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DOE OSTI2020

Evaluation of Material Balance Approaches for Hanford Direct-Feed Low Activity Waste Processing - 20022

The Hanford Site has accumulated millions of gallons of tank waste from reprocessing spent fuel to recover plutonium, uranium, cesium, and strontium. The supernatant from the accumulated tank waste will be treated using a Direct Feed Low-Activity Waste approach. The supernatant will be treated to remove solids and cesium in the Tank-Side Cesium Removal process in the Hanford tank farm, then vitrified in a semi-batch process in the Hanford Waste Treatment and Immobilization Plant (WTP). Currently, each batch of feed is sampled at three locations prior to being fed to the melter: the feed qualification tank in the Hanford tank farm as well as the concentrate receipt vessel (CRV) and melter feed preparation vessel (MFPV) in the WTP. The feed qualification sample is taken from a large batch of accumulated feed, only two to three samples are expected each year. Approximately 275 samples from the CRVs and 1100 samples from the MFPV are expected each year. An evaluation was performed to determine if a material balance based on the feed qualification sample could replace most of the sampling in the CRVs and MFPVs. The evaluation consisted of three elements: (1) determination of the practicality of using a material balance to estimate the stream composition of the CRV and MFPV contents, (2) evaluation of whether the material balance could be automated using the existing process control system, and (3) determination of the uncertainty in glass composition using the material balance approach. It is assumed that periodic sampling at the CRV and MFPV would be performed periodically to re-baseline the material balance, evaluations are in progress to determine the frequency of this periodic sampling. Process sample locations downstream of the melter were reviewed as well, but the partitioning of semi-volatile species in the melter was determined to preclude extending the material balance approach past the melter. It was determined that replacement of the CRV sample location was feasible and did not increase process uncertainty or significantly impact waste loading. Replacement of the MFPV sample was also determined to be feasible, but that measurement of the glass former chemical addition may be needed prior to addition of these chemicals to the MFPV. This measurement could be performed by an in situ laser-induced breakdown spectroscopy (LIBS) system. Limited tests were performed to evaluate LIBS for direct measurements of the low-activity waste melter feed. The use of a material balance would eliminate over 1200 samples each year if only 10% of the CRV and MFPV batches are sampled and could likely allow the WTP laboratory to operate on days only versus 24/7 operation. (authors)

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DOE OSTI2015

Treatment of radioactive liquid waste by the forced-air exhaust system - 15190

A Natural Evaporation Facility (NEF) was constructed to treat low level (LL) radioactive liquid waste arising from a Radioactive Waste Treatment Facility (RWTF) at the Korea Atomic Energy Research Institute (KAERI). In order to process LL radioactive liquid waste, we investigated the influence of climate condition, mainly atmospheric humidity and temperature, on the amount of evaporation in a forced-air exhaust system in KAERI. Evaporation media were made of cotton and polyester. Air exhaust in the facility was forced by exhaust fans. The evaporation rate and the decontamination factor were calculated and reported as shown in the results section. The evaporation and air supply flow increased at the same rate. In addition, the liquid waste feed rate and temperature of supplied air increased. Regarding the relative humidity of supplied air, the evaporation rate increased as the relative humidity decreased. As the result of this study, the operating conditions for the NEF were optimized in the following areas: The air temperature above 8 deg. C, the air relative humidity below 70%, the air flow rate 1.14-1.47 m/sec, and the liquid waste feed rate 4.6 l/hr.m{sup 2}. The decontamination factor was 1.1x10{sup 3}. (authors)

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DOE OSTI2019

Treatment of Foreign Radioactive Waste at the BELGOPROCESS Plant - 19345

Belgium has used a large number of nuclear technologies during the last 60 years. Power generation, reprocessing, fuel fabrication, isotope production and research activities have produced a large variety of radioactive waste streams. BELGOPROCESS is the operator of the centralised waste treatment facilities located in the Campine region in Flanders. Over the years BELGOPROCESS has developed the know-how and facilities to treat each one of these waste streams. BELGOPROCESS is a subsidiary of NIRAS, the governmental institution responsible for the safe management of radioactive waste and enriched fissile materials in Belgium. A number of these waste streams are only produced during a short period of time. After processing the existing stockpile the treatment facility is put into operational standby and eventually prepared for decommissioning. In order to treat the stockpile of some types of legacy waste within an acceptable time frame, the treatment facilities needed to be designed at a certain treatment capacity. Since the majority of the legacy waste is processed nowadays, a significant amount of treatment capacity becomes available. In June 2006 the Belgian government decided that this know-how and facilities should become available to foreign waste producers. This would allow BELGOPROCESS to fill up the available capacity of its facilities and consequently lower the cost for the Belgian waste producers. Nowadays two facilities are open to accept foreign radioactive waste. The CILVA-facility can be used for sorting, cutting, incineration, precompaction, supercompaction and cementation of low level solid waste. The PAMELA-facility - a former vitrification plant - can be used for the treatment of medium level and alpha-bearing waste. In this paper we will discuss the technical characteristics of the CILVA-incinerator and a showcase for the treatment of foreign waste at our facilities. In 2016 BELGOPROCESS started a cooperation with MATE-FIN for the incineration of low level burnable waste from the Cernavoda NPP in Romania. (authors)

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DOE OSTI2016

Aging Infrastructure: Application of System Health Monitoring at the Defense Waste Processing Facility - 16069

The disposition of high level waste (HLW) into borosilicate glass canistered waste forms began at SRS in 1996. Since that time, the Defense Waste Processing Facility (DWPF) has produced nearly four thousand HLW canisters. The DWPF is now approaching 20 years of radioactive service and 25 years of operation since the commencement of nonradioactive testing in 1990. Many facility systems are challenged by equipment or component obsolescence, which places a premium on proper work planning, engineering, and maintenance resource allocation to ensure facility reliability. To this end, Savannah River Remediation (SRR) implements programs and procedures for performance monitoring of structures, systems, and components (SSCs) within all of its processing facilities. As these programs have matured, they have helped to identify vulnerabilities within facility systems and provided a framework for ranking system assets to ensure correct priority is established and maintained. This paper provides an overview of how SRR structures and implements system performance monitoring with specific focus on the recent replacement of the DWPF Safety Grade Purge Flowmeters. This case study will discuss: 1) How the performance monitoring process identified an availability risk due to the minimal vendor support for the in-use flow transmitters, 2) The design changes developed as part of the project to transition to a different measurement technology, which includes operational lessons learned from the original installation, 3) The execution of the replacement effort, which was complicated by the fact that the purge system is a Vital Safety System (VSS) and not easily taken out of service for extended maintenance, and 4) The need to consider future maintainability, particularly given that continued acceleration in technical innovation will likely increase the frequency of equipment obsolescence. (authors)

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DOE OSTI2020

Panel Session 55: US DOE Savannah River Operations Office: Liquid Waste Progress

This panel provided an overview of the SRS HLW Liquid Waste Program (history, challenges, opportunities, and future) including discussions of the Defense Waste Processing Facility (DWPF), Tank Closures, Actinide Removal Process/Modular Caustic Side Solvent Extraction Unit, Tank Closure Cesium Removal, and the Salt Waste Processing Facility (SWPF). Panelists with presentations: SRS Liquid Waste Program (James Folk); SRS Liquid Waste Progress (Mark Schmitz); Salt Waste Processing Facility (Frank Sheppard)

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DOE OSTI2017

Implementation of Glycolic Acid Flowsheet in the Defense Waste Processing Facility (DWPF) - 17348

The disposition of high level waste (HLW) into canistered waste forms began at Savannah River Site (SRS) in 1996. For its entire processing history, DWPF has added formic acid as a pretreatment step in producing acceptable melter feed from incoming waste sludges sent from the HLW Tank Farms. While this process has been used to successfully produce more than 4,000 radioactive canisters, the addition of formic acid and its tendency to catalytically decompose in the presence of trace quantities of noble metals adds substantial process complexity in order to safely accommodate hydrogen generation and pH shifting during batch operations. Research efforts have been in progress for several years to identify a chemical surrogate for formic acid which performs the essential process functions without the associated negative consequences. The final analytical work to demonstrate the effectiveness of glycolic acid in the DWPF is nearly complete and a project team has been created to safely integrate the new acid into the chemical process. The project team is currently executing an implementation schedule which addresses safety basis development, revision to operations procedures and training, facility modifications, chemical procurement, and updated process monitoring. Introduction of glycolic acid into DWPF is currently scheduled to take place in early 2017, beginning with a blended flowsheet of glycolic and formic acid and ultimately transitioning to glycolic-only processing. This paper provides an overview of planned glycolic acid flowsheet implementation in the DWPF. This paper discusses several issues related to implementing such a significant change to the chemical process including: 1) Process chemistry differences between the two flowsheets with quantitative comparison for several properties of interest; 2) Technical approach to hazards analysis and safety basis development strategy; 3) Status of project execution for major deliverables; and 4) Path forward for facility modifications related to legacy formic acid support equipment. (authors)

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DOE OSTI2024

Improving Efficiency of DWPF Operations via Automating Process Calculations and Vitrifying High-Curie Feed- 24529

The Liquid Waste Organization (LWO) at the Savannah River Site (SRS) uses a “Power As One®” motto to process and dispose of radioactive waste. The Defense Waste and Processing Facility (DWPF) treats the high-level waste through a process of vitrification. The DWPF receives three incoming waste streams that are added to the Sludge Receipt and Adjustment Tank (SRAT): sludge, Monosodium Titanate/Sludge Solids (MST/SS), and Strip Effluent (SE). The liquid waste is mixed with pre-fabricated frit and treated with high temperatures in the melter prior to being poured into stainless steel canisters. The liquid cools to form solid glass within these canisters that are suitable for long-term storage and disposal. Savannah River Mission Completion (SRMC), the SRS liquid waste contractor for the U.S. Department of Energy, has implemented several facility improvements to further enhance the operations to support the Liquid Waste Operation (LWO) mission. Two of these operational enhancements include implementing the electronic Material Tracking Program Calculator (eMTPC)software and increasing the DWPF canister heat rate limit.

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DOE OSTI2016

Interim glycol flowsheet reduction/oxidation (redox) model for the Defense Waste Processing Facility (DWPF)

Control of the REDuction/OXidation (REDOX) state of glasses containing high concentrations of transition metals, such as High Level Waste (HLW) glasses, is critical in order to eliminate processing difficulties caused by overly reduced or overly oxidized melts. Operation of a HLW melter at Fe +2 /ΣFe ratios of between 0.09 and 0.33, a range which is not overly oxidizing or overly reducing, helps retain radionuclides in the melt, i.e. long-lived radioactive 99 Tc species in the less volatile reduced Tc 4+ state, 104 Ru in the melt as reduced Ru +4 state as insoluble RuO 2 , and hazardous volatile Cr 6+ in the less soluble and less volatile Cr +3 state in the glass. The melter REDOX control balances the oxidants and reductants from the feed and from processing additives such as antifoam. Currently, the Defense Waste Processing Facility (DWPF) is running a formic acid-nitric acid (FN) flowsheet where formic acid is the main reductant and nitric acid is the main oxidant. During decomposition formate and formic acid releases H 2 gas which requires close control of the melter vapor space flammability. A switch to a nitric acid-glycolic acid (GN) flowsheet is desired as the glycolic acid flowsheet releases considerably less H 2 gas upon decomposition. This would greatly simplify DWPF processing. Development of an EE term for glycolic acid in the GN flowsheet is documented in this study.

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DOE OSTI2017

A New Group Crystallization Approach in Used Nuclear Fuel Recycling: A Look at Fission Product Behaviors

Nuclear energy has the potential to play a large role in meeting future needs of the U.S. sustainable energy strategy, as a clean, carbon-free power source. However, to take advantage of the inherent benefits, many hurdles must be overcome in order for nuclear power to achieve these expectations. One of the largest challenges faced, is the management of used nuclear fuel (UNF). Currently, no strategy has been identified to deal with the growing UNF problem, and so the UNF remains stored on site at nuclear power plants either in cooling pools or in dry cask storage. The amount of stored UNF continues to grow, and with it, so does the concern surrounding its long-term radiotoxicity. This material could be reprocessed to recover the actinides, maximizing the energy potential of the UNF and minimizing the waste generation. Not only do U and Pu, the major energy producing actinides, need to be recovered, but recovering Np and Am would make waste generated from reprocessing significantly safer for long-term storage in a geological repository. There are a number of methods currently in use to reprocess UNF; throughout Europe, the Purex process is widely implemented, while the Japanese have developed the Next process. The Next process removes a bulk of the U by crystallization prior to a Purex-like separation scheme. In the Next process, it is undesirable to have any other species removed from solution in the crystallization step, but rather, the goal is to obtain a pure uranyl nitrate hexahydrate (UNH) product. Alternatively, a group separation of U-Am by a crystallization process would provide an elegant solution. Such a process has become attainable with the renewed interest in accessing Am(VI) in nitric acid solutions. In order to synthesize and stabilize Am(VI) which has only been demonstrated on a limited basis, a highly oxidizing system is required. By oxidizing Np, Pu, and Am to their hexavalent state, they can be co-crystallized with UNH, and thereby removed from solution. Recent results have demonstrated that all the hexavalent actinide (An) species are removed from solution in near proportion, by a simple reduction of temperature in 6-8 M HNO{sub 3}.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

Development and Implementation of a Scaled Saltstone Facility at Savannah River National Laboratory - 13346

The Savannah River National Laboratory (SRNL) has supported the Saltstone Production Facility (SPF) since its conception. However, bench scaled tests have not always provided process or performance data related to the mixing, transfer, and other operations utilized in the SPF. A need was identified to better understand the SPF processes and to have the capabilities at SRNL to simulate the SPF unit operations to support an active low-level radioactive waste (LLW) processing facility. At the SPF, the dry premix is weighed, mixed and transferred to the Readco '10-inch' continuous mixer where it is mixed with the LLW salt solution from the Salt Feed Tank (SFT) to produce fresh Saltstone slurry. The slurry is discharged from the mixer into a hopper. The hopper feeds the grout pump that transfers the slurry through at least 457.2 meters of piping and discharges it into the Saltstone Disposal Units (SDU) for permanent disposal. In conjunction with testing individual SPF processes over several years, SRNL has designed and fabricated a scaled Saltstone Facility. Scaling of the system is primarily based on the volume capacity of the mixer and maintaining the same shear rate and total shear at the wall of the transfer line. At present, SRNL is utilizing the modular capabilities of the scaled Saltstone Facility to investigate the erosion issues related to the augers and paddles inside the SPF mixer. Full implementation of the scaled Saltstone Facility is still ongoing, but it is proving to be a valuable resource for testing alternate Saltstone formulations, cleaning sequences, the effect of pumping Saltstone to farther SDU's, optimization of the SPF mixer, and other operational variables before they are implemented in the SPF. (authors)

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DOE OSTI2019

Engineering Controls for the Processing and Repackaging of Gloveboxes Contaminated with Plutonium - 19629

Gloveboxes used for processing plutonium (Pu)-238 at the US DOE Mound site in Miamisburg, Ohio, in support of the United States space mission were packaged in non-standard waste boxes and sent to the INL Site for temporary disposal. In addition to the gloveboxes, an assortment of smaller items, all with varying levels of Pu-238 contamination, were found in the non-standard waste boxes. The characteristics of Pu-238 make it especially difficult to contain and control contamination; alpha recoil phenomena and nanometer particle sizes lend to the contamination exhibiting near gas like characteristics. The Idaho Cleanup Project Core contractor, Fluor Idaho, LLC, engineered and deployed a waste processing containment with specialized ventilation and processes to characterize and disposition this waste. This paper summarizes the successes and lessons learned from processing Pu-238-contaminated waste. (authors)

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DOE OSTI2018

Treatment of Problematic Reactive Metal Wastes Using the GeoMelt{sup R} In-Container Vitrification (ICV{sup TM}) Process - 18639

Decommissioning of sodium cooled reactors and fast reactor technologies have generated a number of reactive metal waste configurations that are problematic to treat and typically lack cost effective treatment methods and disposition options. As a result, Veolia Nuclear Solutions, under contract with Idaho National Laboratory (owned by the U.S. Department of Energy and managed and operated by Battelle Energy Alliance, LLC) demonstrated its GeoMelt{sup R} In-Container Vitrification (ICV){sup TM} technology to safely convert sodium metal to a non-reactive vitrified oxide form. The demonstration project, supported by glass formulation and crucible testing, consisted of a series of ICV{sup TM} melts that processed elemental sodium into stable non-reactive glass. Battelle Energy Alliance, LLC is evaluating the GeoMelt{sup R} technology as a means to safely and reliably convert radioactive reactive metal residues that contaminate components from sodium cooled reactors into waste forms that comply with existing disposition pathways. Reactive metal wastes at INL require treatment in order to remove the Resource Conservation and Recovery Act (RCRA) reactivity and ignitability characteristics in order to comply with land disposal restrictions. GeoMelt{sup R}, which is an alternative to other potential treatment approaches, provides a robust approach that chemically converts the reactive metals to an inert oxide while also immobilizing radionuclides in a vitrified waste form with durability equal to or better than vitrified nuclear fuel reprocessing wastes (very robust and inert waste forms). Most other treatment approaches generate hydrogen gas which is problematic. In 2016, Veolia Nuclear Solutions demonstrated the effectiveness of the GeoMelt{sup R} ICV{sup TM} process in deactivating reactive sodium metal. Crucible, bench-scale, and engineering-scale demonstrations were conducted on several surrogate waste configurations with various ratios of sodium metal and glassformers. Each ratio and configuration demonstrated complete deactivation of the surrogate sodium metal. Follow-on work in 2017 demonstrated the deactivation of reactive sodium by GeoMelt{sup R} ICV{sup TM} at a higher waste loading relative to previously demonstrated work performed in 2016. The higher waste loading optimized glass chemistry while enhancing the economical full-scale treatment of reactive metals. The GeoMelt{sup R} technology is a proven radioactive waste treatment technology capable of immobilizing radioactive wastes, including bulk rubble such as drums and other steel vessels usually without pretreatment. Utilizing the GeoMelt{sup R} technology to treat reactive metals eliminates pretreatment steps resulting from having to separate the reactive metal from steel containers or jackets as GeoMelt{sup R} can easily operate at temperatures sufficient to melt the steel and expose the reactive metal for treatment. Eliminating handling steps of reactive metals is a significant safety advantage since reactive metals are pyrophoric. The results of this two-year demonstration program and future planned activities are presented in the paper. (authors)

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DOE OSTI2026

Induction Melter Processing Alternatives for High Level Radioactive Wastes – 26198

This work investigates the potential to vitrify nuclear fuel directly, as well as the vitrification potential associated with experimental dissolver solutions. Greater understanding of the exothermicity is needed to quantify processing risk, especially with respect to potential phase changes and associated explosion hazards present in some systems. Thermal analysis was carried out on various simulants to elucidate the exothermic reaction potential from solid metal dissolution in glass and from the drying of alternative process dissolver solutions. Results from experimental testing of simulants to demonstrate vitrification potential and compatibility indicate that alternative dissolver flowsheets suppress the heat released during processing and that common silicate- and phosphate- based glass systems are potential candidates for direct vitrification. Direct vitrification (conversion) of fuel simulants was assessed using laboratory scale glass melts to obtain qualitative information on dissolution rates and waste loadings. Initial tests were successful to dissolve and incorporate metal directly into glass, although the kinetics and limits of dissolution and incorporation into glass are not fully understood.

Amoroso, Jake [Savannah River National Laboratory ↗

DOE OSTI2019

Treatment of Uranium Catalyst Waste for High Volume Reduction of the Final Waste to be Disposed - 19023

A large volume of problematic radioactive waste of spent uranium catalyst, approximately 7,000 drums was generated by a chemical company in Korea. The spent uranium catalyst has since been stored onsite awaiting a management strategy that is to minimize the volume/cost requirements for final disposal of the uranium-bearing waste. This work developed for the first time a process to greatly reduce the waste volume to be disposed while the secondary wastes generated from the process being released to the environment are controlled to meet the discharge criteria in Korea. (authors)

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DOE OSTI2020

Panel Session 110B: US DOE High-Level Radioactive Waste (HLW) Interpretation

This panel focused on the US DOE High-Level radioactive Waste (HLW) definition interpretation. Discussion integrated topics such as the issuance of the HLW interpretation Federal Registration Notice, the status of the draft Environmental Assessment that analyzes the treatment and commercial disposal of up to 10,000-gallons of SRS Defense Waste Processing Facility recycle wastewater in South Carolina, and the path forward. Panelists with presentations: Status on the Environmental Assessment for the Commercial Disposal of Defense Waste Processing Facility Recycle Wastewater from SRS (Theresa Kliczewski); Enhancing Stakeholder Engagement on the HLW interpretation (Kara Colton, Rick McLeod); Technical Aspects of Potential Disposal of SRS DWPF Recycle Wastewater (Kent Rosenberger)

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DOE OSTI2024

Application of Process Chemical Modeling to Optimize Radioactive Waste Disposal at the Savannah River Site - 24242

The Technical Optimization Model (TOM) is used by Savannah River Mission Completion (SRMC) to carry out facility-wide material balance and validate chemistrydependent processes for the purpose of their System Plan.The TOM simulates material movement and chemical reactions at the Tank Farm (TF), Salt Waste Processing Facility (SWPF) and Defense Waste Processing Facility (DWPF).

Georgiou, Andreas↗

DOE OSTI2016

Development of a Holistic Waste Management Flow Sheet for Nuclear New Build in China - 16634

Nuclear new build is continuing to experience rapid expansion globally with China leading the pace of installed and planned new generating capacity. While new generation LWR's can be largely considered to be standardized the host countries may have legislative and regulatory drivers that can impact their environmental performance and design of waste treatment systems. Of particular issue is an increasing divergence in the technical approach to liquid waste processing in order to accommodate local environmental impacts and supply chain issues. As a global provider of LWP technology to nuclear new build EnergySolutions is seeking to standardize world class environmental performance in liquid and solid waste processing using a standardized suite of technology modules tailored for host country's needs. This approach is consistent with standardization efforts seen historically in the USA and supported by past EPRI programs. Details are provided of EnergySolutuions participation in the provision of world class liquid and solid technology to global nuclear new build. Including details of how innovative technology can be tailored to meet host country needs within the broader concept of standardization for liquid waste processing and management of resultant solid secondary waste. (authors)

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DOE OSTI2016

Treatment Study Plan for Nitrate Salt Waste Remediation Revision 1.0

The two stabilization treatment methods that are to be examined for their effectiveness in the treatment of both the unremediated and remediated nitrate salt wastes include (1) the addition of zeolite and (2) cementation. Zeolite addition is proposed based on the results of several studies and analyses that specifically examined the effectiveness of this process for deactivating nitrate salts. Cementation is also being assessed because of its prevalence as an immobilization method used for similar wastes at numerous facilities around the DOE complex, including at Los Alamos. The results of this Treatment Study Plan will be used to provide the basis for a Resource Conservation and Recovery Act (RCRA) permit modification request of the LANL Hazardous Waste Facility Permit for approval by the New Mexico Environment Department-Hazardous Waste Bureau (NMED-HWB) of the proposed treatment process and the associated facilities.

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