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Stabilization of Radionuclides and Metals into Spinels Using THOR{sup R} Fluidized Bed Steam Reforming - 12039

Studsvik, Inc.'s Thermal Organic Reduction (THOR) Process has been operating for over 12 years in the treatment of radioactive liquid and solid wastes from commercial Nuclear Power Plants in the United States. A second THOR plant is starting operations in early 2012 for the treatment of radioactive wastes from US nuclear power research at US DOE's Idaho National Laboratory. THOR has long been recognized for its unique ability to substantially reduce the volume of radioactive wastes - via the removal of water and destruction of organics, nitrates/nitrites, etc.- and produce a dry, stable solid product while retaining extremely high amounts (>99.99%) of the incoming wastes' radioactivity. Moreover, the THOR process has demonstrated the ability to produce radioactive waste forms with leach-resistance as good as or better than vitrified glass. An additional facet of the technology which, to this point, has been marginalized is the production spinels that are essential for capturing volatile, leachable, and large radioactive and hazardous inorganic species. These spinels form from the wide variety of inorganics in the incoming waste feed and proprietary additives used in processing. (authors)

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Treatment of Liquid Radioactive Waste with High Salt Content by Colloidal Adsorbents - 13274

Treatment processes have been fully developed for most of the liquid radioactive wastes generated during the operation of nuclear power plants. However, a process for radioactive liquid waste with high salt content, such as waste seawater generated from the unexpected accident at nuclear power station, has not been studied extensively. In this study, the adsorption efficiencies of cesium (Cs) and strontium (Sr) in radioactive liquid waste with high salt content were investigated using several types of zeolite with different particle sizes. Synthesized and commercial zeolites were used for the treatment of simulated seawater containing Cs and Sr, and the reaction kinetics and adsorption capacities of colloidal zeolites were compared with those of bulk zeolites. The experimental results demonstrated that the colloidal adsorbents showed fast adsorption kinetic and high binding capacity for Cs and Sr. Also, the colloidal zeolites could be successfully applied to the static adsorption condition, therefore, an economical benefit might be expected in an actual processes where stirring is not achievable. (authors)

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Arvia organics destruction: an innovative treatment approach for dioxins, furans and other problematic (Orphan) organic wastes - 15380

The safe and efficient disposal of radionuclide-contaminated organic wastes is a key component in the life-cycle of a nuclear power plant. Technologies are available for the processing of some of these wastes although there are a number of waste streams for which the disposition path is not clear or does not exist at all. For example, waste streams such as dioxin, furan and PCB-laden oils pose an issue for the US Dept. of Energy complex. These wastes present a challenge as they cannot be readily treated using existing treatment technologies because of legislative and material constraints such as gaseous phase radioactivity discharge limits, waste elemental content, by-product generation and waste resistance to thermal decomposition. Alternate methods of organics destruction can circumvent some of these issues by using different mechanisms of organics oxidation. One such innovative method, known as the Arvia organics destruction process, combines adsorption with electrochemical combustion and offers such an alternative. This technology, based on a patented material, Nyex{sup TM} that is able to act as both an adsorbent and as an electrode, allows the electrochemical combustion of adsorbed organics to occur on the material's surface. The result is the destruction of organics whilst the bulk of inorganic and radionuclide components remain in the aqueous phase. These components can subsequently be treated using existing waste processing technologies. This paper outlines recent work conducted on demonstrating the Arvia process against the destruction of dioxin, furan and PCB-laden wastes that demonstrates that the technology successfully removes these target species to below discharge standards in the aqueous phase. Based on these successful trials a test unit has been designed to demonstrate the process on actual radioactive wastes at the DOE Oak Ridge facility. The successful treatment of F027-coded waste has also been demonstrated and will be reported. The above work showcases the development, scaling and application of an innovative technology, first presented to the industry at the Waste Management Symposia in 2012, as another option for the treatment of challenging wastes with expensive, complex or no current disposal routes. The Arvia process adds to the portfolio of available treatment technologies and will help facilitate the decommissioning and waste disposal responsibilities of the nuclear industry. (authors)

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Hanford Tank Waste Treatment and Immobilization Plant (WTP) Waste Feed Qualification Program Development Approach - 13114

The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is a nuclear waste treatment facility being designed and constructed for the U.S. Department of Energy by Bechtel National, Inc. and subcontractor URS Corporation (under contract DE-AC27-01RV14136 [1]) to process and vitrify radioactive waste that is currently stored in underground tanks at the Hanford Site. A wide range of planning is in progress to prepare for safe start-up, commissioning, and operation. The waste feed qualification program is being developed to protect the WTP design, safety basis, and technical basis by assuring acceptance requirements can be met before the transfer of waste. The WTP Project has partnered with Savannah River National Laboratory to develop the waste feed qualification program. The results of waste feed qualification activities will be implemented using a batch processing methodology, and will establish an acceptable range of operator controllable parameters needed to treat the staged waste. Waste feed qualification program development is being implemented in three separate phases. Phase 1 required identification of analytical methods and gaps. This activity has been completed, and provides the foundation for a technically defensible approach for waste feed qualification. Phase 2 of the program development is in progress. The activities in this phase include the closure of analytical methodology gaps identified during Phase 1, design and fabrication of laboratory-scale test apparatus, and determination of the waste feed qualification sample volume. Phase 3 will demonstrate waste feed qualification testing in support of Cold Commissioning. (authors)

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A WM Facility for the Processing of Radioactive Materials Arising from the Decommissioning Activities of the Belgian Nuclear Power Plant of Doel - 16042

Doel Nuclear Power Plant is located in the northern part of Belgium in the harbour of Antwerp on the Scheldt River. On this 80-hectare site Electrabel, the Belgian utility, operates four PWRs with a total generation capacity of 2911 MWe. Those nuclear power units were commissioned in 1975 (Doel 1 and 2 - twin units), in 1982 (Doel 3) and in 1985 (Doel 4). In light of the Belgian governments decision to definitively shutdown the oldest units of the site (Doel 1 and 2 - 433 MWe each) in 2015 at the end of their 40-year design lifetime, Electrabel launched a D and D program in partnership with Tractebel Engineering in July 2012. During the D and D preparation phase prior to the final shutdown, studies performed by Tractebel Engineering showed the added benefit of a purpose-built Waste Management Facility (WMF) specifically equipped to process contaminated and activated components generated by the in-situ dismantling activities. The WMF is a key infrastructure contributing to a large extent to the Waste Minimization Policy within the global D and D program. The present paper aims at presenting its strategic, logistic, infra-structural, and design principles. At the end of 2014, the Belgian government and Electrabel entered into discussions to envisage a 10-year lifetime extension of Doel 1 and 2 units. Therefore, Electrabel decided to put the D and D program on hold. Nevertheless, the preparatory studies for the future D and D activities on Doel site are set on solid assumptions and pave the way towards a safe, cost-effective and waste minimizing decommissioning process. (authors)

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Phyto-screening for Volatile Organic Compounds in a Clay-based Glacial Till at the Niagara Falls Storage Site, New York - 17372

The Niagara Falls Storage Site (NFSS) and associated vicinity properties (VPs) occupy approximately 607 hectares (1,500 acres) of the original 3,035-hectare (7,500-acre) Lake Ontario Ordnance Works (LOOW). The NFSS was used to store radioactive materials, residues and waste from uranium ore processing between 1944 and 1952. The waste-handling practices produced an array of groundwater impacts throughout the site. Although the groundwater at the NFSS is naturally high in total dissolved solids that preclude untreated consumption, the groundwater contaminant levels in some areas of the site are very high and may pose risk to a Construction Worker receptor, as well as limit the options for future site re-use. Volatile organic compound (VOC) impacted soil in the northern portion of the site has produced a groundwater plume with significant concentrations of tetrachloroethene (PCE) and associated daughter compounds. Active groundwater remediation is impractical due to the clayey soils, as evident from adjacent property owners who have unsuccessfully attempted active remediation of their VOC plumes. Consequently, the Buffalo District may evaluate excavation options, with secondary phyto-remediative approaches as passive alternatives to lessen the groundwater impacts during the project's life cycle. A preliminary field-study in the VOC plume area sampled the interior of a Black Locust (Robinia pseudoacacia) tree using a Haglof tree corer. Detectable levels of tetrachloroethene (PCE), trichloroethene (TCE), and cis-1,2,-dichloroethene (cis-DCE) were found in the woody interior of the tree. Subsequent sampling and analyses were performed on proximal trees of varying species in and around the currently delineated VOC plume. The results indicate that poplar and native cottonwood variants preferentially contain the VOCs, whereas collocated hardwood trees produced non-detectable results. Additional sampling of Cottonwoods in the VOC-impacted area promoted further delineations of the impacts. The tree-sampling data can be considered a broad surrogate for soil sampling to optimally delineate the VOC plume without performing additional invasive sampling of the subsurface (e.g., drilling). The results also will be used to determine which native tree species preferentially uptake the VOCs from soil and groundwater. These species, and possibly other hybrid species known for VOC uptake and transpiration, would be propagated to promote interim phyto-extraction of the VOCs during the maintenance period between decision-document completion and likely remedial action under the Formerly Utilized Sites Remedial Action Program (FUSRAP). (authors)

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

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

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Testing of the Defense Waste Processing Facility Cold Chemical Dissolution Method in Sludge Batch 9 Qualification

For each sludge batch that is processed in the Defense Waste Processing Facility (DWPF), the Savannah River National Laboratory (SRNL) tests the applicability of the digestion methods used by the DWPF Laboratory for elemental analysis of Sludge Receipt and Adjustment Tank (SRAT) Receipt samples and SRAT Product process control samples. DWPF SRAT samples are typically dissolved using a method referred to as the DWPF Cold Chemical or Cold Chem Method (CC), (see DWPF Procedure SW4- 15.201). Testing indicates that the CC method produced mixed results. The CC method did not result in complete dissolution of either the SRAT Receipt or SRAT Product with some fine, dark solids remaining. However, elemental analyses did not reveal extreme biases for the major elements in the sludge when compared with analyses obtained following dissolution by hot aqua regia (AR) or sodium peroxide fusion (PF) methods. The CC elemental analyses agreed with the AR and PF methods well enough that it should be adequate for routine process control analyses in the DWPF after much more extensive side-by-side tests of the CC method and the PF method are performed on the first 10 SRAT cycles of the Sludge Batch 9 (SB9) campaign. The DWPF Laboratory should continue with their plans for further tests of the CC method during these 10 SRAT cycles.

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Time dependent zero valent iron oxidation and the reductive removal of pertechnetate at variable pH

Elemental iron Fe 0 is a promising reductant for removal of radioactive technetium-99 (Tc) from complex aqueous waste streams that contain sulfate, halides, and other inorganic anions generated during processing of legacy radioactive waste. The impact of sulfate on the kinetics of oxidation and reduction capacity of Fe 0 in the presence of Tc has not been examined. We investigated the oxidative transformation of Fe 0 and reductive removal of TcO 4 - in 0.1 M Na 2 SO 4 as a function of initial pH (i.e., pH i 4, 7, and 10) under aerobic conditions up to 30 days. Tc reduction was the fastest at pH i 7 and slowest at pHi 10 (Tc reduction rate pHi 7 > 4 > 10). Aqueous fraction of Tc was measured at 0.4% at pH i 7 within 6 h, whereas ≥ 97% of Tc was removed from solutions at pH i of 4 and 10 within 24 h. Solid phase characterization showed that magnetite was the only oxidized crystalline phase for the first 6 h regardless of initial pH. Lepidocrocite was the most abundant oxidized product for pH i 10 after 5 days, but was not observed at pH of 4 or 7.

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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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Development of low-pH cements for immobilisation of intermediate level radioactive waste: achievements and challenges

Although cementation is a widely recognized solidification/ stabilization process for immobilisation of Intermediate Level Radioactive Waste (ILRW), the low resistance to hyper-alkaline pore waters compromises the effectiveness of the process when Portland Cement (PC) is employed. Moreover the manufacture of PC is responsible for significant CO{sub 2} emissions. In this context, low pH cements are environmentally more suitable and have emerged as a potential alternative for obtaining secure waste forms. This paper summarises the achievements on development of low-pH cements and the challenges of using these new materials for the ILRW immobilisation. The performance of waste forms is also discussed in terms of radionuclides release. Reactive magnesium oxide and magnesium phosphate cements are emphasised as they feature important advantages such as consumption of available constituents for controlling acid-base reactions, reduced permeability and higher density. Additionally, in order to identify new opportunities for study, the long-term modelling approach is also briefly discussed. (authors)

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The Cires facility in Morvilliers: from a VLL waste disposal facility to the development of industrial activities - 15252

The disposal facility of CIRES for very low level wastes was designed to provide safe and cost effective services to waste generators, in particular for decommissioning works. The facility could significantly increase waste annual deliveries to meet waste generators' needs and accommodate large disused components as PWR steam generators after decontamination. However an anticipated saturation of its capacity is foreseen as needs for disposal has doubled since its start-up. The geometry of the cells was optimized and the technical capacity of the repository is presently by 40% above its licensed capacity. Saturation has been delayed; however a new facility will be necessary by 2030 and other waste management options are to be considered as recycling of metallic wastes or crushed concrete rubble. CIRES facility also became Andra's industrial platform for the management of institutional wastes. A grouping and a storage facility have been built and a new processing facility should start up in 2016. (authors)

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Simulation of DWPF Chemical Process Cell Using a Reaction Calorimeter

The Defense Waste Processing Facility (DWPF) converts liquid radioactive waste into glass using a Sludge Receipt Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) in addition to other processes. The SRAT and SME processes can be simulated on a small-scale using the Mettler-Toledo Reaction Calorimeter (RC1). This equipment is able to automate the execution and monitoring of experiments involving nonradioactive chemical simulants.

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Spectroscopic Online Monitoring for Process Control and Safeguarding of Radiochemical Fuel Reprocessing Streams - 13553

There is a renewed interest worldwide to promote the use of nuclear power and close the nuclear fuel cycle. The long term successful use of nuclear power is critically dependent upon adequate and safe processing and disposition of the used nuclear fuel. Liquid-liquid extraction is a separation technique commonly employed for the processing of the dissolved spent nuclear fuel. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments such as high radiation fields and aggressive chemical matrices. This paper discusses application of absorption and vibrational spectroscopic techniques supplemented by physicochemical measurements for radiochemical process monitoring. In this context, our team experimentally assessed the potential of Raman and spectrophotometric techniques for on-line real-time monitoring of the U(VI)/nitrate ion/nitric acid and Pu(IV)/Np(V)/Nd(III), respectively, in solutions relevant to spent fuel reprocessing. Both techniques demonstrated robust performance in the repetitive batch measurements of each analyte in a wide concentration range using simulant and commercial dissolved spent fuel solutions. Static spectroscopic measurements served as training sets for the multivariate data analysis to obtain partial least squares predictive models, which were validated using on-line centrifugal contactor extraction tests. Satisfactory prediction of the analytes concentrations in these preliminary experiments warrants further development of the spectroscopy-based methods for radiochemical safeguards and process control. (authors)

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French Innovative Thermal Treatment Processes for The Management Radioactive Organic Liquid Waste - 19017

The nuclear industry as well as hospitals or some research laboratories produce Radioactive Organic Liquid Waste (ROLW) such as scintillation cocktails for medical imaging, labeled molecules for medical research or liquids from former activities. In order to be compatible with disposal facilities requirements, this waste must be inerted and stabilized in solid matrix. The route generally used is incineration in the Centraco facilities at the Marcoule site (Gard, France), ashes produced being cemented downstream. However, liquids whose chemical or radiological characteristics are not compatible with the incinerator operating rules (high level of {sup 14}C or halogen content, corrosive gas production etc.) cannot follow this path. The purpose of the MILOR project is then to propose alternative paths to treat all kinds of ROLW. Initiated in September 2017, this project has the technical objective of developing, over a four-year period, two pilot-scale mineralization processes. Two variants of complementary plasma technologies are evaluated: aerial plasma with the 'IDOHL' process and submerged plasma with the 'ELIPSE' process. Compared with other methods, thermal plasmas have unique characteristics such as rapid decomposition with high throughput, fast startup and shutdown, and high-energy delivery. The process called IDOHL consists of a process involving an induction plasma in which the liquids are directly introduced. It has been designed for organic liquids without mineral load and for treatment rates less than 0.5 l/h. This type of plasma allows the destruction of liquid organohalogens without any prior evaporation or treatment. Specific gas treatment systems have been developed to direct the gas produced to a dry or wet gas treatment system. One aim of the project is to subsequently install the process in a radioactive zone at Saclay DRF. The second called ELIPSE consists of a breakthrough process involving a submerged plasma at the heart of which organic liquids are also introduced. Current research has demonstrated that such a technique may enable the instantaneous and complete destruction of liquids with a wide variety of constituents, such as chlorine, fluorine, or phosphorus. The submersion solution offers many advantages: quenching and cleaning of combustion gases; filtering of the particles they contain; and cooling maintained for the entire process, which guarantees excellent corrosion control. An advantage of this type of design is that the gas treatment system can be reduced to a demister-condenser followed by a simple safety filter, thereby offering the additional advantage of an extremely compact treatment system. This design also allows the ELIPSE process to become by this way an embeddable process if required. The MILOR project, financed by the French State through investment program is a tripartite project bringing together Andra, in charge of radioactive waste management, the CEA in charge of leading scientific and technological researches to propose effective ways of treatment and the company A3i whose role is to ensure the subsequent industrial development. (authors)

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IMPACT OF URANIUM AND THORIUM ON HIGH TIO2 CONCENTRATION NUCLEAR WASTE GLASSES

This study focused on the potential impacts of the addition of Crystalline Silicotitanate (CST) and Monosodium Titanate (MST) from the Small Column Ion Exchange (SCIX) process on the Defense Waste Processing Facility (DWPF) glass waste form and the applicability of the DWPF process control models. MST from the Salt Waste Processing Facility (SWPF) is also considered in the study. The KT08-series of glasses was designed to evaluate any impacts of the inclusion of uranium and thorium in glasses containing the SCIX components. All but one of the study glasses were found to be amorphous by X-ray diffraction (XRD). One of the slowly cooled glasses contained a small amount of trevorite, which is typically found in DWPF-type glasses and had no practical impact on the durability of the glass. The measured Product Consistency Test (PCT) responses for the study glasses and the viscosities of the glasses were well predicted by the current DWPF models. No unexpected issues were encountered when uranium and thorium were added to the glasses with SCIX components.

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Defense Waste Processing Facility (DWPF) Viscosity Model: Revisions for Processing High TiO 2 Containing Glasses

Radioactive high-level waste (HLW) at the Savannah River Site (SRS) has successfully been vitrified into borosilicate glass in the Defense Waste Processing Facility (DWPF) since 1996. Vitrification requires stringent product/process (P/P) constraints since the glass cannot be reworked once it is poured into ten foot tall by two foot diameter canisters. A unique “feed forward” statistical process control (SPC) was developed for this control rather than statistical quality control (SQC). In SPC, the feed composition to the DWPF melter is controlled prior to vitrification. In SQC, the glass product would be sampled after it is vitrified. Individual glass property-composition models form the basis for the “feed forward” SPC. The models transform constraints on the melt and glass properties into constraints on the feed composition going to the melter in order to guarantee, at the 95% confidence level, that the feed will be processable and that the durability of the resulting waste form will be acceptable to a geologic repository. The DWPF SPC system is known as the Product Composition Control System (PCCS). The DWPF will soon be receiving wastes from the Salt Waste Processing Facility (SWPF) containing increased concentrations of TiO 2 , Na 2 O, and Cs 2 O . The SWPF is being built to pretreat the high-curie fraction of the salt waste to be removed from the HLW tanks in the F- and H-Area Tank Farms at the SRS. In order to process TiO 2 concentrations >2.0 wt% in the DWPF, new viscosity data were developed over the range of 1.90 to 6.09 wt% TiO 2 and evaluated against the 2005 viscosity model. An alternate viscosity model is also derived for potential future use, should the DWPF ever need to process other titanate-containing ion exchange materials. The ultimate limit on the amount of TiO 2 that can be accommodated from SWPF will be determined by the three PCCS models, the waste composition of a given sludge batch, the waste loading of the sludge batch, and the frit used for vitrification.

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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)

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