DOE-STD-5506-2021 Implementation Evaluation
Abstract not provided.
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Results for “RADIOACTIVE WASTE PROCESSING”
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Cost Evaluation of Alternative Radioisotope Disposal Methods at the Livermore Site
The purpose of this cost evaluation was to investigate a hypothetical in which radioisotopes were no longer permitted in the City of Livermore sanitary sewer. Three alternative methods were proposed to determine estimated costs and compare current practice costs to the latter. An assumption from the initial steps of the analysis was significant cost savings would be found with the alternative methods.
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Gaseous-fuel nuclear reactor research for multimegawatt power in space
In the gaseous-fuel reactor concept, the fissile material is contained in a moderator-reflector cavity and exists in the form of a flowing gas or plasma separated from the cavity walls by means of fluid mechanical forces. Temperatures in excess of structural limitations are possible for low-specific-mass power and high-specific-impulse propulsion in space. Experiments have been conducted with a canister filled with enriched UF6 inserted into a beryllium-reflected cavity. A theoretically predicted critical mass of 6 kg was measured. The UF6 was also circulated through this cavity, demonstrating stable reactor operation with the fuel in motion. Because the flowing gaseous fuel can be continuously processed, the radioactive waste in this type of reactor can be kept small. Another potential of fissioning gases is the possibility of converting the kinetic energy of fission fragments directly into coherent electromagnetic radiation, the nuclear pumping of lasers. Numerous nuclear laser experiments indicate the possibility of transmitting power in space directly from fission energy. The estimated specific mass of a multimegawatt gaseous-fuel reactor power system is from 1 to 5 kg/kW while the companion laser-power receiver station would be much lower in specific mass.
Thom, K.↗
A 5-GWe nuclear satellite power system conceptual design
This paper presents the results of a brief study performed for MSFC on the conceptual design of a nuclear satellite power station which delivers 5 GWe net power to earth by microwave transmission. The system contains 26 modules each consisting of a reactor, fuel processing plant, Brayton PCU, space radiator, and nuclear shield. A high-temperature, gas-cooled, pebble-bed plutonium breeder concept was selected which is resupplied with fertile U-238. Sections of this core are periodically replaced and the spent fuel is chemically processed, the radioactive wastes separated, and stored for eventual space disposal. Fresh fuel pellets, formed from the U-238 and the bred plutonium, are recycled back to the reactor. The hot (1317 C) helium gas exiting the reactor serves as the working fluid in a 30%-efficient Brayton PCU.
Goodman, M.↗
Consensus Standardization of Slurry Simulant Development Process to Reduce Design Risk Within the DOE Complex
A critical responsibility of DOE’s Office of Environmental Management is the design, construction, and operation of equipment and facilities to process legacy radioactive waste slurries for safe, long-term disposal. The accomplishment of this mission requires the extensive use of physical and chemical simulants to represent the hazardous radioactive waste. Development and characterization of simulants can often be difficult to compare or utilize as questions exist as to the basis, techniques, or consistency associated with reported parameters. In addition, controversy has existed in the interpretation and application of the characterization data for the wastes that the simulants are intended to represent.
design↗
DOE's Safety Analysis Report for Packages (SARP) Review Course: Development and Evolution
Explore the source record for details and available documents.
Radioactive waste processing and disposal↗
Waste Treatment and Immobilization Plant U. S. Department of Energy Office of River Protection Submerged Bed Scrubber Condensate Disposition Project - 13460
The Hanford Waste Treatment and Immobilization Plant (WTP) will generate an off-gas treatment system secondary liquid waste stream [submerged bed scrubber (SBS) condensate], which is currently planned for recycle back to the WTP Low Activity Waste (LAW) melter. This SBS condensate waste stream is high in Tc-99, which is not efficiently captured in the vitrified glass matrix [1]. A pre-conceptual engineering study was prepared in fiscal year 2012 to evaluate alternate flow paths for melter off-gas secondary liquid waste generated by the WTP LAW facility [2]. This study evaluated alternatives for direct off-site disposal of this SBS without pre-treatment, which mitigates potential issues associated with recycling. This study [2] concluded that SBS direct disposal is a viable option to the WTP baseline. The results show: - Off-site transportation and disposal of the SBS condensate is achievable and cost effective. - Reduction of approximately 4,325 vitrified WTP Low Activity Waste canisters could be realized. - Positive WTP operational impacts; minimal WTP construction impacts are realized. - Reduction of mass flow from the LAW Facility to the Pretreatment Facility by 66%. - Improved Double Shell Tank (DST) space management is a benefit. (authors)
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Reaction of Antimony-Uranium Composite Oxide in the Chlorination Treatment of Waste Catalyst - 13521
The effect of oxygen gas concentration on the chlorination treatment of antimony-uranium composite oxide catalyst waste was investigated by adding different concentrations of oxygen at 0-6 vol% to its chlorination agent of 0.6 or 6 vol% hydrogen chloride gas at 1173 K. The addition of oxygen tended to prevent the chlorination of antimony in the oxide. When 6 vol% hydrogen chloride gas was used, the addition of oxygen up to 0.1 vol% could convert the uranium contained in the catalyst to U{sub 3}O{sub 8} without any significant decrease in the reaction rate compared to that of the treatment without oxygen. (authors)
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Comparative Evaluation of Cutting Methods of Activated Concrete from Nuclear Power Plant Decommissioning - 13548
The amount of radioactive wastes from decommissioning of a nuclear power plant varies greatly depending on factors such as type and size of the plant, operation history, decommissioning options, and waste treatment and volume reduction methods. There are many methods to decrease the amount of decommissioning radioactive wastes including minimization of waste generation, waste reclassification through decontamination and cutting methods to remove the contaminated areas. According to OECD/NEA, it is known that the radioactive waste treatment and disposal cost accounts for about 40 percentage of the total decommissioning cost. In Korea, it is needed to reduce amount of decommissioning radioactive waste due to high disposal cost, about $7,000 (as of 2010) per a 200 liter drum for the low- and intermediate-level radioactive waste (LILW). In this paper, cutting methods to minimize the radioactive waste of activated concrete were investigated and associated decommissioning cost impact was assessed. The cutting methods considered are cylindrical and volume reductive cuttings. The study showed that the volume reductive cutting is more cost-effective than the cylindrical cutting. Therefore, the volume reductive cutting method can be effectively applied to the activated bio-shield concrete. (authors)
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Proceedings of GLOBAL 2013: International Nuclear Fuel Cycle Conference - Nuclear Energy at a Crossroads
The Global conference is a forum for the discussion of the scientific, technical, social and regulatory aspects of the nuclear fuel cycle. Relevant topics include global utilization of nuclear energy, current fuel cycle technologies, advanced reactors, advanced fuel cycles, nuclear nonproliferation and public acceptance.
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Current Efforts in Immobilized Low-Activity Waste Glass Testing to Support the Hanford Site Integrated Disposal Facility - 17119
Cleanup plans for Hanford site high-level radioactive mixed tank waste calls for separating the waste into high-level waste (HLW) and low-activity waste (LAW) fractions. The fractions will be vitrified separately at the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The HLW glass will be sent to an offsite geological disposal facility while the immobilized low-activity waste (ILAW) glass will be disposed at the onsite Integrated Disposal Facility (IDF). In order to inform models used to predict the performance of ILAW glass at IDF, it is necessary to understand the corrosion behavior of the glass. In this paper, we discuss tests that account for the effects of 1) solution chemistry, 2) the ion exchange reaction, and 3) secondary phase formation on the glass dissolution rate. To quantify the effects of solution chemistry, specifically changes in pH, T, and Si concentration, the single-pass flow-through (SPFT) test method is applied. A designed set of these tests results in a set of kinetic rate law parameters that are used in a glass dissolution rate law based on transition-state theory. Furthermore, for SPFT tests with relatively high concentrations of Si, an ion exchange rate is provided. Additionally, through tests such as the pressurized unsaturated flow (PUF) test method, a suite of secondary phases can be compiled. The collection of kinetic rate law parameters, the ion exchange term, and the suite of secondary phases are provided to IDF performance assessment (PA) modelers. Lastly, we present a description of the efforts made to verify and validate the use of these parameters and secondary phases through the use of a field lysimeter present at the Hanford site. (authors)
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Low Activity Waste Pretreatment System Bench-Scale Testing: Supporting Integrated Testing and Facility Safety Analyses - 17171
The Low Activity Waste Pretreatment System (LAWPS) is being designed to enable the direct feed of waste to the Hanford Tank Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) facility to be immobilized. Prior to construction of the LAWPS, pilot-scale integrated testing of the key unit operations (crossflow filtration, ion exchange using spherical resorcinol-formaldehyde (sRF) resin) will be conducted by a team led by Washington River Protection Solutions (WRPS) to increase the technology maturation level of the facility's critical technology elements. As a part of this effort, Pacific Northwest National Laboratory (PNNL) has conducted a series of tests to perform two major objectives: (1) support pilot-scale integrated testing of the LAWPS by supplying information or performance data in advance of operating the pilot-scale facility (1/9. scale); and (2) collect data needed to establish or confirm assumptions/approaches planned for implementation in the LAWPS safety basis. The first objective was focused in two technical areas: developing simulants that are representative of expected waste feed and can be produced at larger scales, and using these simulants in a bench-scale crossflow filter to establish expected solid-liquid separation performance. The crossflow filter was also used to observe the efficacy (with respect to filter production rate) of selected operational strategies. The second objective also included two technical areas: measuring the effect of sRF resin on hydrogen generation rate under irradiation, and demonstrating that the planned hydrogen management approach is effective and robust. The hydrogen management strategy involves fluidization of the sRF resin bed in the ion exchange columns and recirculating the liquid, a scenario that is planned for testing at full column height. The full height tests at PNNL also supported full-scale IX column testing conducted as part of the technology maturation plan. The experimental approaches used at PNNL in these four technical areas are summarized and selected key preliminary results are provided. (authors)
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Use of Layered Double Hydroxide (LDH) Sorbents to Remove Radioactive Iodine Species - 17178
To more efficiently manage the radioactive iodine waste, layered double hydroxides (LDHs) were synthesized and characterized using XRD, FT-IR, SEM-EDS, and ICP-AES. Batch adsorption experiments to remove I{sup -} and IO{sub 3}{sup -} under DIW condition were conducted to check overall removal efficiencies with and without calcination using ICP-MS. Synthesized MgFe-Cl-C-21 LDHs consisting of Mg{sup 2+} and Fe{sup 3+} in a 2:1 concentration ratio of the divalent ion to the trivalent ion after calcination showed the highest I- removal efficiency, 89.2 % and synthesized CoCr-Cl-NC-41 LDHs consisting of Co{sup 2+} and Cr{sup 3+} in a 4:1 concentration ratio of the divalent ion to the trivalent ion without calcination showed the highest IO{sub 3}- removal efficiency, 97.0 %. (authors)
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Fabrication and Properties of Advanced Ceramic Technetium Waste Forms - 17311
Technetium-99 (Tc, t{sub 1/2} = 2.13 x 10{sup 5} years) is a challenge from a nuclear waste perspective and is one of the most abundant, long-lived radioisotopes found in used nuclear fuel (UNF). Within the Hanford Tank Waste Treatment and Immobilization Plant, technetium volatilises at typical glass melting temperature, is captured in the off-gas treatment system and recycled back into the feed to eventually increase Tc-loadings of the glass. In this research we are providing a technical solution for continuously depleting the Tc-content of the glass melter feed by precipitating Tc from the off-gas effluents and effectively immobilizing {sup 99}Tc as durable ceramic waste form. We have developed a wet-chemical coprecipitation-based synthesis route to fabricate gram-size Tc-based pyrochlores with the model composition Sm{sub 2}(Tc{sub 0.5-x}Ru{sub x}Ti{sub 0.5}){sub 2}O{sub 7}. These ceramic Tc waste forms exhibit better technetium retention and corrosion resistance in ASTM C1220-10 based testing compared with technetium-containing LAWE4-type borosilicate glass, combined with optimal waste loading. (authors)
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Direct Feed Options for Hanford High Level Waste Vitrification - 17318
In the high level waste (HLW) direct feed (DF) option that is under consideration for early operations of the Hanford Tank Waste Treatment and Immobilization Plant (WTP), the pretreatment facility would be bypassed in order to support an earlier start-up of the HLW vitrification facility. For HLW, this would mean that the ultrafiltration and caustic leaching operations that would otherwise have been performed in the pretreatment facility would either not be performed or would be replaced by an interim pretreatment function (in-tank washing and settling, for example). These changes could result in higher aluminum contents, higher chromium contents, a higher fraction of supernate in the HLW feed, and lower solids contents. This would likely affect glass formulations and waste loadings and have impacts on the downstream vitrification operations. In the present work, new high waste loading glass formulations have been developed and characterized for the DF HLW option. A series of scaled melter tests was then conducted to assess the impacts of increased supernate content resulting from less effective washing on wastes from candidate source tanks for the direct feed option. The melter tests were conducted at solids contents extending below the current WTP baseline, which are likely for the direct feed option. The effects on glass production rate, melter operations, and off-gas carryover were determined, as was the ability of increased bubbling to compensate for the increased evaporative load. A series of waste compositions were investigated that span the range of in-tank washing efficiencies between the baseline WTP full-wash case and the no-wash case. The tests included wastes from tank AY-102 as well as high plutonium wastes from tank A-104. The results from this work provide the initial basis for assessments of the relative merits of progressively more intensive pretreatment in HLW direct feed options versus the practical benefits of minimal pretreatment. Together, these results illustrate that the optimum washing strategy will depend on the specific tank waste composition and will need to be evaluated on a tank-by-tank basis. (authors)
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Evaluating the Impacts of Reactive Gas (Carbon Dioxide and Oxygen) Ingress on Leaching from Cementitious Waste Forms - 17482
Cementitious materials are currently being used as the waste forms for low-activity waste disposal (Saltstone) at the Savannah River Site. In addition, cementitious materials have been selected as the waste form (Cast Stone) for secondary wastes and are being investigated for use for supplemental low-activity waste treatment (Cast Stone) at the Hanford Site. These materials are subjected to numerous aging and degradation phenomena in the disposal environment that potentially impact their long-term durability and performance. One important phenomenon is the ingress of reactive gases (i.e., carbon dioxide and oxygen) that can react with and change the chemistry and pore structure and thus mass transport properties of the waste forms. The reaction of carbon dioxide (i.e., carbonation) reduces the pH of the waste form pore solution impacting the solubility of constituents of potential concern (COPCs). In Cast Stone and Saltstone, the reaction of oxygen (i.e., oxidation) impacts the solubility of redox-sensitive COPCs, notably the technetium-99 (Tc-99) and chromium. The objective of this research is to experimentally characterize the carbonation and oxidation impacts on chemistry and leaching behavior in cementitious waste forms using experimental techniques including United States Environmental Protection Agency (U.S. EPA) Method 1313 and U.S. EPA Method 1315. Results of experimental characterization will be used to parameterize reactive mass transport models to project leaching under a range of field scenarios in future applications. (authors)
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Engaging the Public on Complex Issues: Communication challenges and Solutions - 17521
Public buy-in and support can make or break a project that has local environmental effects and implications, especially if that project takes place in a large community that is in-touch with its local and national government. Because of this, effective public communication is essential to the future of projects that can have an impact on a local community. This paper addresses the challenges and methods of communicating complex, scientific and/or technical issues with the public in an easy-to-understand, engaging manner while still accomplishing the communication goals of the organization. For consideration in particular is the challenge of communicating the proposed path forward and ongoing progress being made at ORP's tank farms and Waste Treatment and Immobilization Plant project to help build support and acceptance with local and national audiences. By using a variety of engagement opportunities and communication tools, we are able to educate the public on the issues being addressed by the organization, the possible solutions to these issues, and the challenges involved. An engaged and well-informed public can assist the organization in accomplishing its mission by educating and influencing audiences that have an impact on the organization. (authors)
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Solid Secondary Waste Form Testing Activities for the Integrated Disposal Facility Performance Assessment Support - 17567
Solid secondary waste (SSW) that will be generated at the Waste Treatment and Immobilization Plant (WTP) and from other activities at the Hanford site is destined for disposal at the Integrated Disposal Facility (IDF). Construction of the IDF was undertaken based on a Performance Assessment (PA) performed in 2001. In 2016, a SSW data package was prepared to support the development of an updated PA (2017 IDF PA). There are many inputs required to support the IDF PA. Data needs were identified based on experience from the previous PAs and from the development of conceptual models and modeling approaches to be applied for the 2017 PA. Interactions with the IDF PA team and the inventory data package developer helped to focus detailed efforts on a few key waste streams, contaminants of concern, and disposal configurations that are expected to be the primary contributors to the dose resulting from SSW disposed at the IDF. The key waste streams include: High Efficiency Particulate Air (HEPA) filters, ion exchange resins, granular activated carbon, and Ag-mordenite. The key parameters addressed in the 2016 data package include: saturated hydraulic conductivity, moisture characteristic curves, effective diffusion coefficient, distribution coefficient and porosity. The IDF PA team also identified a set of key contaminants that are expected to be the primary concern for the PA calculations (Tc-99, I-129, Cs-137, Sr-90, uranium isotopes (and total uranium), chromium, mercury and nitrate). These contaminants were the focus of the SSW data package. Specific formulations had not been identified for cementitious materials that will be used to encapsulate or stabilize SSW, and no IDF-specific experiments had been conducted to obtain the data for the PA. Thus, for the 2016 SSW data package, recommended property values were provided for a range of representative candidate materials based on a review of existing literature. However, starting in FY17, IDF-specific experiments will be performed for Hanford SSW expected to contribute most significantly to the dose for the IDF. This paper will summarize the information obtained for the 2016 data package and the plans for further testing to refine key inputs for the IDF PA. (authors)
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