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DOE OSTI2022

Structural Changes in Molten Salt Fuel and/or Waste Stream Compounds Cs 2 UCl 6 and Cs 2 UO 2 Cl 4 from Room Temperature to Melting and Related Materials [Slides]

Perovskite phases have been explored in recent literature as waste forms for long term storage of radioactive waste from molten salt reactors specifically. Synthesis of new halide perovskite phases can inform these efforts as well as molten salt reactor chemistry in general. Inorganic perovskites have also shown promise as new scintillating materials for detection of X-rays and gamma rays. Modifying the structure and chemistry of these compounds, perhaps with organic components, could be used to synthesize new scintillating compounds or modify the properties of existing ones. Both U 6+ and U 4+ readily form compounds with a striking similarity to vacancy ordered perovskites (Cs 2 UO 2 Cl 4 and Cs 2 UCl 6 ). Each of these compounds is relevant to molten salt reactor chemistry and/or waste streams. These phases could from in uranium chloride salt systems or waste streams: Cs being a fission product.

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DOE OSTI2021

Multicomponent pyrochlore solid solutions with uranium incorporation – A new perspective of materials design for nuclear applications

Multicomponent pyrochlore solid solutions with and without uranium incorporation were fabricated and their thermal-mechanical properties were characterized. Multicomponent pyrochlore solid solutions without uranium exhibit comparable thermal conductivity and higher mechanical strength compared to baseline single component rare-earth titanate pyrochlore (A 2 Ti 2 O 7 ). Uranium incorporation reduces hardness as compared with single component compositions. High entropy pyrochlore with uranium displays the highest thermal conductivity within multicomponent pyrochlore solid solutions with significantly better mechanical properties than UO 2 . The measured thermal conductivity correlates well with A-site cation mixing entropy and a modified size disorder parameter, and thus the size disorder and mixing entropy could be good indicators for predicting thermal conductivity of multicomponent pyrochlore solid solutions. This work opens up the possibility of designing multicomponent oxide solid solutions by controlling their chemical disorder/mixing entropy to achieve acceptable thermal-mechanical properties, desired radiation and corrosion performance for potential nuclear waste form and inert matrix fuel applications.

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DOE OSTI2019

Implementation of Increased Salt and Decreased Sludge Volumes in the Defense Waste Processing Facility - 19597

The Savannah River Site (SRS) Liquid Waste System (LWS) safely stores and treats high-level radioactive waste. The LWS consists of 51 waste storage tanks (eight of which are operationally closed and filled with grout), waste evaporators, treatment facilities, and solidification facilities such as the Defense Waste Processing Facility (DWPF) and Saltstone Production Facility (SPF). The disposition of high-level waste (HLW) into a glass waste form poured into canisters at the DWPF began in 1996. DWPF processed a sludge only flowsheet until 2008 when the Actinide Removal Process (ARP) and Modular Caustic Side Solvent Extraction Unit (MCU) began treating the salt waste from the underground liquid storage tanks by striking with monosodium titanate (MST) to remove actinides, when necessary to meet downstream acceptance criteria, and removing the MST and entrained sludge solids (SS) by filtration before using an organic solvent to strip the cesium from the salt, creating two new concentrated HLW streams. This change caused the DWPF to begin processing under both a sludge-only and a coupled operations flowsheet where salt streams would only be added when available. This strategy allows for continuous glass canister production at DWPF. Recently, DWPF completed a 10-month melter replacement in December 2017 and the Concentrate, Storage, and Transfer Facility (CSTF) has completed repairs to the 3H Evaporator system. These facility outages have caused delays in preparing the next batch of sludge feed for DWPF. Concurrent with these facility outages, the Salt Waste Processing Facility (SWPF) has been undergoing system testing and commissioning, working toward startup. SWPF is designed to process salt waste in the same manner as ARP/MCU, but with much higher throughput volumes produced per unit time. In order to process the SWPF salt streams efficiently, DWPF will implement a coupled operation only flowsheet. Based on the facility outages mentioned above, efforts have been made to identify strategies to minimize sludge volumes and maximize salt volumes in the DWPF until the next sludge batch is ready for processing in DWPF. Savannah River Remediation (SRR) recently performed an evaluation that examines the effects of reducing the volume of sludge and maximizing the addition of the cesium-rich concentrated salt solution in each Sludge Receipt and Adjustment Tank (SRAT) batch in order to extend the existing sludge batch as long as possible until the next sludge batch is ready. These effects include a reduction of weight percent solids in the DWPF processing vessels, an increase in cycle time, a new canister heat generation value that remains within regulatory limits, and higher gamma source that remains within bulk facility gamma shielding. Balancing the above parameters resulted in a small reduction of sludge volume per SRAT batch under SWPF operation and maximizing the salt volumes for both ARP/MCU and SWPF operation. This slight reduction in sludge volume and increase in salt volumes allows SRR to mitigate some of the schedule impact incurred by the facility outages. This strategy was evaluated for both ARP/MCU and SWPF operation. This strategy has been successfully implemented under ARP/MCU operation using small step changes in the current process. This paper provides an overview of the evaluation performed for DWPF with ARP/MCU and DWPF with SWPF operations. (authors)

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DOE OSTI2019

External Engagement During Evaluations of Shutdown Nuclear Power Plant Sites - 19221

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 U.S. Department of Energy Office of Integrated Waste Management (DOE-IWM) continues to conduct evaluations of removing spent nuclear fuel (SNF) from shutdown nuclear power plant sites. The 14 shutdown sites included in the evaluation to date have nuclear power reactors that have been permanently shut down and the sites have been decommissioned or are undergoing decommissioning. The 14 shutdown sites are Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, San Onofre, Vermont Yankee, and Fort Calhoun. The Oyster Creek Nuclear Generating Station ceased operations in September 2018 and will be included in future evaluations. A key part of the shutdown site evaluations are visits to the sites. These site visits typically take place over 3 days. The first day is spent at the nuclear power plant site. The second day is spent evaluating the near-site transportation infrastructure, including rail infrastructure, potential heavy-haul routes, and potential rail and barge transload locations. The third day is often spent meeting with community engagement or advisory panels. External engagement activities during site visits are conducted in two areas. The first area involves inviting tribal and state representatives, State Regional Group (SRG) representatives, and Federal Railroad Administration (FRA) representatives to participate in the entire site visit. The SRG representatives are from the Southern States Energy Board, the Western Interstate Energy Board, the Council of State Governments - Midwest, or the Council of State Governments - Eastern Regional Conference, according to the location of the nuclear power plant. State representatives typically represent their state department of energy, state department of environmental or natural resources, State Rail Safety Participation Program, state police, state department of transportation, radiation protection organization, or emergency management organization. Tribal representatives explore issues associated with cultural affiliation and tribal involvement with past and present site activities. As the regulator of the U.S. railroads, the FRA representatives bring unique experience to the site visits. They also coordinate meetings with the railroads that serve the shutdown sites. The second area of external engagement activities during site visits involves meeting with local community engagement or advisory panels. Meeting with local community engagement or advisory panels provides an opportunity to inform these panels on DOE activities and the roles and responsibilities of federal agencies during site decommissioning. (authors)

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DOE OSTI2015

NDA Challenges during Cleanup of the Hanford 618-10 Burial Ground Trenches - 15542

The 618-10 Burial Ground on the Hanford Reservation was an early trench waste site used for the disposal of waste from the 300 Area at Hanford. Much of this waste consisted of residues from evaluations of failed fuel as well as various process and separations studies associated with the Department of Energy (DOE) mission which were performed in the 300 Area between March 1954 and September 1963. Much of this waste contained significant concentrations of fission products as well as Transuranic (TRU) radionuclides. Waste forms consist of drums and boxes which were disposed of in the trenches. Many of the drums are shielded containers, cemented sludge, or cemented debris. There are no identifying marks on the containers to correlate them to a specific waste stream. In addition there is some waste in the trenches which came from other locations outside of the 300 area. Cleanup of the 618-10 burial ground is covered under the Washington Closure Hanford (WCH) contract. Due to the proximity of the 618-10 site to the Columbia River, waste is being removed from the 618-10 site, stabilized, and disposed of into either the Environmental Restoration Disposal Facility (ERDF) if the containers are determined to be LLW or transferred to the Hanford Central Waste Complex (CWC) if the containers are suspected to be TRU waste. In order to determine the correct disposal path Non Destructive Assay (NDA) must be performed on all of the containers excavated from the trenches. The NDA of these containers is performed using germanium gamma spectroscopy with an In Situ Object Counting System (ISOCS) and with a neutron slab counter. Due to the challenges associated with these difficult waste streams, matrices, and lack of information to associate these containers with a particular process, the NDA program has had to develop a more robust process to ensure the correct disposition of the waste. (authors)

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DOE OSTI2016

Reconfirmation of frit 803 based on the January 2016 sludge batch 9 reprojection

On January 11, 2016, Savannah River Remediation (SRR) provided the Savannah River National Laboratory (SRNL) with a Sludge Batch 9 (SB9) reprojection that was developed from the analyzed composition of a Tank 51 sample. This sample was collected after field washing had been completed in Tank 51 to support the alternate reductant task. Based on this reprojection, Frit 803 is still a viable option for the processing of SB9 under sludge-only operations and coupled (Actinide Removal Process (ARP) product with and without monosodium titanate (MST)) operations. The maximum projected volumes of ARP product that can be transferred from the Precipitate Reactor Feed Tank (PRFT) per Sludge Receipt and Adjustment Tank (SRAT) batch and the resulting Na 2 O concentrations in the SRAT for coupled operations were determined. The Na 2 O concentrations in the SRAT resulting from the maximum projected ARP product transfer volumes are consistent with those from the previous assessments that were based on the August 2015 projections. Regardless of the presence or absence of MST in the ARP product, the contribution of Na 2 O to the resulting glass will be similar at the same waste loading (WL). These projected volumes of ARP product are not anticipated to be an issue for SB9. The actual transfer volumes from the PRFT to the SRAT are determined based upon the analyzed Na 2 O concentrations in the PRFT samples, which has resulted in larger transfer volumes than those allowed by the projections for Sludge Batch 8 (SB8). An operating window of 32-40% WL around the nominal WL of 36% is achievable for both sludge-only and coupled operations; however, each of the glass systems studied does become limited by waste form affecting constraints (durability) at higher volumes of ARP product and WLs of 41-42%.

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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)

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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)

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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)

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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)

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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)

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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.

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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.

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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)

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