Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “RADIOACTIVE WASTE PROCESSING”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,315 records · Page 73
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Verification and Analysis of Spent Fuel Safety Information Database in Korea - 18252

The Public Engagement Commission on SNF management (PECOS) published the recommendation report on the spent fuel management of Korea in June, 2015. Based on this report, the national basic policy for high level radioactive waste management of Korea was established and promulgated, taking into consideration the national/international trends on policy and technology development. According to the national plan, the siting for deep geological disposal facility of high level radioactive waste will be proceeded by 2028. Once the site is selected at somewhere, centralized spent fuel storage facility will be constructed by 2035 at the same site while research program is ongoing to determine the compatibility of the site for disposal facility. According to the enforcement decree of the nuclear safety act, the definition of 'interim storage' is that a safe storage of spent fuel which is received from NPP operator for a specified period, until it is processed or permanently disposed of. It means that spent fuels should be taken over with integrated information of spent fuel from NPP operator to the KORAD (Korea Radioactive waste agency). The aim of this study is to confirm and analyze present and future spent fuel data in order to provide the basic information to the safety research on the transport and storage of spent fuel in Korea. AMORES (Automatic Multi-batch ORIGEN Runner for Evaluation of Spent fuel) was used for the aim of this study. All 16,160 bundles of PWR spent fuel database as of the end of 2015 and the expected spent fuel which will be generated from 2016 to 2053 are evaluated and verified based on national basic plan. Through scenario analysis, it was estimated that approximately 15,000 MTU of spent fuel will be transported to intermediate storage facilities by 2052. Therefore, 44%, 6%, and 33% of spent fuel will be left in Kori, Hanbit and Hanul sites, respectively. (authors)

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

Off-Gas Generation Rate during Chemical Cleaning Operations at the Savannah River Site - 12499

The enhanced chemical cleaning process (ECC) is being developed at the Savannah River Site (SRS) to remove the residual radioactive sludge heel that remains in a liquid waste storage tank. Oxalic acid is the chemical agent utilized for this purpose. However, the acid also corrodes the carbon steel tank wall and cooling coils. If the oxalic acid has little interaction with the sludge, hydrogen gas could conceivably evolve at cathodic areas due to the corrosion of the carbon steel. Scenarios where hydrogen evolution could occur during ECC include the initial filling of the tank prior to agitation and near the end of the process when there is little or no sludge present. The purpose of this activity was to provide a bounding estimate for the hydrogen generation rate during the ECC process. Sealed vessel coupon tests were performed to estimate the hydrogen generation rate due to corrosion of carbon steel by oxalic acid. These tests determined the maximum instantaneous hydrogen generation rate, the rate at which the generation rate decays, and the total hydrogen generated. The tests were performed with polished ASTM A285 Grade C carbon steel coupons. This steel is representative of the Type I and II waste tanks at SRS. Bounding conditions were determined for the solution environment. The oxalic acid concentration was 2.5 wt.% and the test temperature was 75 deg. C. The test solution was agitated and contained no sludge simulant. Duplicate tests were performed and showed excellent reproducibility for the hydrogen generation rate and total hydrogen generated. The results showed that the hydrogen generation rate was initially high, but decayed rapidly within a couple of days. A statistical model was developed to predict the instantaneous hydrogen generation rate as a function of exposure time by combining both sets of data. An upper bound on the maximum hydrogen generation rate was determined from the upper 95% confidence limit. The upper bound limit on the maximum instantaneous generation rate at 5 hours was 6.1 x 10{sup -5} m{sup 3}/m{sup 2}/minute. After two and five days the upper bound limit decayed to 7.9 x 10{sup -6} and 1.3 x 10{sup -6} m{sup 3}/m{sup 2}/minute, respectively. The total volume of hydrogen gas generated during the test was calculated from the model equation. An upper bound on the total gas generated was determined from the upper 95% confidence limit. The upper bound limit on the total hydrogen generated during the 163 hour test was 0.101 m{sup 3}/m{sup 2}. Corrosion rates were determined from the coupon tests and also calculated from the measured hydrogen generation rates. Excellent agreement was achieved between the time averaged corrosion rate calculated from the hydrogen generation rates and the corrosion rates determined from the coupon tests. The corrosion rates were on the order of 0.45 mmpy. Good agreement was also observed between the maximum instantaneous corrosion rate as calculated from the hydrogen generation rate and the corrosion rate determined by previous electrochemical tests. (authors)

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

Progress in demonstrating the performance of Engineered Barriers for the disposal of HLW and SNF: lab studies and demonstration experiments - 20293

Disposal of high-level radioactive waste (HLW) and spent nuclear fuels (SNF) in a geological disposal facility requires a fundamental understanding of the degradation processes of materials used in its engineered barrier system (EBS) over very long timescales. Beyond the wasteform itself (i.e. the borosilicate glass for HLW or the fuel matrix and cladding for SNF), materials often employed in a disposal concept include metallic containers (typically copper and/or carbon steel) and clay-based buffer materials (typically bentonite). In this paper we explore key developments on the long-term evolution of materials employed for the immobilisation, isolation and containment of HLW and SNF in a geological disposal facility, including nuclear glasses, candidate container materials and clay buffers. Our discussion is based on the latest findings from both experimental studies and in-situ demonstration experiments. We present relevant findings from the long-term evolution of HLW glasses (UK composition), metallic materials for the containment system (envisaged in Switzerland, Sweden, Canada and other countries), and bentonite clays (envisaged in many national disposal programmes). (authors)

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

A paradigm shift: an introduction to structured decision making for sustainable waste management and remediation - 15649

transport of contaminants and human health risk assessment. Most often these modeling activities are carried out in a conservative fashion, usually beyond even the conservatism that is built into the associated regulations and guidance. Decisions are aimed primarily at compliance with regulatory objectives, rather than optimization. This leads to unnecessarily costly remediation and under-utilization of the nation's relatively few radioactive waste disposal facilities. Risk assessment decisions under CERCLA and RCRA might seem to perform better, but if anything this is because the focus is on shorter-term decision-making because long term fate and transport modeling is not considered necessary in most cases. However, some more complex contamination problems suffer from the same basic issues as radioactive waste disposal in terms of the need for complex fate and transport modeling and subsequent risk assessment. At least CERCLA addresses optimization in the feasibility study, but it does not address how to perform that optimization in the face of uncertainty or stakeholders who have different competing objectives. Other environmental issues that fall outside the realm of DOE Orders, NRC regulations, and regulations such as CERCLA and RCRA, have found a different path towards effective decision-making. This includes land reuse and watershed management decisions, for which there is a focus on sustainability and on stakeholder involvement throughout the decision making process. EPA is using 'structured decision analysis' to achieve the primary goal of maximizing societal welfare. These same tools and approaches can be brought into environmental and waste management decisions that need to be made by DOE. Structured decision analysis provides a formal process for capturing not only the science side of the problem, but also the costs and value judgments of the stakeholders, to reach an optimal solution. But this approach requires a paradigm shift within the DOE and NRC environmental and waste management programs. That is, it requires a willingness to engage a new approach that will provide a path towards effective optimization in the decision-making process for the complex environmental problems faced by DOE and NRC. Because of the current nature of problem solving for these types of problems, which involves many layers of conservatism, this structured decision analysis path towards optimization will also realize substantial cost savings for these programs, while maintaining or improving defensibility and transparency in the decision making process. (authors)

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

Comparative lifecycle assessment of alternatives for waste management in Rio de Janeiro – Investigating the influence of an attributional or consequential approach

Highlights: • Management of organic solid waste in Rio de Janeiro was investigated through LCA. • An attributional approach was compared to a consequential. • Two approaches for identification of electricity marginal were compared. • The chosen approach can affect hierarchy between compared treatment alternatives. - Abstract: The environmental impacts from three management alternatives for organic fraction of municipal solid waste were compared using lifecycle assessment methodology. The alternatives (sanitary landfill, selective collection of organic waste for anaerobic digestion and anaerobic digestion after post-separation of organic waste) were modelled applying an attributional as well as consequential approach, in parallel with the aim of identifying if and how these approaches can affect results and conclusions. The marginal processes identified in the consequential modelling were in general associated with higher environmental impacts than average processes modelled with an attributional approach. As all investigated waste management alternatives result in net-substitution of energy and in some cases also materials, the consequential modelling resulted in lower absolute environmental impacts in five of the seven environmental impact categories assessed in the study. In three of these, the chosen modelling approach can alter the hierarchy between compared waste management alternatives. This indicates a risk of underestimating potential benefits from efficient energy recovery from waste when applying attributional modelling in contexts in which electricity provision historically has been dominated by technologies presenting rather low environmental impacts, but where projections point at increasing impacts from electricity provision in coming years. Thus, in the present case study, the chosen approach affects both absolute and relative results from the comparison. However, results were largely related to the processes identified as affected by investigated changes, and not merely the chosen modelling approach. The processes actually affected by future choices between different waste management alternatives are intrinsically uncertain. The study demonstrates the benefits of applying different assumptions regarding the processes affected by investigated choices – both for provision of energy and materials substituted by waste management processes in consequential LCA modelling, in order to present outcomes that are relevant as decision support within the waste management sector.

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

25 years of environmental remediation in the general separations area of the Savannah river site: lessons learned about what worked and what did not work in soil and groundwater cleanup - 15270

The Savannah River Site, which is owned and administered by the US Department of Energy, covers approximately 900 square kilometers. The General Separation Area is located in the center of the site and includes: radioactive material chemical separations facilities, radioactive waste tank farms, radioactive seepage basins, and the radioactive waste burial grounds. Radioactive wastes were disposed in the General Separation Area from the mid-1950's through the mid-1990's. Operations at the F Canyon began in 1954 and at H Canyon began in 1955. Waste water disposition to the F and H Seepage Basins began soon after operations started in the canyons. The Old Radioactive Waste Burial Ground began operations in 1952 to manage solid waste that could be radioactive from all the site operations, and ceased receiving waste in 1972. The Mixed Waste Management Facility and Low Level Radioactive Waste Disposal Facility received radioactive solid waste from 1969 until 1995. Environmental legislation enacted in the 1970's through 1990's led to changes in waste management and environmental cleanup practices. The US Congress passed the Clean Air Act in 1970 and the Clean Water Act in 1972; the Resource Conservation and Recovery Act was enacted in 1976; the Comprehensive Environmental Response Compensation, and Liability Act was enacted by Congress in 1980; and the Federal Facilities Compliance Act was signed into law in 1992. Environmental remediation at the Savannah River Site essentially began with a 1987 Settlement Agreement between the Savannah River Site and the State of South Carolina. The site manages several of the larger groundwater remedial activities under the Resource Conservation and Recovery Act for facilities recognized early on as environmental problems. All subsequent remediation projects are managed under a tri-party agreement (Department of Energy, Environmental Protection Agency, and South Carolina Department of Health and Environmental Control) through the Federal Facilities Agreement. Environmental remediation activities during the past 25 years have stabilized and capped seepage basins as well as consolidated and capped waste units and burial grounds in the General Separation Area. Groundwater activities include: pump and treat systems in the groundwater, installation of deep subsurface barrier systems to manage groundwater flow, in situ chemical treatments in the groundwater, and captured contaminated groundwater discharges at the surface for management in a forest irrigation system. Concentrations of contaminants in the aquifers beneath the General Separation Area and in surface water streams in the General Separation Area have decreased over the past 25 years. Closure of 65 waste sites and 4 Resource Conservation and Recovery Act facilities has been successfully accomplished. Environmental clean-up has progressed to the stage where most of the work involves monitoring, optimization, and maintenance of existing remedial systems. Many lessons have been learned in the process of remediating the General Separations Area. Geo-textile covers outperform low permeability clay caps, especially with respect to the amount of repairs required to upkeep the drainage layers as the caps age. Passive and enhanced natural processes to address contamination are much more cost effective than pump and treat systems. Forest irrigation can be very effective in water management situations with non-accumulative contaminants such as tritium and volatile organic contaminants. (authors)

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

Impact of Permanganate Oxidation of Glycolate on Corrosion of the Defense Waste Processing Facility (DWPF) Recycle Collection Tank (RCT), Transfer-line and Waste Tank Materials of Construction (MoC)

Savannah River Remediation (SRR) is evaluating the use of glycolic acid to replace formic acid as an alternate reductant for the Defense Waste Processing Facility (DWPF) at Savannah River Site (SRS). The replacement is aimed at reducing facility hazards, improving pH stability and rheological control, and lowering off gas production. Glycolate, which is the product that may remain, can also be carried downstream to High-Level Waste and Low-Level Waste facilities. Complete oxidation of glycolate and other organic species from the Sludge Receipt and Adjustment Tank (SRAT) may be necessary to minimize thermolytic hydrogen generation that can cause flammability issues. The glycolate oxidation may be performed in the Recycle Collection Tank (RCT). The RCT heel would be adjusted with sodium hydroxide to maintain the solution caustic and sodium nitrite as corrosion inhibitor prior to the addition of condensate that can be from the contents of the Slurry Mix Evaporator Condensate Tank (SMECT) or Offgas Condensate Tank (OGCT). The RCT tank collects the waste to be transferred via the Low Point Pump Pit – Recycle Pump Tank (LPPP-RPT) through a jacketed transfer-line to the Tank Farm facility. Scoping studies were performed to evaluate an oxidizing agent and sodium permanganate was selected as the best option for glycolate destruction. Testing was recommended to determine the impact on the Materials of Construction (MoC) due to the addition of sodium permanganate. The targeted concentration of sodium permanganate used equated to a 7:1 molar permanganate to glycolate ratio (nominally 0.002 M permanganate). However, to conservatively maximize the permanganate corrosivity, in some cases no glycolate was added.

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

Crystalline Ceramic Waste Forms: Reference Formulation Report

The research conducted in this work package is aimed at taking advantage of the long term thermodynamic stability of crystalline ceramics to create more durable waste forms (as compared to high level waste glass) in order to reduce the reliance on engineered and natural barrier systems. Durable ceramic waste forms that incorporate a wide range of radionuclides have the potential to broaden the available disposal options and to lower the storage and disposal costs associated with advanced fuel cycles. Assemblages of several titanate phases have been successfully demonstrated to incorporate radioactive waste elements, and the multiphase nature of these materials allows them to accommodate variation in the waste composition. Recent work has shown that they can be successfully produced from a melting and crystallization process. The objective of this report is to explain the design of ceramic host systems culminating in a reference ceramic formulation for use in subsequent studies on process optimization and melt property data assessment in support of FY13 melter demonstration testing. The waste stream used as the basis for the development and testing is a combination of the projected Cs/Sr separated stream, the Trivalent Actinide - Lanthanide Separation by Phosphorous reagent Extraction from Aqueous Komplexes (TALSPEAK) waste stream consisting of lanthanide fission products, the transition metal fission product waste stream resulting from the transuranic extraction (TRUEX) process, and a high molybdenum concentration with relatively low noble metal concentrations. In addition to the combined CS/LN/TM High Mo waste stream, variants without Mo and without Mo and Zr were also evaluated. Based on the results of fabricating and characterizing several simulated ceramic waste forms, two reference ceramic waste form compositions are recommended in this report. The first composition targets the CS/LN/TM combined waste stream with and without Mo. The second composition targets with CS/LN/TM combined waste stream with Mo and Zr removed. Waste streams that contain Mo must be produced in reducing environments to avoid Cs-Mo oxide phase formation. Waste streams without Mo have the ability to be melt processed in air. A path forward for further optimizing the processing steps needed to form the targeted phase assemblages is outlined in this report. Processing modifications including melting in a reducing atmosphere, and controlled heat treatment schedules are anticipated to improve the targeted elemental partitioning.

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

Improving in situ liquid SEM imaging of particles

This work presents in situ imaging of synthesized boehmite (γ-AlOOH) particles ranging from 20 to 100 nm, suspended in liquid, in a vacuum compatible microfluidic sample holder using a scanning electron microscopy (SEM) under the high vacuum mode and highlights the advantage of in situ liquid imaging of colloids. Nanometer-sized boehmite particles in high-level radioactive wastes at the Hanford site are known to be difficult to dissolve and cause rheological problems for processing in the nuclear waste treatment plant. Therefore, it is important to characterize boehmite particles and understand how they form aggregates in the liquid state. Several technical advancements are made to optimize in situ liquid SEM chemical imaging resulting in the improved ability to obtain secondary electron (SE), backscattered electron (BSE) images, and energy dispersive X-ray spectroscopy (EDX) spectra. Moreover, our results show mixed particles could be studied and identified based on the particle shape and elemental composition using in situ SEM imaging and EDX. Thus, we provide a new and improved approach to observe the evolution of particle dispersion and stability in liquid under conditions similar to those in the waste tank.

47 OTHER INSTRUMENTATION↗
DOE OSTI2018

Molecular Hydrogen Yields from the Alpha Self-Radiolysis of Nitric Acid Solutions Containing Plutonium or Americium

The yield of molecular hydrogen, as a function of nitric acid concentration, from the a-radiolysis of aerated nitric acid and sulfuric acid mixtures containing plutonium or americium has been investigated. Comparison of experimental measurements with predictions of a Monte Carlo radiation track chemistry model shows that, in addition to scavenging of the hydrated electron, its precursor, and the hydrogen atom, the quenching of excited state water is important in controlling the yield of hydrogen. In addition, increases in solution acidity cause a significant change in the track reactions that can be explained as resulting from scavenging of e aq - by H aq + to form H • . While plutonium has been shown to be an effective scavenger of precursors of molecular hydrogen below 0.1 mol dm-3 nitrate, previously reported effects of plutonium on G(H 2 ) a between 1 and 10 mol dm -3 nitric acid were not reproduced. Modelling results suggest that plutonium is unlikely to effectively compete with nitrate ions in scavenging the precursors of molecular hydrogen at higher nitric acid concentrations and this was confirmed by comparing hydrogen yields from plutonium solutions with those from americium. Finally, comparison between the radionuclide, ion accelerator experiments and model predictions leads to the conclusion that the high dose rate of accelerator studies does not significantly affect the measured hydrogen yield. These reactions provide insight into the important processes for liquors common in the reprocessing of spent nuclear fuel and the storage of highly radioactive liquid waste prior to vitrification.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Design and Analysis of a NPO2 Target for the Production of 238PU

Current production of 238Pu at Oak Ridge National Laboratory uses NpO2–aluminum cermet pellets as the feed material. The advantage of cermet is that its thermal conductivity is much higher than that of the oxide alone, which greatly reduces pellet temperatures. However, the melting point of the material is quite low due to the aluminum in the matrix, and the recovery of the 238Pu results in substantial radioactive liquid waste. There is potentially both a significant production gain and a reduction in post-irradiation processing cost and waste associated with using a pure NpO2 pellet. This work describes ongoing efforts to design a NpO2 pellet to maximize production of 238Pu in the High Flux Isotope Reactor.

McDuffee, Joel↗
DOE OSTI2012

Evaluating the efficiency of municipalities in collecting and processing municipal solid waste: A shared input DEA-model

Highlights: Black-Right-Pointing-Pointer Complexity in local waste management calls for more in depth efficiency analysis. Black-Right-Pointing-Pointer Shared-input Data Envelopment Analysis can provide solution. Black-Right-Pointing-Pointer Considerable room for the Flemish municipalities to improve their cost efficiency. - Abstract: This paper proposed an adjusted 'shared-input' version of the popular efficiency measurement technique Data Envelopment Analysis (DEA) that enables evaluating municipality waste collection and processing performances in settings in which one input (waste costs) is shared among treatment efforts of multiple municipal solid waste fractions. The main advantage of this version of DEA is that it not only provides an estimate of the municipalities overall cost efficiency but also estimates of the municipalities' cost efficiency in the treatment of the different fractions of municipal solid waste (MSW). To illustrate the practical usefulness of the shared input DEA-model, we apply the model to data on 293 municipalities in Flanders, Belgium, for the year 2008.

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

Evaluation of the Waste Streams from Management of Reprocessing Off-Gas Streams - 18483

Nuclear fuel reprocessing is being considered in the United States as part of an effort to help meet US energy needs and as a means to improve waste management. During the chemical processing of UNF to recover the valuable components for reuse, a number of volatile radionuclides will be released into the off-gas streams of the reprocessing plant. These volatile radionuclides include H-3, C-14, Kr-85, and I-129. Off-gas abatement systems for the volatile radionuclides will be required to ensure that air emissions from a UNF reprocessing facility can meet or exceed emission requirements set forth by the US Environmental Protection Agency and the US Nuclear Regulatory Commission. An initial engineering evaluation of offgas abatement systems was completed recently. The evaluation addressed the designs of the head-end offgas treatment system, which typically includes both the shear off-gas and the dissolver off-gas and vessel off-gas treatment system. Based on this study, the waste volumes for each of the four primary volatile radionuclides (H-3, C-14, Kr-85, and I-129) were estimated. Included in these estimates were the nonradioactive isotopes and the 'tramp' elements found in air (e.g., Kr and C-12 as CO{sub 2}) and process chemicals (e.g., I and Cl). Because the final waste form for these radionuclides has not been selected, estimated volumes were calculated for various alternative waste forms as appropriate. The resulting waste volume for I was found to be highly dependent on the final waste form selected and increased by a factor of 2 to 4 by tramp halogens that would be released into the off-gas stream from the process chemicals. The volume of tritiated water from the fuel was quite small, but the estimated volume of the final waste form depended on the tritium management approach employed within the reprocessing plant. The Kr waste volume was calculated assuming pure Kr (i.e., assuming complete separation of Xe and Ar). For the case of no Xe separation, the estimated waste volume increased by factor of ∼10. Argon from the air, if not removed, increased the estimated volume by a factor of 425. The estimated volume of C-14 waste was dominated by the CO{sub 2} from the air leaking into the processing cell. (authors)

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

Working in partnership with the UK nuclear industry to package waste - 16263

Radioactive Waste Management (RWM) is a wholly owned subsidiary of the Nuclear Decommissioning Authority (NDA), which is responsible for managing the effective clean-up of the UK's nuclear legacy. RWM is responsible for delivering a geological disposal facility and providing radioactive waste management solutions. To this end, RWM provides advice to waste owners to help them retrieve, condition and package waste now in ways that make them suitable for disposal later. To facilitate planning decisions and hazard reduction, RWM operates the Disposability Assessment process. This is used to provide advice to waste producing organisations on approaches to waste packaging and interim storage of packaged wastes. This advice is intended to ensure compatibility with future geological disposal requirements, as far as this is possible in advance of the availability of Waste Acceptance Criteria for a Geological Disposal Facility (GDF). As such, it is an enabler for early hazard reduction on UK nuclear sites through the facilitation of packaging. This paper explains how RWM is working with waste owners through the Disposability Assessment process to deliver optimised waste packaging solutions through proactive and early engagement. This includes specific examples of how the process has been used in a flexible way to contribute to accelerated and sustainable hazard reduction at UK sites using a partnership approach. (authors)

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

Deployment of large tank monosodium titanate strike to increase salt processing at the Savannah river site - 15115

The treatment of high level waste (HLW) salt solutions using the Actinide Removal Process (ARP) and Modular Caustic Side Solvent Extraction Unit (MCU) began at Savannah River Site (SRS) in 2008. Since that time, over 6 million gallons of liquid HLW have been treated and the resulting low level waste (LLW) effluent immobilized in cementitious grout at salt-stone. In addition to the continued stabilization of liquid HLW, operating experience from the ARP/MCU facilities has also been valuable in considering the design and construction of the Salt Waste Processing Facility (SWPF) which will utilize similar unit operations on a much larger scale. ARP removes actinides by contacting waste with monosodium titanate (MST) which is a solid adsorbent for soluble strontium and actinides. Once the sorption reaction is complete, the MST solids are separated from salt solution via ultrafiltration through a crossflow filter. The throughput of initial operations of ARP/MCU was limited by the flowrate through the extraction process. However, after the implementation of a new solvent flowsheet, filtration of MST will be the rate limiting step in salt processing. For the entirety of ARP/MCU operation, Savannah River Remediation (SRR) has sought to improve process efficiency and performance through optimization and modification of the original facility design. These efforts have resulted in a best-single-year throughput of 4.9 million liters of decontaminated salt solution. In keeping with these practices, SRR recently initiated a project to improve filtration throughput by relocating the MST strike from small batches immediately upstream of the filtration unit to the underground waste tank used to batch and store salt solution feed for ARP/MCU. This process change will allow the bulk of MST solids to decant in the waste tank prior to filtration which should improve both the nominal operational filter flux as well as the achieved utility of filter operation. (authors)

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

Immobilization of high-level waste salt in dechlorinated zeolite waste forms. Final report

This project addressed the main challenges associated with the disposal of electrorefiner salt: maximization of fission products in a final waste form and the associated processing costs. Since over half of the mass of the ER salt is chlorine in the form of non-radioactive chloride ions, our approach for achieving dechlorination is through the formation of HCl gas during the exchange of the metal ions into ultrastable H-Y (USHY) zeolite. The zeolite containing the fission products was then sintered into a chemically durable waste form.

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

How to Replace Co-60 Tracer when Faced with Unknown Variability of Initial Cobalt Content - 18059

Electricite de France (EDF), radioactive waste producer, is responsible for managing waste from dismantling of its permanently shut-down nuclear power plants. The process has been opened for more than 10 years for some plants of the first generation and is under preparation for the first pressurized water reactor (PWR) plants to be dismantled. Nowadays it is generally admitted that waste disposal is the main constraint of dismantling. In France, waste producer is required by safety rules to obtain an agreement from French governmental agency (ANDRA) before any waste production and for each type of waste under project. Such an agreement which is a technical file, is the proof that the container, the way of waste conditioning, waste composition itself and especially the activity of each radionuclide (RN), comply with all the requirements of the disposal safety studies. We should also notice in French regulation that the declaration process of activities i.e. the way the radiological content of each container is computed and declared is a prerequisite to its disposal shipment. Taking into account that EDF has already prepared numerous files describing the means to fill the container itself, the main problem to cope with now is the declaration of each RN which is present in each container ready to be shipped to the disposal. The paper points out the general case of Difficult- To-Measure (DTM) radionuclide management and specially the high difference between waste from plants in operation and waste from dismantling plants. Starting from a '3D' neutron flux cartography computation at 100% power, a specific global strategy of DTM activation RN evaluation for each type of waste has been developed. As the only tracer is Co- 60, which is almost always the case in EDF dismantling, the way to compute the global scaling factors of mix of different waste initial composition is easy when the cobalt initial composition (before irradiation by neutrons) is known. But, in some cases, cobalt is not included in the steel composition required for conception purposes but only considered as an impurity which maximum quantity is known. The result is that, even for a given family of waste corresponding to the same object for dismantling, the tracer value variability is not any more correlated with the irradiation level of the activated waste, which is completely spoiling any of its tracer property. The more the cobalt level variability is high, the more the necessary conservative choice of minimum cobalt level to compute the activation scaling factor will lead to an important overestimation of DTM RN using these scaling factor with respect to Co-60. The paper describes a method using residual thermal power measurement and removing computed Co-60 residual thermal power with the help of gamma measurement to give a new tracer without any chaotic behavior. The possible application is studied on interim storage of irradiated graphite sleeves mixed with stainless steel wires. This method avoids overestimation of graphite inventories coming from activation by neutron flux and using scaling factors with respect to any failing gamma tracer. (authors)

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

Defense Waste Processing Facility Canister Closure Weld Current Validation Testing

Two closure welds on filled Defense Waste Processing Facility (DWPF) canisters failed to be within the acceptance criteria in the DWPF operating procedure SW4-15.80-2.3 (1). In one case, the weld heat setting was inadvertently provided to the canister at the value used for test welds (i.e., 72%) and this oversight produced a weld at a current of nominally 210 kA compared to the operating procedure range (i.e., 82%) of 240 kA to 263 kA. The second weld appeared to experience an instrumentation and data acquisition upset. The current for this weld was reported as 191 kA. Review of the data from the Data Acquisition System (DAS) indicated that three of the four current legs were reading the expected values, approximately 62 kA each, and the fourth leg read zero current. Since there is no feasible way by further examination of the process data to ascertain if this weld was actually welded at either the target current or the lower current, a test plan was executed to provide assurance that these Nonconforming Welds (NCWs) meet the requirements for strength and leak tightness. Acceptance of the welds is based on evaluation of Test Nozzle Welds (TNW) made specifically for comparison. The TNW were nondestructively and destructively evaluated for plug height, heat tint, ultrasonic testing (UT) for bond length and ultrasonic volumetric examination for weld defects, burst pressure, fractography, and metallography. The testing was conducted in agreement with a Task Technical and Quality Assurance Plan (TTQAP) (2) and applicable procedures.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗