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At least 793 records · Page 44
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Zero Valent Iron for Reductive Removal of Technetium-99 from Aqueous Sulfate Solutions - 20347

This research investigates the reductive removal of technetium-99 ({sup 99}Tc) by zero valent iron (ZVI) from low activity waste (LAW) off-gas condensate simulant as a secondary treatment after recovery from vitrification planned at the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Due to its high volatility, only a fraction of Tc{sup (VII)} will be incorporated into glass waste forms. Volatilized Tc will be captured by an off-gas treatment system with current plans to recycle off-gas condensate back to the vitrification facility. The scheme with off-gas recycling will increase Tc loading in the glass waste, but will also increase the concentrations of sulfate, halides, and other problematic constituents impeding overall LAW processing and increasing volume of the glass product. This study focuses on Tc{sup (VII)} removal by ZVI as a feasible pathway to minimize off-gas condensate recycling, which may result in a reduction in the volumes of LAW waste to be immobilized and subsequent cost savings. ZVI is an established treatment agent for redox-active contaminants such as trichloroethylene, nitrate, arsenic, chromium, phenol, and others. It is a commercially available and cost-effective material. Our previous experiments showed that ZVI is very efficient for the reductive removal of {sup 99}Tc. In this work we studied ZVI oxidation under aerobic conditions at pH 7 with 0.1 M Na{sub 2}SO{sub 4} solution (ionic strength, IS, 0.3 M) in the presence and absence of {sup 99}Tc. The concentration of {sup 99}Tc and the changes in pH, dissolved oxygen (DO), and oxidation-reduction potential (ORP) of the simulated solutions were monitored over 8 days. The formation of iron oxide phases was probed by X-ray diffraction (XRD). Obtained results suggest that ZVI contact time with {sup 99}Tc containing solutions for 6 hours resulted in nearly complete removal of {sup 99}Tc from the aqueous phase. XRD analysis showed that the oxidation of ZVI is rapid with formation of magnetite (Fe{sub 3}O{sub 4}) with minor percentage of goethite and maghemite. This work is a part of a larger set of studies investigating the feasibility of {sup 99}Tc reductive removal from the LAW off-gas condensate. (authors)

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

Execution Strategy Analysis (ESA) Origin Sites Readiness Model - 19073

This is a technical paper that does not take into account the contractual limitations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). Under the provisions of the Standard Contract, DOE does not consider spent nuclear fuel in multi-assembly canisters to be an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract provisions prevail. The DOE Office of Nuclear Energy (DOE-NE) Office of Spent Fuel and Waste Disposition (SFWD) Integrated Waste Management System (IWMS) program is developing an Execution Strategy Analysis (ESA) Origin Sites Readiness Model using the Goldsim{sup TM} software application to provide information on potential future approaches for removing SNF and HLW from origin sites such as commercial nuclear power plant sites. The purpose of this ESA model is to apply integrated waste management system analysis, system engineering, and decision analysis principles to inform future decisions regarding potential future nuclear waste management system architectures. The ESA Origin Sites Readiness Model is a dynamic simulation modeling capability. It is developed for use in the analysis and comparison of potential future implementation strategies associated with an integrated nuclear waste management program. The model is designed to be independent of the destination of the SNF or HLW so the results from the Origin Sites Readiness Model are equally applicable to a wide range of scenarios involving shipments to generic destinations in an integrated waste management system. The model explicitly takes significant assumptions, uncertainties, and risks into account. Currently, the ESA Origin Sites Readiness Model includes 14 shutdown sites (Maine Yankee, Yankee Rowe, Connecticut Yankee, Vermont Yankee, La Crosse, Big Rock Point, Zion, Kewaunee, Fort Calhoun, Trojan, Rancho Seco, Humboldt Bay, San Onofre, and Crystal River) and is designed to be modular so that other sites can be added in the future. The model also includes multiple transportation modes (heavy haul truck, rail, and barge), activities related to the acquisition of transportation assets such as railcars and transportation casks, and site-specific considerations of on-site and near-site transportation infrastructure. The model was developed using the ESA process with Subject Matter Experts (SME); it incorporates uncertainty and risks and provides a structured, systematic methodology for evaluating potential SNF campaign scenarios. The model also complements other system-level logistics tools being developed by SFWD. (authors)

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

Radionuclide Behaviour During Thermal Treatment of LLW and ILW Using In-Container Vitrification - 19239

Thermal treatment is being considered in the UK for the treatment of a wide range of low and intermediate level wastes (LLW and ILW). To support this objective, a range of in-container vitrification trials were conducted to test a wide range of simulated waste streams. The vitrified products have been visually inspected and samples removed for chemical analysis. This paper presents preliminary results of this analysis with some comment on the spatial and chemical heterogeneity of the products. Treatment of a sludge simulant in a steel skip failed to fully melt the metal and produced a glass encapsulated waste-form with clearly visible porosity. Clinoptilolite produced a dense uniform glass but with some chemical heterogeneity. Some operational issues were encountered when treating water-rich sludge, but these were successfully resolved and resulted in a broadly homogenous vitrified product. Melting simulated mixed decommissioning waste produced a chemically homogenous product, despite the heterogenous nature of the waste stream. Uranium and concrete contained in a steel 'top hat' presented significant treatment issues. The top hat failed to fully melt, and this appears to have prevented the waste contained within the top hat from mixing with the glass external to the top hat. As with the steel skip containing sludge, this again left a glass encapsulated, rather than fully vitrified, product. Attempts to melt bulk PVC (an analogue of soft PCM) initially led to significant operational issues with the melt being terminated early and the waste apparently combusting rather than melting. This was resolved by pre-treating the PVC by shredding it and mixing this with the frit to give a larger contact area for melting. This approach produced a homogenous well mixed glass. Overall it appears the properties of the melt product are highly dependent on the physical and chemical properties of the waste stream being treated. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Benchmarking the New RESRAD-OFFSITE Source Term Model with DUST-MS and GoldSim - 13377

RESRAD-OFFSITE is a computer code developed by Argonne National Laboratory under the sponsorship of U.S. Department of Energy (DOE) and U.S. Nuclear Regulatory Commission (NRC). It is designed on the basis of RESRAD (onsite) code, a computer code designated by DOE and NRC for evaluating soil-contaminated sites for compliance with human health protection requirements pertaining to license termination or environmental remediation. RESRAD-OFFSITE has enhanced capabilities of modeling radionuclide transport to offsite locations and calculating potential radiation exposure to offsite receptors. Recently, a new source term model was incorporated into RESRAD-OFFSITE to enhance its capability further. This new source term model allows simulation of radionuclide releases from different waste forms, in addition to the soil sources originally considered in RESRAD (onsite) and RESRAD-OFFSITE codes. With this new source term model, a variety of applications can be achieved by using RESRAD-OFFSITE, including but not limited to, assessing the performance of radioactive waste disposal facilities. This paper presents the comparison of radionuclide release rates calculated by the new source term model of RESRAD-OFFSITE versus those calculated by DUST-MS and GoldSim, respectively. The focus of comparison is on the release rates of radionuclides from the bottom of the contaminated zone that was assumed to contain radioactive source materials buried in soil. The transport of released contaminants outside of the primary contaminated zone is beyond the scope of this paper. Overall, the agreement between the RESRAD-OFFSITE results and the DUST-MS and GoldSim results is fairly good, with all three codes predicting identical or similar radionuclide release profiles over time. Numerical dispersion in the DUST-MS and GoldSim results was identified as potentially contributing to the disagreement in the release rates. In general, greater discrepancy in the release rates was found for short-lived, fast-moving radionuclides than for long-lived, slow-moving radionuclides. (authors)

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DOE OSTI2016

One System Overview - 16579

The U.S. Department of Energy (DOE), Office of River Protection (ORP) is responsible for managing and completing the River Protection Project mission, which is to 'safeguard the nuclear waste stored in Hanford's 177 underground tanks, and to manage the waste safely and responsibly until it can be treated in the Waste Treatment and Immobilization Plant for final disposition'. The Office of River Protection's work scope is currently performed through two primary contracts: - The Tank Operations Contract, held by Washington River Protection Solutions LLC, (WRPS) includes the construction, operation, and maintenance activities necessary to safely store, retrieve, and transfer tank wastes; provide supplemental tank waste treatment; provide storage and/or disposal of primary and secondary waste streams; design, construction and commissioning of the Low-Activity Waste Pretreatment System; and support ORP to integrate Hanford Tank Waste Treatment and Immobilization Plant (WTP) completion and startup with Tank Farms operations. - The WTP Contract, held by Bechtel National, Inc. (BNI), includes the design, construction, and commissioning of a Pretreatment Facility, two vitrification facilities (one for high-level waste and one for low-activity waste), a dedicated analytical laboratory, and supporting facilities to treat radioactive tank wastes by immobilizing them into a stable glass waste form. One System was established as a means to improve the integration between the tank farms and the WTP project to support its safe and successful startup. One System's scope and purpose focus on mission integration consistent with DOE's plans to treat tank waste in phases, starting with direct feed of the low-activity waste stream to the WTP's Low-Activity Waste (LAW) Vitrification Facility (referred to as DFLAW). The One System organization now has four major focus areas: - Flowsheet Integration? Integrated River Protection Project (RPP) flowsheet development and management; developing a gaps, risks, and opportunities management plan and technical roadmap; and waste feed qualification. - Mission Analysis and Planning? Strategic planning, system planning, and modeling integration. - WTP Startup, Commissioning, and Operations Integration? Interface Control Document (ICD) management, coordinating facilities startup-to-commissioning-to-operations, contract directives and standards identification, and DFLAW program integration. - Project Integration and Controls? Integrated schedule, metrics/dashboards monitoring, and One System risk register. ORP established the DFLAW Program Office in February 2015 as one of its key performance goals. ORP staff, along with contractor managers, have identified roles and responsibilities for the each of the four major focus areas. One System has established an Executive Council and Governance Board to drive program progress and make decisions on DFLAW. This approach allows appropriate government and contractor senior management involvement in the DFLAW Program. Monthly critical path analysis on the integrated DFLAW schedule is being performed. Interface control documents specific to DFLAW have been developed, and system planning and modeling tools are being used to help define the DFLAW flowsheet. (authors)

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

Integration of Models for the Hanford Integrated Disposal Facility Performance Assessment - 18541

The Department of Energy Office of River Protection and its contractors have performed a performance assessment for the near-surface disposal of low-level and mixed low-level waste at the Hanford Integrated Disposal Facility. The long-term performance of the Hanford Integrated Disposal Facility to be protective of human health and the environment was evaluated under the requirements of United States Department of Energy Order 435.1, Radioactive Waste Management. Computer simulations were performed to evaluate whether or not the disposal facility would comply with the performance objectives specified in the order. The evaluation was performed using numerical models to simulate source term releases from vitrified and cement-based waste forms. Source term releases were subsequently transported through the facility into the natural system beneath the facility (comprised of an 85-meter thick vadose zone and thinner but highly conductive saturated zone overlying basalt) to a point of compliance 100-meters downstream from the disposed waste. Due to the differing scales in the finite difference models developed for the source terms and those developed for the natural system, numerical models for source term release and natural system flow and transport were not directly coupled. Instead, the output from the source term simulations were used as input to the natural system flow and transport simulations. The decoupling of the modeled systems makes it difficult to explore synergies and investigate the impact of parameter uncertainties. In order to explore system synergies and evaluate parameter uncertainty, an integrated system model was developed that includes source term release models, natural system transport models, and dose calculations. This paper will describe how the numerical models were used to develop abstractions that could be incorporated into the integrated system level model that was then used to explore system synergies and parameter uncertainties. (authors)

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

DOC-WMF - A Generic Model for Assessment of Waste Disposal Facilities - 19278

Numerical models and computer codes play an important role in the safety assessment of nuclear waste disposal facilities, as well as in the regulatory review of proponents' safety assessments. An important part of a safety assessment is the radiation dose calculation. This paper presents a computer code called DOC-WMF for calculating radiation dose from nuclear waste disposal facilities. DOC-WMF is a generic code that could be used for dose calculations for deep geological or near surface disposal facilities of nuclear wastes. DOC-WMF is much faster than complicated 3-D numerical models in terms of computational speed since the model uses the commercial code GoldSim as the computation engine which is based on analytical solutions for contaminant transport. As a result, it enables a user to quickly assess the dose consequences of multiple options or scenarios for a nuclear waste disposal facility. DOC-WMF was adapted from, and an improvement to, the Scoping of Options and Analyzing Risk (SOAR) model developed by the United States Nuclear Regulatory Commission (US NRC). A key improvement over the SOAR model is that DOC-WMF allows the calculation of dose contributions from other pathways such as food ingestion and air inhalation in addition to the drinking water pathway which is the sole dose contribution considered in the SOAR model. DOC-WMF also allows for: the consideration of 55 radionuclides with option to add more if required; the choice of source release method; various pathways to the biosphere; and dose calculations including an extended array of exposure pathways. In addition, DOC-WMF can simulate more complex hydrogeological settings and waste forms than SOAR does. DOC-WMF was compared against known analytical solutions for verification. It was then applied to two test cases and the results were compared. The first was a hypothetical deep geological disposal repository. The second test case was a near-surface nuclear waste disposal facility. DOC-WMF independently modeled these two test cases and achieved similar results as the more complicated numerical models did. (authors)

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DOE OSTI2013

Irradiated Nuclear Fuel Management: Resource Versus Waste

Management of irradiated fuel is an important component of commercial nuclear power production. Although it is broadly agreed that the disposition of some fraction of the fuel in geological repositories will be necessary, there is a range of options that can be considered that affect exactly what fraction of material will be disposed in that manner. Furthermore, until geological repositories are available to accept commercial irradiated fuel, these materials must be safely stored. Temporary storage of irradiated fuel has traditionally been conducted in storage pools, and this is still true for freshly discharged fuel. Criticality control technologies have led to greater efficiencies in packing of irradiated fuel into storage pools. With continued delays in establishing permanent repositories, utilities have begun to move some of the irradiated fuel inventory into dry storage. Fuel cycle options being considered worldwide include the once-through fuel cycle, limited recycle in which U and Pu are recycled back to power reactors as mixed oxide fuel, and advance partitioning and transmutation schemes designed to reduce the long term hazards associated with geological disposal from millions of years to a few hundred years. Each of these options introduces specific challenges in terms of the waste forms required to safely immobilize the hazardous components of irradiated fuel.

irradiated fuel management↗
DOE OSTI2015

Process intensification of biodiesel production by using microwave and ionic liquids as catalyst

The energy crisis pushes the development and intensification of biodiesel production process. Biodiesel is produced by transesterification of vegetable oils or animal fats and conventionally produced by using acid/base catalyst. However, the conventional method requires longer processing time and obtains lower yield of biodiesel. The microwave has been intensively used to accelerate production process and ionic liquids has been introduced as source of catalyst. This paper discusses the overview of the development of biodiesel production through innovation using microwave irradiation and ionic liquids catalyst to increase the yield of biodiesel. The potential microwave to reduce the processing time will be discussed and compared with other energy power, while the ionic liquids as a new generation of catalysts in the chemical industry will be also discussed for its use. The ionic liquids has potential to enhance the economic and environmental aspects because it has a low corrosion effect, can be recycled, and low waste form.

09 BIOMASS FUELS↗
DOE OSTI2016

Sensitivity Analysis of Kinetic Rate-Law Parameters Used to Simulate Long-Term Weathering of ILAW Glass. Erratum

This report refers to or contains K g values for glasses LAWA44, LAWB45 and LAWC22 affected by calculations errors as identified by Papathanassiu et al. (2011). The corrected K g values are reported in an erratum included in the revised version of the original report. The revised report can be referenced as follows: Pierce E. M. et al. (2004) Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment. PNNL-14805 Rev. 0 Erratum. Pacific Northwest National Laboratory, Richland, WA, USA.

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

Integrated Disposal Facility FY 2012 Glass Testing Summary Report, Erratum

This report refers to or contains K g values for glasses LAWA44, LAWB45 and LAWC22 affected by calculations errors as identified by Papathanassiu et al. (2011) The corrected K g values are reported in an erratum included in the revised version of the original report. The revised report can be referenced as follows: Pierce E. M. et al. (2004) Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment. PNNL-14805 Rev. 0 Erratum. Pacific Northwest National Laboratory, Richland, WA, USA.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Granite disposal of U.S. high-level radioactive waste.

This report evaluates the feasibility of disposing U.S. high-level radioactive waste in granite several hundred meters below the surface of the earth. The U.S. has many granite formations with positive attributes for permanent disposal. Similar crystalline formations have been extensively studied by international programs, two of which, in Sweden and Finland, are the host rocks of submitted or imminent repository license applications. This report is enabled by the advanced work of the international community to establish functional and operational requirements for disposal of a range of waste forms in granite media. In this report we develop scoping performance analyses, based on the applicable features, events, and processes (FEPs) identified by international investigators, to support generic conclusions regarding post-closure safety. Unlike the safety analyses for disposal in salt, shale/clay, or deep boreholes, the safety analysis for a mined granite repository depends largely on waste package preservation. In crystalline rock, waste packages are preserved by the high mechanical stability of the excavations, the diffusive barrier of the buffer, and favorable chemical conditions. The buffer is preserved by low groundwater fluxes, favorable chemical conditions, backfill, and the rigid confines of the host rock. An added advantage of a mined granite repository is that waste packages would be fairly easy to retrieve, should retrievability be an important objective. The results of the safety analyses performed in this study are consistent with the results of comprehensive safety assessments performed for sites in Sweden, Finland, and Canada. They indicate that a granite repository would satisfy established safety criteria and suggest that a small number of FEPs would largely control the release and transport of radionuclides. In the event the U.S. decides to pursue a potential repository in granite, a detailed evaluation of these FEPs would be needed to inform site selection and safety assessment.

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

Software Quality Assurance for LeachXS/ORCHESTRA as a Tool for Assessment of Waste Management Systems Performance - 17486

Qualified modeling and simulation tools and experimental data are needed to provide a reasonable basis for the incorporation of realistic wasteform retention assumptions (i.e., those without orders-of-magnitude conservatism) into performance assessments and regulatory decisions. Use of highly conservative assumptions in performance assessments has historically led to the overprediction of release and near-field transport by as much as 4- to 6-orders of magnitude thereby limiting the nature and quantities of certain contaminants of concern in near surface disposal facilities. This project provides peer-reviewed models and characterization methods to provide more reasonable assumptions for use in performance assessments and regulatory decisions. Simulations tools have been developed to predict hydraulic and chemical performance of waste-forms and barriers used in nuclear applications under several important aging and degradation scenarios, including sulfate attack, carbonation, oxidation, cracking, and leaching; these tools are complaint with DOE Order 414.1D and ASME NQA-1. Wasteform characterization methods have been developed that are consistent with US EPA's Leaching Environmental Assessment Framework (LEAF) currently part of EPA's Hazardous Waste Test Methods (SW-846); the resulting experimental data are evaluated using LeachXS. ORCHESTRA models, that can be called seamlessly from LeachXS, have been developed representing both laboratory and field conditions; these models have undergone an extensive Software Quality Assurance process where they have been calibrated (where necessary), verified, and validated to be compliant with relevant guidelines. The models used for field predictions have also been incorporated into the Cementitious Barriers Partnership (CBP) Software Toolbox that allows probabilistic evaluation of field scenarios. The foci over the next year will be on providing 1) a mechanistic basis for projecting the retention of Tc-99, I-129, and other important radionuclides in near-surface disposal facilities at the Savannah River and Hanford Sites and 2) a technical basis for assessing the efficacy of macro-encapsulation of mercury-contaminated debris at the Oak Ridge Reservation. (authors)

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

What's in WIPP? Packaging TRU Waste to Enhance WIPP's Capacity - 16515

Beginning in 1999, and continuing for 15 years, the US DOE Waste Isolation Pilot Plant (WIPP) repository successfully disposed of defense Transuranic (TRU) waste. Six of the originally planned ten disposal panels were filled before two unrelated accidents temporarily halted disposal operations in February 2014. Based on counting the gross volume of all containers emplaced in WIPP to date, a little more than half of the legislated volume capacity has been subscribed (91,000 cubic meters). This paper describes the many different packaging configurations that have been used to ship TRU waste to WIPP, and makes observations that may enhance the repository capacity when shipments resume. The contents of shipments to WIPP are limited by many requirements of the repository waste acceptance criteria. These criteria derive from three primary sources; 1) the Certificate of Compliance (CoC), issued by the NRC for Type B shipping packaging used by WIPP; 2) The Hazardous Waste Facility Permit, issued by the New Mexico Environment Department; and 3) the WIPP facility's own Documented Safety Analysis (DSA), developed under DOE's regulation 10CFR830. To meet these criteria, various TRU waste forms have been packaged in different ways, with varying packaging efficiency. Of course, there is an overall upper limit on WIPP's volume capacity that was legislated as part of the WIPP Land Withdrawal Act. The volume counted against this limit has been calculated by simply summing the volumes of individual payload containers shipped to and emplaced at WIPP. This method counts significant non-waste volumes inside the containers due to over-packing inner containers and partially filled containers. Without some change to how disposal volumes are calculated, or how waste is packaged, the amount of TRU waste yet to be shipped to WIPP could exceed the legislated capacity. Concepts to enhance the packaging efficiency of future waste shipments to WIPP are presented. Approximate estimates of the waste versus non-waste volumes that have been emplaced to date are discussed. This paper only considers contact handled TRU waste (i.e., contact dose rates < 2 mSv/hr). (authors)

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

Reuse of Ion Exchange Resins for Reduction of Spent Radioactive Resins - 17125

A process is proposed to reuse ion exchange resins. The process consists of ion exchange beds for the ion exchange and regeneration, regeneration solution (regenerant) make-up tanks, an evaporator for concentration of spent regenerant, a dryer for the drying of concentrate, a pulverizer for grinding dried salt and a mixer for polymer solidification of the salt. A spent resins storage tank and a spent regenerant storage tank are also involved in the process. The role of each component in the process for reuse of the resins is explained. In addition, the effect of the reuse of ion exchange resins by regeneration was estimated in view of the reduction of spent resins. As the number of regenerations increases, the volume of spent resins is obviously inversely proportional to the number of regenerations. Although a polymer waste form is generated instead of spent resins when the ion exchange resins are reused, its volume is only about one-fifth of that of spent resins. (authors)

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

Chemical Reactivity and Recommended Remediation Strategy for Los Alamos Remediated Nitrate Salt (RNS) Wastes

Nitrate salt wastes were generated during historical processing operations to recover and purify plutonium at Los Alamos National Laboratory (LANL). These nitrate salts have the potential to behave as oxidizers, and to contain entrained liquids, both of which are prohibited in the Waste Isolation Pilot Plant (WIPP). A process was developed to remediate nitrate salts by mixing salts with an inorganic zeolite pet litter to absorb free liquids, and remove the oxidizing characteristic of the wastes. Process changes introduced instead an organic wheat-­-based pet litter along with organic neutralizers, generating a new waste form referred to as Remediated Nitrate Salt (RNS) waste. The February 2014 radiological release at the WIPP site was traced to a LANL RNS waste drum number 68660. Experimental and modeling studies performed at LANL indicate that mixtures of metal nitrate salts (oxidizer) with Swheat organic kitty litter (fuel) create the potential for exothermic chemical reactions. Even in the presence of triethanol-­-ammonium nitrate (TEAN) created during neutralization, relatively high temperatures, greater than approximately 160 °C (320 °F), are required for runaway exothermic reactions. Evaluation of Drum 68660 characteristics coupled with extensive chemical testing indicate that, in addition to the nitrate salt/Swheat organic kitty litter mixture, an additional trigger mechanism (or mechanisms) is likely required to raise the internal drum temperature high enough to initiate the nitrate salt/Swheat organic kitty litter reaction. A combination of chemical conditions were identified that may lower the temperature for reaction, including initial high acid concentration of free liquids; significant quantities (> 1 gal) of neutralized, absorbed free liquids; the presence of reactive or catalytic metals like magnesium, iron, or lead; the presence of bismuth containing glovebox gloves; and the presence of natural biological activity. Complex surrogate nitrate salt mixtures prepared to simulate wastes, particularly those containing iron and magnesium, can generate NOx gases that partially nitrate the organic Swheat kitty litter and form a more energetic fuel. These complex surrogate salt mixtures display exothermic behavior at temperatures as low as 60 °C (140 °F) which is still well above the ambient temperature conditions experience by a drum. Neutralization of free liquids, and sorption onto Swheat establishes conditions (moisture) with near-neutral pH) that will support natural biological activity. Spontaneous self-heating generated by low-level chemical reactions and/or the respiration of bacteria, molds, and microorganisms is potentially important in the early stages and may be sufficient to raise the temperature as high as 60 °C (140 °F), where the other exothermic chemical reactions can take place. Additional studies are being conducted to evaluate the role biological activity may have played in initiating the event. The use of Swheat absorbent in the processing of nitrate salt wastes can be pinpointed as the critical processing decision that led to the failure of drum 68660, regardless of the ii details of the thermal processes that enabled the drum to achieve temperatures sufficient to initiate the chemical reactions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Generic Natural Systems Evaluation - Thermodynamic Database Development and Data Management

Thermodynamic data are essential for understanding and evaluating geochemical processes, as by speciation-solubility calculations, reaction-path modeling, or reactive transport simulation. These data are required to evaluate both equilibrium states and the kinetic approach to such states (via the affinity term or its equivalent in commonly used rate laws). These types of calculations and the data needed to carry them out are a central feature of geochemistry in many applications, including water-rock interactions in natural systems at low and high temperatures. Such calculations are also made in engineering studies, for example studies of interactions involving man-made materials such as metal alloys and concrete. They are used in a fairly broad spectrum of repository studies where interactions take place among water, rock, and man-made materials (e.g., usage on YMP and WIPP). Waste form degradation, engineered barrier system performance, and near-field and far-field transport typically incorporate some level of thermodynamic modeling, requiring the relevant supporting data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

The benefits of a fast reactor closed fuel cycle in the UK

The work has shown that starting a fast reactor closed fuel cycle in the UK, requires virtually all of Britain's existing and future PWR spent fuel to be reprocessed, in order to obtain the plutonium needed. The existing UK Pu stockpile is sufficient to initially support only a modest SFR 'closed' fleet assuming spent fuel can be reprocessed shortly after discharge (i.e. after two years cooling). For a substantial fast reactor fleet, most Pu will have to originate from reprocessing future spent PWR fuel. Therefore, the maximum fast reactor fleet size will be limited by the preceding PWR fleet size, so scenarios involving fast reactors still require significant quantities of uranium ore indirectly. However, once a fast reactor fuel cycle has been established, the very substantial quantities of uranium tails in the UK would ensure there is sufficient material for several centuries. Both the short and long term impacts on a repository have been considered in this work. Over the short term, the decay heat emanating from the HLW and spent fuel will limit the density of waste within a repository. For scenarios involving fast reactors, the only significant heat bearing actinide content will be present in the final cores, resulting in a 50% overall reduction in decay energy deposited within the repository when compared with an equivalent open fuel cycle. Over the longer term, radiological dose becomes more important. Total radiotoxicity (normalised by electricity generated) is lower for scenarios with Pu recycle after 2000 years. Scenarios involving fast reactors have the lowest radiotoxicity since the quantities of certain actinides (Np, Pu and Am) eventually stabilise. However, total radiotoxicity as a measure of radiological risk does not account for differences in radionuclide mobility once in repository. Radiological dose is dominated by a small number of fission products so is therefore not affected significantly by reactor type or recycling strategy (since the fission product will primarily be a function of nuclear energy generated). However, by reprocessing spent fuel, it is possible to immobilise the fission product in a more suitable waste form that has far more superior in-repository performance. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗