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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Salt Waste Processing Facility project lessons learned: Acquisition through construction - 15518

The Salt Waste Processing Facility (SWPF), located at the Savannah River Site (SRS) in Aiken, South Carolina, is a multi-billion dollar Liquid Radioactive Waste (LRW) processing facility that represents the keystone for the U.S. Department of Energy (DOE)-EM clean-up mission at SRS. The SWPF facilitates DOE-EM clean-up at SRS by decontaminating the salt waste, which is the dominant volumetric component in the LRW tanks, thereby allowing the cleaned salt waste to be removed from the tanks and stabilized into an approved grout-based final waste form. The small volume of removed radionuclides is sent to the Defense Waste Processing Facility (DWPF) for vitrification. Emptied LRW storage tanks are then grouted into an approved configuration for final closure. The SWPF project is currently 100% complete with design, 80% complete with construction, and 10% complete with commissioning. To date there have been numerous lessons learned with regard to project acquisition, design, procurement, construction, and quality assurance (QA) approaches (both good practices and mistakes) that are considered to be of significant value in improving the state of major project execution within the DOE-EM clean-up program as well as to the wider DOE enterprise. An overview of the current SWPF project status is provided and key lessons learned in the aforementioned areas are presented relative to improving future DOE major project execution performance. Specific areas to be addressed include strategies to optimize integration of design and nuclear safety, constructability tools for design and construction synchronization, commercial grade dedication (CGD) approaches to mitigate atrophy in the nuclear supply chain, and QA perspectives to facilitate effective and efficient project execution. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Analysis of Monolith Cores from an Engineering Scale Demonstration of a Prospective Cast Stone Process

The primary disposition path of Low Activity Waste (LAW) at the DOE Hanford Site is vitrification. A cementitious waste form is one of the alternatives being considered for the supplemental immobilization of the LAW that will not be treated by the primary vitrification facility. Washington River Protection Solutions (WRPS) has been directed to generate and collect data on cementitious or pozzolanic waste forms such as Cast Stone. This report documents the coring and leach testing of monolithic samples cored from an engineering-scale demonstration (ES Demo) with non-radioactive simulants. The ES Demo was performed at SRNL in October of 2013 using the Scaled Continuous Processing Facility (SCPF) to fill an 8.5 ft. diameter x 3.25 ft. high container with simulated Cast Stone grout. The Cast Stone formulation was chosen from the previous screening tests. Legacy salt solution from previous Hanford salt waste testing was adjusted to correspond to the average LAW composition generated from the Hanford Tank Waste Operation Simulator (HTWOS). The dry blend materials, ordinary portland cement (OPC), Class F fly ash, and ground granulated blast furnace slag (GGBFS or BFS), were obtained from Lafarge North America in Pasco, WA. In 2014 core samples originally obtained approximately six months after filling the ES Demo were tested along with bench scale molded samples that were collected during the original pour. A latter set of core samples were obtained in late March of 2015, eighteen months after completion of the original ES Demo. Core samples were obtained using a 2” diameter x 11” long coring bit. The ES Demo was sampled in three different regions consisting of an outer ring, a middle ring and an inner core zone. Cores from these three lateral zones were further segregated into upper, middle and lower vertical segments. Monolithic core samples were tested using the Environmental Protection Agency (EPA) Method 1315, which is designed to provide mass transfer rates (release rates) of inorganic analytes contained in monolithic material under diffusion controlled release conditions as a function of leaching time. Compressive strength measurements and drying tests were also performed on the 2015 samples. Leachability indices reported are based on analyte concentrations determined from dissolution of the dried samples.

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

LDR Inorganic Spiked Simulants in Pozzolan-Substituted Cast Stone for Hanford Pretreated Low-Activity Tank Waste

Pretreated Low Activity Waste (PTW/LAW) at the Hanford Site, is currently projected for disposition through a cementitious waste form. The current standard formulation, Cast Stone, uses Class F fly ash, blast furnace slag, and Portland cement. However, the future availability of Class F fly ash is uncertain, as many coal-fired power plants—the source of this byproduct—are being decommissioned or converted to natural gas. This issue has prompted research into identifying and evaluating suitable alternative materials to replace fly ash in the cementitious waste forms used by the Department of Energy (DOE). The Savannah River National Laboratory (SRNL) successfully identified four natural pozzolans for the replacement of Class F fly ash in the current Savannah River Site (SRS) low-level waste (LLW) form, Saltstone.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Process Control Approach to Implement High Waste Loading Glass Formulations for Hanford Low Activity Waste Vitrification - 18435

Acceptable glass formulations for vitrification of Hanford low activity waste (LAW) at The Hanford Tank Waste Treatment and Immobilization Plant (WTP) must meet a variety of product quality, processability, and waste loading requirements. The glass formulation development work underpinning the WTP baseline LAW glasses was captured in a vitrification process control approach developed by the Vitreous State Laboratory (VSL) and Atkins Energy Federal EPC, Inc. that permits the calculation of the glass formulation required for each particular batch of LAW delivered to the LAW vitrification facility. Since that time, we have developed a wide range of glass formulations that achieve considerably higher waste loadings and which therefore are more economically attractive in terms of reduced amounts of glass produced and decreased time for treatment. That work was done in support of the Department of Energy's Office of River Protection program that is examining options to optimize the LAW facility and LAW glass waste form. Implementation of these high waste loading formulations at the WTP requires the development of a new LAW vitrification process control algorithm, which is the subject of this paper. Waste loadings in LAW glasses are typically limited by either the alkali loading for LAW with low sulfate content or the SO{sub 3} loading for LAW with high sulfate content. In the former case, waste loadings are limited by refractory corrosion and glass corrosion on the vapor hydration test (VHT) and product consistency test (PCT), all of which tend to increase with higher alkali concentrations. In the latter case, waste loadings are limited by the formation of a molten sulfate phase in the melter, which has deleterious effects. The LAW vitrification process control strategy must mitigate these effects while optimizing waste loadings. Given the composition of any LAW stream, the new LAW vitrification process control algorithm permits the calculation of a suitable high waste loading glass composition and the required glass forming additives such that all processing, waste loading, and product quality requirements are met. This algorithm was developed from data for a large number of glasses that were formulated, and characterized with respect to a variety of properties affecting processability and product quality; many of these formulations were also subjected to melter testing to verify acceptable processing characteristics, processing rates, and the absence of secondary sulfate phase formation. This paper describes the development of the present algorithm, comparisons to the present lower waste loading baseline WTP process control approach, and plans for further development and testing. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Separating and Stabilizing Phosphate from High-Level Radioactive Waste: Process Development and Spectroscopic Monitoring

Removing phosphate from alkaline high-level waste sludges at the Department of Energy's Hanford Site in Washington State is necessary to increase the waste loading in the borosilicate glass waste form that will be used to immobilize the highly radioactive fraction of these wastes. We are developing a process which first leaches phosphate from the high-level waste solids with aqueous sodium hydroxide, and then isolates the phosphate by precipitation with calcium oxide. Tests with actual tank waste confirmed that this process is an effective method of phosphate removal from the sludge and offers an additional option for managing the phosphorus in the Hanford tank waste solids. The presence of vibrationally active species, such as nitrate and phosphate ions, in the tank waste processing streams makes the phosphate removal process an ideal candidate for monitoring by Raman or infrared spectroscopic means. As a proof-of-principle demonstration, Raman and Fourier transform infrared (FTIR) spectra were acquired for all phases during a test of the process with actual tank waste. Quantitative determination of phosphate, nitrate, and sulfate in the liquid phases was achieved by Raman spectroscopy, demonstrating the applicability of Raman spectroscopy for the monitoring of these species in the tank waste process streams.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

NEAMS Update: Quarterly report for January - March 2012

Quarterly highlights are: (1) The integration of Denovo and AMP was demonstrated in an AMP simulation of the thermo-mechanics of a complete fuel assembly; (2) Bison was enhanced with a mechanistic fuel cracking model; (3) Mechanistic algorithms were incorporated into various lower-length-scale models to represent fission gases and dislocations in UO2 fuels; (4) Marmot was improved to allow faster testing of mesoscale models using larger problem domains; (5) Component models of reactor piping were developed for use in Relap-7; (6) The mesh generator of Proteus was updated to accept a mesh specification from Moose and equations were formulated for the intermediate-fidelity Proteus-2D1D module; (7) A new pressure solver was implemented in Nek5000 and demonstrated to work 2.5 times faster than the previous solver; (8) Work continued on volume-holdup models for two fuel reprocessing operations: voloxidation and dissolution; (9) Progress was made on a pyroprocessing model and the characterization of pyroprocessing emission signatures; (10) A new 1D groundwater waste transport code was delivered to the used fuel disposition (UFD) campaign; (11) Efforts on waste form modeling included empirical simulation of sodium-borosilicate glass compositions; (12) The Waste team developed three prototypes for modeling hydride reorientation in fuel cladding during very long-term fuel storage; (13) A benchmark demonstration problem (fission gas bubble growth) was modeled to evaluate the capabilities of different meso-scale numerical methods; (14) Work continued on a hierarchical up-scaling framework to model structural materials by directly coupling dislocation dynamics and crystal plasticity; (15) New 'importance sampling' methods were developed and demonstrated to reduce the computational cost of rare-event inference; (16) The survey and evaluation of existing data and knowledge bases was updated for NE-KAMS; (17) The NEAMS Early User Program was launched; (18) The Nuclear Regulatory Commission (NRC) Office of Regulatory Research was introduced to the NEAMS program; (19) The NEAMS overall software quality assurance plan (SQAP) was revised to version 1.5; and (20) Work continued on NiCE and its plug-ins and other utilities, such as Cubit and VisIt.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Enhanced Corrosion of Magnesium as a Treatment for Legacy SNF - 19369

The ability of high concentrations of MgCl{sub 2} to induce activity-controlled dissolution of magnesium has been investigated in the context of providing a means of stabilising reactive metal fuel wastes, as oxide/hydroxide corrosion products, to enhance the resilience of the waste form following disposal. Kinetic parameters were measured using a novel immersion testing method that utilise the gravimetric measurement of evolving hydrogen produced by the oxidation of the metal. This method demonstrates extremely high precision relative to a traditional volumetric approach and appears to be relatively robust in its accuracy despite its sensitivity to changing temperature. An effective activation energy of 10.2 kJmoL{sup -1} was measured for the reaction using the Arrhenius equation over a temperature range of 30-70 deg. C. This implies that the reaction remains controlled by the diffusion of reactants to the metal surface and fails to corroborate the effect previously observed by Casey and Bergeron. This may be explained by differences in initial conditions, or it could be that the system transitions into the previously observed state as it reaches equilibrium, a possibility that was not accounted for in this study as the experiments focussed on measuring the initial rate of reaction. Repeat testing and further study is warranted given that enhanced corrosion could offer a means to produce a self-immobilising waste form with minimal intrusion, offering both improved safety tolerances and cost savings for the decommissioning and waste disposal efforts ongoing at the Sellafield site. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Mercury Speciation in Savannah River Site High-Level Waste: Determination and Implications for Waste Handling and Disposal - 16178

As part of a larger program to address mercury issues at the Savannah River Site, a number of samples from across the salt processing flowsheet have been collected and analyzed for several mercury species. The results indicate the presence of organomercury species, predominately methylmercury cation. This result was not expected, and it represents a significant paradigm shift in the understanding of mercury chemistry in high-level tank waste. Concentrations of total and soluble mercury have been measured as high as 0.5 g/L with methylmercury concentrations of at least fifty percent of that value. The distribution of mercury species being observed have ramifications to waste processing strategies. In particular, the mercury behavior across the cesium solvent extraction process has been examined. Leachates, from toxicity testing of the cementitious waste form used for decontaminated salt waste at Savannah River, have been tested and shows methylmercury as the dominant extractable species. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Integration of National Laboratory and Low-Activity Waste Pretreatment System Technology Service Providers - 16435

The National Laboratories are a critical partner and provide expertise in numerous aspects of the successful execution of the Direct-Feed Low Activity Waste Program. The National Laboratories are maturing the technologies of the Low-Activity Waste Pretreatment System consistent with DOE Order 413.3B, 'Program and Project Management for the Acquisition of Capital Assets', expectations. The National Laboratories continue to mature waste forms, i.e., glass and secondary waste grout, for formulations and predictions of long-term performance as inputs to performance assessments. The working processes with the National Laboratories have been developed in procurements, communications, and reporting to support the necessary delivery-based technology support. The relationship continues to evolve from planning and technology development to support of ongoing operations and integration of multiple highly coordinated facilities. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Vitrification of High-Level Waste: how to Select the Best Suited Technology - 18117

Since more than 30 years now, a number of countries have selected and operated vitrification as the best industrial practice for safely and durably immobilizing the most active liquid waste. This now significant worldwide experience covers a wide range of waste compositions, from concentrated alkaline defense waste sludges with a variety of salts and chemicals (Na, Fe, Al, Cr, sulphate,... ) to relatively clear acidic HLLW from commercial, low salt, PUREX-type processes, waste that include dissolved cladding material (Al, Mg, stainless steel), waste that include high levels of noble metals and fines... The activity levels vary from 40- year old defense waste to relatively young waste from high burn-up commercial fuel. The current throughputs vary from about 5 kg/h of glass to about 90 kg/h of glass per melter. Even higher throughputs are currently being considered for Hanford facility for instance. In order to cope with the range of waste compositions, glass compositions have also evolved from low melting temperature borosilicate or phosphate glass to glass or glass ceramics with higher melting temperatures. More complex matrices are currently under study in a number of countries, to immobilize more complex waste and / or increase waste loadings. Each challenge is unique and, for each challenge, it is necessary to select the best combination of waste form and technology. For instance for very active solutions with high levels of noble metal particles that could settle, increase apparent glass viscosity, and impact electrical conductivity, one would select borosilicate glass in the mechanically-stirred hot wall induction melter with bottom pouring that has proven its efficiency in industrial conditions at La Hague. For high glass throughputs with low solids in a borosilicate melt (typically Defense waste), ceramic melters have proven their reliability and efficiency. For matrices that need higher temperatures, or with very high corrosiveness, Cold-Crucible Induction Melters are now considered, with one first industrial implementation at La Hague to produce borosilicate glass and a glass-ceramic. All these technologies have been improved over the years to increase throughput, waste loading, melting temperature, reliability... Other types of waste are now being studied worldwide, and for each, it is necessary to carefully select the adapted waste form and technology. The technologies studied will range from traditional technologies (such as hot-wall induction melter technology and liquid-fed ceramic melter technologies) to the more recent advanced technologies (such as Cold Crucible Induction Melter or Advanced Ceramic melters). In addition, the field of vitrification is extending to other types of waste; the current advances in this field will also be described. (authors)

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

Optimization of the Post-Operational Phase on Two Belgian Multi-Unit Nuclear Power Plants: the Case of the Non-Fissile Irradiated Core Items - 20156

The current legal framework in Belgium foresees the progressive phase out of nuclear power between October 2022 (Doel 3) and December 2025 (Doel 2). Upon its definitive shutdown, each unit of the Tihange and Doel sites will enter a Post-Operational Phase (POP) and be prepared for its Decontamination and Decommissioning (D and D). Prior to obtaining the D and D license, the Operator Electrabel is legally required to remove any non-fissile irradiated core items stored in the deactivation pools. The non-fissile irradiated core items consist essentially of control rods, poison rods and source thimbles as well as thimble plugs and foreign materials irradiated during operation: - Their significant content in highly radiant radionuclides (up to 6 TBq of Co-60 per kg of irradiated material) renders all existing operational waste management processes inadequate due to insufficient biological shielding; - Their high concentrations in long-lived radionuclides call for their disposal in a geological repository for which no final design nor waste acceptance criteria are expected prior to 2050. Uncertainties in the Belgian energy supply and security, however, require the Operator to be prepared for a partial nuclear phase out, where one or more units would benefit from lifetime extension while the remaining units would undergo decommissioning. The present paper aims at presenting how Electrabel, in partnership with Tractebel, addressed this challenge by maximizing the use of synergies within the respective sites as well as between both sites themselves, all the while accounting for site specificities. The most recent results and state of progress of the project will be detailed and the first lessons learned will be shared. The project has been split in multiple tasks and phased as follows: - An inventory phase aimed at mapping the contents, origin, composition and history of the non-fissile irradiated core items; - A pre-characterization phase based on neutron activation models; - A waste sorting phase aimed at separating waste forms for which an evacuation route exists from those for which such route does not exist; - A feasibility phase aimed at exploring all possible scenarios for the management of non-fissile irradiated core items and identifying the optimal feasible solution for each site; - A preparation phase (currently ongoing), developing further the optimal solution and ensuring that back-up solutions are available for any foreseeable change of context (licensing issue, modification in the nuclear phase-out program, etc.) and initiating early contacts with potential subcontractors for segmentation works and cask manufacturers, as well as the Belgian regulatory body and waste management agency. This phase also foresees the investigation of destructive and non-destructive radiological measurements to support the detailed characterization of the waste forms; - A realization phase (future work). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Installation and operation of GeoMelt{sup R} in container vitrification in NNL Central Lab active rig hall at Sellafield - 15328

Cleanup and subsequent immobilization of wastes on the UK's NDA estate brings major challenges. With the exception of HLW where vitrification is utilized for immobilization, cement is typically used in the management and immobilization of non-heat generating wastes. However some legacy feeds such as those containing reactive metals can be incompatible with cementation technologies, and the application of engineered cement encapsulating product to raw wastes increases wasteform volume. Cementation has also been developed to be compatible with the repository concept, which provides it with a level of protection from the environment. Potential delays in repository implementation and as yet to be determined repository waste form acceptance criteria for some wastes may require the duration of above ground storage of immobilized wastes to be extended with potential effect on the integrity of the wasteform. Vitrification is a treatment option which destroys organics and hazardous inorganics such as asbestos, reduces wasteform volume, and produces a robust leach resistant wasteform. Kurion's GeoMelt{sup R} In-container Vitrification (ICV){sup TM} is a vitrification process wherein the treatment vessel can serve as an interim-storage, transport, and final long-term storage vessel - thus negating the need to pour glass into a separate container, effectively leading to a much simpler system than traditional waste melters. The inherent volume reduction from vitrification results in a dense product with self-shielding properties, and the GeoMelt treatment process has been demonstrated to homogenize elemental metals and radionuclide contaminants of concern, eliminating hot spots and mitigating criticality concerns. All of which opens up alternatives strategies for storage and disposal. GeoMelt has been proven in the commercial and US DOE waste treatment environment with over 26,000 tonnes of hazardous waste treated to date (half of this radioactive). Kurion and the UK's National Nuclear Laboratory (NNL) plan to demonstrate its application to Sellafield radioactive wastes. A small (engineering) scale demonstrator is being installed in the active rig hall at NNL's Central Laboratory at Sellafield UK with a twofold mission; 1) demonstration of the vitrification of ILW/LLW, and 2) processing of smaller volume orphan wastes that currently have no treatment path. The NNL facility will allow the handling of uranium containing materials which form a substantial component of ILW currently stored on the Sellafield site. Accurate DFs across the melter and off gas system are often difficult to obtain using inactive materials at the low concentrations present in radioactive wastes. Installation in the NNL facility will allow trace quantities of active species such as Cs-137 and Tc-99 to be processed as part of a simulant ILW stream and monitoring will allow accurate decontamination factors (DFs) to be established to support future vitrification facility design. In addition to underpinning melter operation and obtaining DFs across the system, the facility will allow the processing of radioactive waste arisings of low activity which may either be typical of larger waste volumes or may present as orphans without a disposition route. In either case the melter facility will be able to vitrify these wastes into a form that will be suitable for interim storage and disposal. Among the feeds envisaged for processing will be asbestos from power stations, contaminated soils, feeds containing uranium metals, ponds solids, sludge, and miscellaneous active laboratory wastes. The NNL Central Lab capability will be used to analyze the active product to support disposal requirements whether through letters of compliance for the Geological Disposal Facility (GDF) or waste acceptance criteria for the Low Level Waste Repository (LLWR). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Safety Case Considerations for Deep Borehole Disposal of Cs/Sr Capsules - 16294

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE) performed an assessment of disposal options that recommended the consideration of deep borehole disposal of smaller DOE-managed waste forms, such as cesium (Cs) and strontium (Sr) capsules. To assess the feasibility of the deep borehole disposal concept for Cs/Sr capsules, safety case considerations are identified and examined. A safety case includes quantitative (e.g., safety assessments) and qualitative information related to both pre-closure (operational) and post-closure safety. For deep borehole disposal of Cs/Sr capsules, pre-closure safety considers potential hazards associated with waste package surface handling and downhole emplacement activities; post-closure safety considers scenarios for long-term radionuclide transport to the biosphere. A preliminary list of qualitative indicators of pre-closure and post-closure safety is presented. Research, development, and demonstration (RD and D) activities that can provide a quantitative basis for these indicators are discussed. In particular, DOE-NE has initiated a Deep Borehole Field Test, using surrogate test packages without radioactive waste, to further investigate various aspects of site characterization, deep drilling, and waste package handling and emplacement that would be needed for a full-scale disposal facility for Cs/Sr capsules or other DOE-managed waste forms. Finally, preliminary results from pre-closure hazard analyses and post-closure performance assessment (PA) calculations are presented as examples of supporting information for the quantitative safety case considerations. These preliminary results suggest that a favorable safety case can be developed for deep borehole disposal of Cs/Sr capsules. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Hot Isostatic Pressing of Engineered Forms of I-AgZ

Hot isostatic pressing (HIP) is being considered for direct conversion of 129 I-bearing materials to a radiological waste form. The removal of volatile radioactive 129 I from the off-gas of a nuclear fuel reprocessing facility will be necessary to comply with regulatory requirements regarding reprocessing facilities sited within the United States, and any iodine-containing media or solid sorbents generated by offgas abatement will require disposal. Zeolite minerals such as silver-exchanged mordenite (AgZ) have been studied as potential iodine sorbents and will contain 129 I as chemisorbed AgI. Oak Ridge National Laboratory (ORNL) has conducted several recent studies on the HIP of both iodine-loaded AgZ (I-AgZ) and other iodine-bearing zeolite minerals. The goal of these research efforts is to achieve a stable, highly leach resistant material that is reduced in volume as compared to bulk iodine-loaded I-AgZ. Through the use of HIP, it may be possible to achieve this with the addition of little or no additional materials (waste formers). Other goals for the process include that the waste form will be tolerant to high temperatures and pressures, not chemically hazardous, and that the process will result in minimal secondary waste generation. This document describes the preparation of 27 samples that are distinct from previous efforts in that they are prepared exclusively with an engineered form of AgZ that is manufactured using a binder. Iodine was incorporated solely by chemisorption. This base material is expected to be more representative of an operational system than were samples prepared previously with pure minerals.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Expression of Interest for off-gas system design build

Idaho National Laboratory (INL) and Pacific Northwest National Laboratory (PNNL) are studying a used nuclear fuel reprocessing system that could be used to recycle used nuclear fuel by separating waste fission products from re-useable, valuable uranium that can be used in making new nuclear fuel. This process includes dissolving the mixture of uranium and fission products in a nitric and hydrofluoric acids, then using an organic extraction process to remove and recover the uranium. The waste fission products remain in the acidic aqueous “raffinate” stream.” The fission products in this raffinate waste stream are expected to require stabilization in a vitrified glass waste form for safe disposal. INL and PNNL plan to construct and operate a facility in which to test the performance of the electric melter process that would receive a non-radioactive but chemically similar simulant of this raffinate. This test system will be located at INL in or near Idaho Falls, ID. In the melter, which operates at a temperature of about 1,050 to 1,150oC, the water, nitric acid, and other volatile and semi-volatile materials evolve into an off-gas stream, while the fission products (in this case, non-radioactive simulants of fission products) are melted with glass forming chemicals into a molten glass. Upon cooling, the molten glass provides a durable waste form for long-term disposal. Expressions of interest are requested from companies who would like to receive a request for proposal to design and build this off-gas system. We expect to request proposals and desire to award a contract prior to the end of September 2019.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Processing and disposition of remote-handled transuranic liquid waste generated at Oak Ridge National Laboratory - 15113

Oak Ridge National Laboratory (ORNL) produces transuranic (TRU) isotopes for the US Department of Energy's (US DoE's) heavy-element research program and other programs for medical and industrial applications. As a result, the research facilities generate remote-handled TRU liquid low-level waste (LLLW). This LLLW is presently collected and mixed with larger volumes of legacy and environmental management waste streams for future treatment. Once the legacy waste is disposed of as part of the US DOE Environmental Management program, the large-scale waste treatment processes will be decommissioned, and new capabilities will be needed to characterize, treat, package, and dispose of the newly generated LLLW at ORNL. Various approaches are being considered for addressing treatment and disposal of the future LLLW streams. A variety of processing methods and final waste forms is under consideration. The impacts of the chemical and radionuclide compositions of the waste stream are being evaluated. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Alloying of U-Al-SS as a Simulant for Pu-Al-SS Alloying

The Savannah River National Laboratory evaluated several options for disposition of stainless steel (SS)–clad plutonium metal, particularly Pu-10.6 at. % Al (Pu- 1.3 wt% Al) alloy fuel. One technology considered was alloying fuel with SS. The goal of the alloying would be to make a SS-Pu alloy that was a nonproliferable waste form with secondary Pu-rich microencapsulated regions distributed throughout the refractory SS. The microencapsulation of the Pu regions should therefore allow the waste form to meet the requirements for a low attractiveness waste as defined by the U.S. Department of Energy. Plutonium-bearing alloys at these levels could potentially be suitable for disposal at a waste isolation pilot plant. Four metal ingots were successfully fabricated using U and Al as a surrogate for Pu-Al. The U was distributed and microencapsulated by the alloy matrix, thereby setting the stage for subsequent tests using SS-clad fuel elements containing Pu-10.6Al.

36 MATERIALS SCIENCE↗