquescence. While NaCl dilution greatly decreases steady state hydration and eliminated deliquescence, storage volume is increased ~10x.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Chloride-Based Volatility for Waste Reduction and/or Reuse of Metallic-, Oxide- and Salt-Based Reactor Fuels

The objective of the chloride based volatility project (CBV) was to demonstrate the ability to separate uranium from used fuel to enable process improvements resulting in 10x reduction of waste volume, while maintaining safeguards standards and global backend costs at $\$$1/MW-hr. Current industrial practices perform separations of used fuel using solvent media in the form of aqueous and molten salt processes, resulting in contaminated process waste. The CBV approach utilized solid state chemistry with no solvent media and was successful in chlorinating uranium and fission product oxides that sublimed into the gas phase and were collected in targeted condensation zones based on temperature gradients of chlorinated products. Recovery of better than 95% of initial uranium in the form of UCl4 was demonstrated when simulated used nuclear fuel was used. Laser induced breakdown spectroscopy, LIBS, and ultraviolet-visible spectroscopy, UV-Vis, were combined into a high temperature flow cell design and utilized as process monitoring techniques to observed chlorinated products leave the reaction vessel in real time.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Deliquescence of Eutectic LiCl-KCl Diluted with NaCl for Interim Waste Salt Storage

Molten eutectic LiCl-KCl salt is a widely used electrolyte for electrorefining uranium from spent nuclear fuel. Due to the hygroscopic nature of this salt, such operations must be performed under controlled atmospheric conditions, and waste salts require careful storage to avoid deliquescence and corrosion of container materials. This study investigated a potential processing path for reducing the degree of deliquescence through dilution to varying extents with NaCl. Here, the hydration behavior of LiCl-KCl salts diluted with NaCl was evaluated in terms of mass gain due to water absorption, degree of deliquescence (including first appearances of standing water), and evidence of corrosion to stainless steel containers in a humid air environment (40°C, 20% relative humidity). In this humid air environment, pure eutectic LiCl-KCl exhibited a 50 mass % increase due to water absorption and showed evidence of standing water after 24 h. Waste salt diluted with NaCl required loadings of 89 mass % NaCl in order to prevent deliquescence and exhibited a 3 mass % increase due to water absorption. After periodic observation for 48 h, standing water was observed near all ingots with the exception of the 89 mass % NaCl samples. Dilution with 89% NaCl was also found to reduce evidence of corrosion when stored in stainless steel crucibles. While dilution with NaCl greatly decreases steady-state hydration, the storage volume is increased ~10× through this procedure.

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

Sodium-Cooled Fast Reactor Proliferation Resistance and Physical Protection White Paper

The Sodium-Cooled Fast Reactor (SFR) system was identified during the Generation IV Technology Roadmap as a promising technology to perform the actinide management mission and, if enhanced economics for the system could be realized, also the electricity and heat production missions. The main characteristics of the SFR that make it especially suitable for the actinide management mission are: Consumption of transuranics in a closed fuel cycle, thus reducing the radiotoxicity and heat load which facilitates waste disposal and geologic isolation; Enhanced utilization of uranium resources through efficient management of fissile materials and multi-recycle; and, High level of safety achieved through inherent and passive means that accommodate transients and bounding events with significant safety margins.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Thermal Process Technology for Nuclear Applications - 20379

ANSTO's Synroc technology has been developed to provide a safe, secure matrix for the immobilization and final disposal of radioactive waste. Synroc technology will be used to manage radioactive wastes from the production of the radioisotope Molybdenum-99 (Mo-99). This paper shall outline various stages of the process development with specific reference to the thermal treatment technology of calcination. Calcination is a key step in the Synroc process [1-2].The rotary thermal processing system includes: an advanced heating element design for increased robustness and ease of remote operation and maintenance, an enhanced modular design of components for ease of remote maintenance in a hot cell and in compliance with hot cell radioactive environment requirements for safety, reliability and maintainability. In addition to thermal treatment of waste from nuclear medicine production, this technology provides solutions for a variety of nuclear materials processing applications including sintering UO{sub 2} pellets for reactor fuel rods, oxidation of UO{sub 2} pellets, swarf, and powder to U{sub 3}O{sub 8}, de-nitration of Uranyl nitrate and hydrofluorination of UO{sub 2} pellets. The paper will also discuss thermal processing solutions for a range of nuclear applications. (authors)

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

Evaluation of a Combined Raffinate Waste Stream for Mo-99 Production

Niowave, Inc., based in Lansing, Michigan, intends to produce Mo-99 from the fission of low enriched uranium (LEU) targets. In FY20, SRNL was tasked to provide technical assistance in identifying the primary waste raffinate present in Niowave’s commercial scale system and evaluating methods for disposal. Due to limited information on Niowave’s complete flowsheet and isotope recovery plans, a combined raffinate stream including the entire radionuclide inventory from the irradiated fuel was used for this evaluation. Results of the evaluation indicated the majority of the waste generated can be expected to be Class A; however radioisotopes such as Pu-239, Cs-137, Sr-90, and some of the shorter lived (< 5 year half-life) may challenge Niowave’s ability to meet the waste acceptance criteria for land burial. With appropriate waste management strategies, it is probable that Niowave’s waste could meet Class C requirements for all constituents, with the possible exception of Pu-239. More detailed information of Niowave’s processes are needed to form a more accurate understanding required to execute and effective, economic, and feasible waste management strategy.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Characterization of a Uranium Groundwater Plume along the Columbia River

The 300 Area is adjacent to the Columbia River on the southeastern end of Hanford, approximately one mile north of the city of Richland. Uranium fuel fabrication took place in the 300 Area from 1943 to 1988. In this chapter, the historical waste releases that occurred within the 300 Area of the Hanford Site, and specifically the resulting uranium plume, will be used as a case study. There are 3 different operable units (OU) that were established in the 300 Area for cleanup purposes: The 300-FF-1 OU, 300-FF-2 OU and the 300-FF-5 groundwater OU. To determine the location of contaminant plumes within the Hanford Site’s 300 Area, several puzzle pieces must be put together using multiple resources. The first piece of the puzzle is determining the location and chemistry of waste disposed of in the area. Records beginning in the 1940’s can help piece this information together, and the known waste sites and waste disposed of are an example of that process knowledge. Characterization of the subsurface and groundwater is important to determine where the contaminants are now, and what form they are in. This information can be used to design a remediation strategy, as well as to inform modeling efforts to predict where the contaminants will move to next.

300-FF-5, uranium, 300 area, hanford book project↗
DOE OSTI2020

Analysis of Using a Hybrid ZIRCEX Process to Generate High-Assay Low-Enriched Uranium (HALEU) - 20158

Idaho National Lab (INL) is investigating the feasibility of providing High-Assay Low Enriched Uranium (HALEU) by recovering Highly Enriched Uranium (HEU) from spent nuclear fuel (SNF) and downblending the HEU with natural, low-enriched, or depleted uranium (NU, LEU, or DU). That investigation identified that significant quantities of HALEU could be produced with a Hybrid ZIRCEX process to further advanced reactor research while reducing INL's environmental liability. Many new advanced reactor designs are being developed with improved safety, efficiency, and economics. Most require nuclear fuel with U-235 enrichment between 5% and 20%, which is defined as HALEU. Although technically feasible, there is no current domestic capability to make HALEU in the US. While it is anticipated that industry will provide HALEU through commercial enrichment once advanced reactors mature, until a market is established, an interim source of HALEU is needed to enable research and demonstration. An interim source of HALEU could be provided by recovering the HEU in some DOE-managed SNF and downblending it to between 5% and 20% using the Hybrid ZIRCEX Process. ZIRCEX is a dry head-end process to remove cladding (zirconium or aluminum) from SNF. Drying of the SNF is needed prior to introduction to ZIRCEX. After cladding removal, uranium and fission products in the bed material are oxidized and elutriated. In a Hybrid ZIRCEX process, the HEU process stream is sent to a very compact, modular solvent extraction process for uranium purification. The uranyl nitrate from solvent extraction is denitrated and calcined to produce a HALEU product ready for fuel fabrication. The long-lived fission products from solvent extraction are immobilized in glass using a small in-can melt. By removing the cladding before dissolution, the volume of glass waste is reduced by a factor of up to 300 times (when compared to not removing the cladding from the fuel). Other radioactive wastes from the process could be disposed of as Class A or Class B low-level waste. In addition to producing HALEU for advanced reactors, consumption of SNF would reduce DoE's environmental liability. Incorporating low-enriched spent fuel (versus exclusive use of LEU, DU, or NU) into the downblending scheme could reduce the SNF inventory even further. Several downblending and enrichment scenarios using INL fuels were investigated in this study. The results of this analysis show that the Hybrid ZIRCEX process could make a significant contribution to needed HALEU feedstock and reduce the environmental liability of managing SNF at INL. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗