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DOE OSTI2016

Applicability of Wireless Power Transfer for Monitoring Technology of Radioactive Waste Geological Disposal - 16156

To assure the engineered barrier of a high-level radioactive waste repository functions correctly, long-term monitoring inside the backfilled deposition holes and tunnels is planned. Cables are used to power monitoring equipment and send signals however complete isolation will not be possible. Previously developed monitoring systems send information wirelessly and used batteries. Due to the limited battery life, these systems cannot be used for long-term monitoring. In this paper a wireless power transfer method based on magnetic resonant coupling is presented. The necessary transmitter and receiver coils have been developed. The performance of the individual coils has been tested. The performance of the system of transmitter and receiver with both air and bentonite as medium was tested. The results showed that the system with bentonite performed similarly well as the system with air. Wireless power transfer based on magnetic resonant coupling is therefore a feasible method for high-level radioactive waste repository monitoring applications which could in principle work for a period of 100 year or longer. (authors)

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DOE OSTI2016

Results for the Fourth Quarter Calendar Year 2015 Tank 50H Salt Solution Sample

In this memorandum, the chemical and radionuclide contaminant results from the Fourth Quarter Calendar Year 2015 (CY15) sample of Tank 50H salt solution are presented in tabulated form. The Fourth Quarter CY15 Tank 50H samples were obtained on October 29, 2015 and received at Savannah River National Laboratory (SRNL) on October 30, 2015. The information from this characterization will be used by Defense Waste Processing Facility (DWPF) & Saltstone Facility Engineering for the transfer of aqueous waste from Tank 50H to the Salt Feed Tank in the Saltstone Production Facility, where the waste will be treated and disposed of in the Saltstone Disposal Facility. This memorandum compares results, where applicable, to Saltstone Waste Acceptance Criteria (WAC) limits and targets. Data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals will be documented at a later time per the Task Technical and Quality Assurance Plan (TTQAP) for the Tank 50H saltstone task. The chemical and radionuclide contaminant results from the characterization of the Fourth Quarter Calendar Year 2015 (CY15) sampling of Tank 50H were requested by SRR personnel and details of the testing are presented in the SRNL Task Technical and Quality Assurance Plan.

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DOE OSTI2016

Results For The Third Quarter Calendar Year 2016 Tank 50H Salt Solution Sample

In this memorandum, the chemical and radionuclide contaminant results from the Third Quarter Calendar Year 2016 (CY16) sample of Tank 50H salt solution are presented in tabulated form. The Third Quarter CY16 Tank 50H samples (a 200 mL sample obtained 6” below the surface (HTF-5-16-63) and a 1 L sample obtained 66” from the tank bottom (HTF-50-16-64)) were obtained on July 14, 2016 and received at Savannah River National Laboratory (SRNL) on the same day. Prior to obtaining the samples from Tank 50H, a single pump was run at least 4.4 hours, and the samples were pulled immediately after pump shut down. The information from this characterization will be used by Defense Waste Processing Facility (DWPF) & Saltstone Facility Engineering for the transfer of aqueous waste from Tank 50H to the Saltstone Production Facility, where the waste will be treated and disposed of in the Saltstone Disposal Facility. This memorandum compares results, where applicable, to Saltstone Waste Acceptance Criteria (WAC) limits and targets. Data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals will be documented at a later time per the Task Technical and Quality Assurance Plan (TTQAP) for the Tank 50H saltstone task. The chemical and radionuclide contaminant results from the characterization of the Third Quarter CY16 sampling of Tank 50H were requested by Savannah River Remediation (SRR) personnel and details of the testing are presented in the SRNL TTQAP.

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DOE OSTI2018

Application of SIAL{sup R} Solidification Technology for Processing of Problematic Waste Streams - 18267

SIAL{sup R} is a proven, safe, cost-effective and licensed waste solidification technology that has been used successfully for 20 years in Slovakia and the Czech Republic for processing of various problematic radioactive waste streams. This abstract presents and summarises the most interesting performance records of the SIAL{sup R} geopolymer solidification technology, the applied mobile devices as well as the characteristics of final products. Radioactive waste solidification into matrix SIAL{sup R} is based on polycondensation of inorganic aluminosilicate components with chemical composition similar to average cement. The main components of matrix SIAL{sup R} are natural inorganic raw materials containing SiO{sub 2}, Al{sub 2}O{sub 3}, CaO, MgO, Na{sub 2}O, K{sub 2}O, Fe{sub 2}O{sub 3}, and TiO{sub 2}. The treatment/solidification process is modified for individual radioactive streams with different compositions. To determine the most appropriate composition, the SIAL{sup R} application technology is laboratory tested on real samples of each individual waste stream identified for solidification. Then an appropriate mixture of aluminosilicate composition and other inorganic compounds is defined for a given solidification application. When determining the most appropriate mixture, other requirements are taken into consideration: maximum dose rate on the drum with solidified product, mass, volume, alpha radionuclide content in the final product, need for a waste's dewatering before solidification, dry matter content in the solidification product as well as other specific waste acceptance criteria (i.e., compressive strength, leachability, diffusion coefficients, radiation stability, biodegradability of matrix, distribution and partition coefficients, influence of frost and water presence to the product stability including long term physical stability in water, dustiness, flammability, time dependence of compression strength and water reduction, product homogeneity and also radiolytic G-value for hydrogen release, GH{sub 2}). The Amec Foster Wheeler baseline approach to implementing the SIAL{sup R} waste treatment process is structured on three steps - waste characterization, waste retrieval and pre-treatment and at last the solidification of pre-treated waste itself. The equipment used for the solidification process is relatively small, lightweight and designed for the specific application on site, directly 'in situ' or local to the tanks storing the sludge's/resins/other waste streams. This mitigates the problems normally experienced by transferring sticky, tight waste through long pipes. The equipment is normally 'tailor-made' to take into account the real space conditions and other existing facility requirements. At Bohunice and Dukovany sites, the facilities for encapsulation were designed to occupy existing areas of the plants in close proximity to the tanks where the waste was being stored. This allowed for the removal of the waste using Remote Operated Vehicles and for the waste to be immediately treated on site without the need for transportation thus saving both, time and resource. To date, SIAL{sup R} has been used successfully to immobilize approximately 2 x 10{sup 6} kg of waste which includes sludge, resin, crystalline borates and contaminated organic waste from Bohunice and Mochovce, Slovakia and approximately 300 m{sup 3} of spent ion exchange resins and sludge from tanks on site at the Dukovany NPP (Nuclear Power Plant) in the Czech Republic. The technology now has a track record of over 20 years which includes on-going research and development. The International missions WANO (the World Association of Nuclear Operators) and OSART (the Operational Safety Review Team) evaluated the SIAL{sup R} matrix technology at Dukovany NPP as an example of good practice. (authors)

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DOE OSTI2017

SRS Defense Waste Processing Facility - the Nation's Only Operating Vitrification Facility - Completes 20 Years of Radioactive Operations - 17407

Savannah River Remediation (SRR) is the current Liquid Waste contractor at the U.S. Department of Energy's (DOE) Savannah River Site (SRS), located near Aiken, South Carolina, and operates the Defense Waste Processing Facility (DWPF), along with the other SRS liquid waste facilities. The DWPF, the nation's only operating vitrification facility, completed 20 years of radioactive operations in 2016. During this period, DWPF has safely treated the high-level liquid waste at this DOE facility since radioactive operations began with the first transfer of sludge waste feed to DWPF on March 12, 1996. A little over a month later, on April 29, 1996, the first radioactive canister was poured. DWPF poured its 4,000. canister of vitrified waste on December 31, 2015. As of September 30, 2016, DWPF has removed approximately 59.9 million curies from the liquid waste system at SRS and placed them in 4,106 canisters. SRS has operationally closed 8 high-level waste tanks. The DWPF melter has poured nearly 7.14 million kg of molten glass, and the facility has used only two melters in its lifetime. Melter 1 poured 1,339 canisters (1996-2002), while melter 2 to date (September 30, 2016) has poured 2,767 canisters (2003-present). Melter 2 continues to safely and efficiently operate. In order to achieve these accomplishments, the facility has chronicled a history of continued operational improvements such as the installation of bubblers in the melter to increase production, incorporation of glass compositions to achieve higher waste loadings, and improvements to melter design to extend melter life. As SRR moves forward to the next 20 years, continued improvements as well as maintaining the aging infrastructure through a proactive system health program are key priorities. This paper chronicles 20 years of challenges, improvements and accomplishments, and discusses future operations and infrastructure upgrades including implementation of a key flowsheet change to significantly reduce hydrogen generation in the Chemical Process Cell. (authors)

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DOE OSTI2024

New Tank Mapping Method Improves Waste Removal Process

Savannah River Mission Completion is the Liquid Waste (LW) contractor at the Savannah River Site (SRS). The LW mission is tasked with treating and disposing of legacy nuclear waste. There are multiple facilities involved in this work, including the Concentration, Storage, and Transfer Facilities (CSTF), the Defense Waste Processing Facility (DWPF), the Salt Waste Processing Facility (SWPF), and the Saltstone Production Facility (SPF). The CSTF includes 43 underground waste tanks used to store and support processing of radioactive liquid waste. Waste removal activities, such as salt dissolution campaigns and sludge agitation, are conducted within the CSTF waste tanks to convert the waste into a form that allows for downstream processing at other LW facilities. While performing these waste removal campaigns, camera inspections are performed to assess the quantity and distribution of the remaining waste within the waste tank (i.e. saltcake or sludge). Understanding the quantity and distribution of the salt/sludge within the waste tanks allows for improved waste removal strategies (e.g. mixing pump operation) and refined safety controls. Typically, several camera inspections are performed during a waste removal transfer to verify the elevation of the visible salt/sludge mounds against the known elevation of the liquid surface. The camera inspection footage must then be interpreted by a trained engineer who will develop a 2-D map that depicts the waste distribution at various elevations within the waste tank. This tank mapping is then used in conjunction with conservative assumptions to evaluate the volume of saltcake or sludge that is present within the waste tank.

Mini, Melany↗

DOE OSTI2019

Cradle-to-Grave: Assessment Methods for Hazardous Waste Management

The Resource Conservation and Recovery Act (RCRA) was passed to set up a framework for the proper management of waste. The 40CFR260-273 regulation established the regulatory framework to handle solid and hazardous waste, from the point of generation to its disposal; or cradle-to-grave management. Per these regulations, a hazardous waste generator is responsible for its waste from initial generation through its ultimate disposal and beyond. Nothing will relinquish a generator of this responsibility. There is no expiration date or time limit, and hiring a someone else to transport and dispose of waste does not transfer responsibility. If an incident occurs at any point during the life cycle of a waste, the generator can be held liable. The first step to compliance with RCRA is proper characterization of the waste. Waste may either be listed specifically or may show a characteristic as defined in the regulation: Is my waste hazardous? Why is it Hazardous? What should I do as a generator? Plan chemical use: - Create a chemical inventory for your research project. - Use existing chemicals if available instead of ordering new ones. - Brainstorm methods that will help to reduce toxicity and minimize waste. Waste Inventory: - Label all waste bottles with: ID no., What is it, Whose is it, Hazards, and Date. - Add waste containers into the inventory log. - Initiate/complete the container sheet to describe the waste. Satellite Areas: - Store waste in containers compatible with the waste with lids on tight. - Store waste containers properly and segregate them according to their specific hazards. - Use secondary containment for liquids. Store acids and bases separately; apart from cyanides or sulfides. Acids should never be put into steel containers. Water-reactive, strong acids such as organic acid halides, organic acid anhydrides, inorganic acid anhydrides, and strong acidic salts must be kept apart from both alkalis and water. Oxidizing agents must be kept apart from reducing agents and organic compounds. Water-reactive agents must be stored apart from water, aqueous solutions, and acids. Air-reactive materials must be packed in containers that are sealed off from the atmosphere. Explosive and shock-sensitive materials present risks that require special handling.

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DOE OSTI2015

Computational fluid dynamics modeling of high level waste plug formation - 15228

At Hanford, an extensive network of pipelines traversing several miles is used to transfer the high level radioactive nuclear waste from tanks to the treatment facilities. During transfer operations, however, there is a potential risk for the transport lines to plug, causing significant delays, increasing operation costs and creating hazardous conditions for personnel and the environment. In this study we investigate the phenomena of plug formation in the pipelines due to solids depositing during the waste transfer process. A three-dimensional (3-D) Computational Fluid Dynamics (CFD) model has been developed using Comsol Multiphysics 4.3b by Florida International University (FIU) to evaluate the factors that can contribute to pipeline plug formation. Several virtual scenarios are developed simulating the settling mechanics as a function of particle size, solids volume fraction and solids density. The numerical results are validated with the available experimental results and the established critical velocity correlations. The 3-D numerical results were also compared with the previous 2-D numerical studies to understand the trade-off between the two studies in terms of computing speed and numerical accuracy. (authors)

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DOE OSTI2012

Thermal Analysis for Ion-Exchange Column System

Models have been developed to simulate the thermal characteristics of crystalline silicotitanate ion exchange media fully loaded with radioactive cesium either in a column configuration or distributed within a waste storage tank. This work was conducted to support the design and operation of a waste treatment process focused on treating dissolved, high-sodium salt waste solutions for the removal of specific radionuclides. The ion exchange column will be installed inside a high level waste storage tank at the Savannah River Site. After cesium loading, the ion exchange media may be transferred to the waste tank floor for interim storage. Models were used to predict temperature profiles in these areas of the system where the cesium-loaded media is expected to lead to localized regions of elevated temperature due to radiolytic decay. Normal operating conditions and accident scenarios (including loss of solution flow, inadvertent drainage, and loss of active cooling) were evaluated for the ion exchange column using bounding conditions to establish the design safety basis. The modeling results demonstrate that the baseline design using one central and four outer cooling tubes provides a highly efficient cooling mechanism for reducing the maximum column temperature. In-tank modeling results revealed that an idealized hemispherical mound shape leads to the highest tank floor temperatures. In contrast, even large volumes of CST distributed in a flat layer with a cylindrical shape do not result in significant floor heating.

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DOE OSTI2017

Developments in the Regulatory Framework of Interim Storage of Spent Nuclear Fuel (SNF) in Germany - 17374

According to article 6 of the German Atomic Energy Act, storage of nuclear fuel requires a license, granted by the competent authority in this field, which has been the Federal Office of Radiation Protection (BfS) until July 2016 and currently is the Federal Office for the Safety of Nuclear Waste Management (BfE). A first license was granted in 1983 for storage of spent nuclear fuel (SNF) and vitrified high level waste (HAW) at the interim storage facility in Gorleben. This license was reviewed in 1995. Dry Interim storage at the NPP sites in its current form started in 2002 in the interim storage facility next to the NPP Lingen. Since that time at each site of an operational NPP storage facilities were licensed and erected. Thus additionally to the three central storage facilities (Ahaus, Gorleben and Rubenow) there are now 12 storage facilities onsite the NPPs. All of these facilities use dry storage in metallic dual purpose casks (DPCs). The actual storage licenses for SNF and the HAW (which is stored in Gorleben and Rubenow) are limited to 40 years after closing the transport and storage cask. The first license to expire will be the storage license for the interim storage facility in Gorleben in the year 2034. The German 'Programme for the safe and responsible management of spent fuel and radioactive waste' expects to take the disposal facility into operation in the year 2050. Thus necessary preparations for interim storage till that time have to be taken on the side of the regulator. In the German Atomic Energy Act precautions have already been taken and a consulting of the German parliament (Bundestag) is mandatory if inevitable events would require an extension of currently licensed interim storage period. In the year 2016, the regulatory body in Germany was given a new structure. The BfS was formerly competent authority in the field of radiation protection since 1989. It had competencies for ionizing and non-ionizing radiation protection (e.g. mobile phone radiation, high voltage power lines and the effects of electric fields), licensing storage of spent nuclear fuel and as the operator of facilities for final disposal. Now there is a new structure with two Federal Offices (BfS and BfE) and a federal company for final disposal (BGE). The BfE is responsible for regulation in the field of nuclear disposal since 30.07.2016. So the package design approval for DPCs and the storage licenses for SNF are now granted by this office. The search for a repository site will be coordinated there as well. The site selection, construction and operation of the repository will be in the responsibility of BGE. The parts of the former BfS dealing with recovery of the waste from Asse (a former salt mine used for disposal research) and the construction of the repository Konrad (for low and medium level radioactive waste) will be transferred into that company. It is planned to incorporate the German company for construction and management of waste disposal facilities (DBE) which is owned by the German utilities as well. The BfS will focus on the scientific foundations and further advancement of radiation protection. (authors)

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DOE OSTI2017

Thermal Analysis of a Pipe Over-pack Container with a Combustible Waste Matrix for Los Alamos National Laboratory - 17307

Los Alamos National Laboratory (LANL) Technical Area 55 (TA 55) utilizes several different container types to store and transfer special nuclear material and waste for numerous programs. The Nuclear Process Infrastructure - Infrastructure Operations Group (NPI-2) analyzes and certifies containers to meet the specifications for distinct processes and storage. Engineering analysis performed on these containers include but are not limited to: leak testing, filter efficiency testing, pressure drop test, drop testing, water penetration testing, polymer (o-ring) hardness testing, and thermal steady state testing. Of all of these tests, thermal steady-state testing is most crucial to ensure critical components for specific containers do not fail under an internal heat payload. This testing provides data for setting heat payload limits for different containers. The Pipe Over-pack Container (POC) is a vented carbon steel container with a removable lid designed for storing or transferring nuclear waste. Additionally, TA 55 has been tasked to manage and store all of its Transuranic (TRU) waste for the foreseeable future. Being able to place more material into a single container conserves the physical storage space available at TA 55. The Pipe Over pack Handling and Operations Manual requires each user to ensure that the maximum temperature of each component in the POC is not exceeded. Conducting a thermal steady-state test with a combustible waste matrix on the POC can benefit stakeholders on increasing the material limits of the container. The results of these measurements will establish new wattage limits to meet LANL's transportation requirements. The current LANL Transportation Safety Document (TSD) limits the amount of heat source plutonium to 10 grams per POC, corresponding to approximately 5 Watts of heat. The result of this test was provided to the Packaging and Transportation (OS-PT) group to evaluate new wattage limits that can satisfy the needs of TA 55, while maintaining a safe packaging configuration. (authors)

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DOE OSTI2016

Brine Transport Experiments in Granular Salt

To gain confidence in the predictive capability of numerical models, experimental validation must be performed to ensure that parameters and processes are correctly simulated. The laboratory investigations presented herein aim to address knowledge gaps for heat-generating nuclear waste (HGNW) disposal in bedded salt that remain after examination of prior field and laboratory test data. Primarily, we are interested in better constraining the thermal, hydrological, and physicochemical behavior of brine, water vapor, and salt when moist salt is heated. The target of this work is to use run-of-mine (RoM) salt; however during FY2015 progress was made using high-purity, granular sodium chloride.

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DOE OSTI2013

Design of the Long-term Waste Management Facility for Historic LLRW Port Hope Project - 13322

The Municipality of Port Hope is located on the northern shores of Lake Ontario approximately 100 km east of Toronto, Ontario, Canada. Starting in the 1930's, radium and later uranium processing by Eldorado Gold Mines Limited (subsequently Eldorado Nuclear Limited) (Eldorado) at their refinery in Port Hope resulted in the generation of process residues and wastes that were disposed of indiscriminately throughout the Municipality until about the mid-1950's. These process residues contained radium (Ra- 226), uranium, arsenic and other contaminants. Between 1944 and 1988, Eldorado was a Federal Crown Corporation, and as such, the Canadian Federal Government has assumed responsibility for the clean-up and long-term management of the historic waste produced by Eldorado during this period. The Port Hope Project involves the construction and development of a new long-term waste management facility (LTWMF), and the remediation and transfer of the historic wastes located within the Municipality of Port Hope to the new LTWMF. The new LTWMF will consist of an engineered above-ground containment mound designed to contain and isolate the wastes from the surrounding environment for the next several hundred years. The design of the engineered containment mound consists of a primary and secondary composite base liner system and composite final cover system, made up of both natural materials (e.g., compacted clay, granular materials) and synthetic materials (e.g., geo-synthetic clay liner, geo-membrane, geo-textiles). The engineered containment mound will cover an area of approximately 13 hectares and will contain the estimated 1.2 million cubic metres of waste that will be generated from the remedial activities within Port Hope. The LTWMF will also include infrastructure and support facilities such as access roads, administrative offices, laboratory, equipment and personnel decontamination facilities, waste water treatment plant and other ancillary facilities. Preliminary construction activities for the Port Hope LTWMF commenced in 2012 and are scheduled to continue over the next few years. The first cell of the engineered containment mound is scheduled to be constructed in 2015 with waste placement into the Port Hope LTWMF anticipated over the following seven year period. (authors)

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DOE OSTI2012

HLW Return from France to Germany - 15 Years of Experience in Public Acceptance and Technical Aspects - 12149

Since in 1984 the national reprocessing concept was abandoned the reprocessing abroad was the only existing disposal route until 1994. With the amendment of the Atomic Energy Act in 2001 spent fuel management changed completely since from 1 June 2005 any delivery of spent fuel to reprocessing plants was prohibited and the direct disposal of spent fuel became mandatory. Until 2005 the total amount of spent fuel to be reprocessed abroad added up to 6080 t HM, 5309 t HM thereof in France. The waste generated from reprocessing - alternatively an equivalent amount of radioactive material - has to be returned to the country of origin according to the commercial contracts signed between the German utilities and COGEMA, now AREVA NC, in France and BNFL, now INS in UK. In addition the German and the French government exchanged notes with the obligation of both sides to enable and support the return of reprocessing residues or equivalents to Germany. The return of high active vitrified waste from La Hague to the interim storage facility at Gorleben was demanding from the technical view i. e. the cask design and the transport. Unfortunately the Gorleben area served as a target for nuclear opponents from the first transport in 1996 to the latest one in 2011. The protection against sabotage of the railway lines and mass protests needed highly improved security measures. In France and Germany special working forces and projects have been set up to cope with this extraordinary situation. A complex transport organization was established to involve all parties in line with the German and French requirements during transport. The last transport of vitrified residues from France has been completed successfully so far thus confirming the efficiency of the applied measures. Over 15 years there was and still is worldwide no comparable situation it is still unique. Summing up, the exceptional project handling challenge that resulted from the continuous anti-nuclear civil disobedience in Germany over the whole 15-year long project running time could be faced efficiently. It has to be concluded that despite of all problems the anti-nuclear activities have caused so far, all transports of vitrified HLW have always been completed successfully by adapting the commonly established safety, security and public acceptance measures to the special conditions and needs in Germany and coordinating the activities of all parties involved but at the expense of high costs for industry and government and a challenging operational complexity. Apart from an anticipatory project planning a good communication between all involved industrial parties and the French and the German government was the key to the effective management of such shipments and to minimize the radiological, economic, environmental, public and political impact. The future will show how efficiently the gained experience can be used for further return projects which are to be realized since no reprocessed waste has yet been returned from UK and neither the medium-level nor the low-level radioactive waste has been transferred from France to Germany. (author)

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DOE OSTI2015

Non-Invasive pipeline unplugging technology for Hanford high-level waste asynchronous pulsing system - 15706

With the plugging of pipelines obstructing the transfer of high-level waste (HLW) from single shell tanks (SSTs) to double shell tanks (DSTs), effective unplugging technology is prudent. Commercial techniques utilize invasive methods that lead to contamination and subsequent clean-up. FIU's Applied Research Center has developed a non-invasive unplugging technology that can prove advantageous in the transfer of high-level waste, the Asynchronous Pulsing System (APS). This is based on the principle of utilizing asynchronous pressure waves on either end of the plug in order to clear the pipeline blockage. (authors)

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DOE OSTI2020

Reduced Neutralization Feasibility Study for H-Canyon Accelerated Basin De- Inventory (ABD) Program

An alternative approach to Spent Nuclear Fuel (SNF) and Nuclear Material Processing was developed for future H-Canyon (HCAN) and L-Area operations that involves a paradigm shift from current HCAN, Concentrate, Storage, and Transfer Facility (CSTF), and Defense Waste Processing Facility (DWPF) operations. The alternative, referred to as Accelerated Basin De-inventory (ABD), requires that all Domestic and Foreign Research Reactor SNF currently at the Savannah River Site (SRS) will be dissolved, stored, and then transferred to CSTF without recovery of Highly Enriched Uranium (HEU). Concentrated nitric acid is utilized to dissolve aluminum spent nuclear fuel (ASNF) in HCAN. The vessels and piping in HCAN are fabricated from 304L stainless steel and are ideally suited to handle the acidic waste stream. However, as the waste is transferred to the CSTF and DWPF, it will contact the carbon steel waste tanks in CSTF. In order to prevent corrosion of the carbon steel, the acidic waste is neutralized (i.e., pH adjusted over 11) by the addition of sodium hydroxide (NaOH). The NaOH is added until the final solution contains 1.2 M excess -OH. Additionally, if the waste is not neutralized to a pH greater than 11, then aluminum hydroxide (Al(OH) 3 ) would form, and solids may form in the piping as it is transferred to CSTF. This document presents an analysis of the influence of reducing the excess caustic that is added to the neutralization tanks on the corrosion protection scheme primarily for the CSTF waste tanks. The implications to HCAN and DWPF were also assessed.

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DOE OSTI2025

Geographic, sectoral, and constituent characteristics of US off‐site manufacturing wastewater disposal

This study employs recent US Environmental Protection Agency off-site manufacturing wastewater disposal data and a socioenvironmental assessment tool for the United States to understand the characteristics of such disposal in terms of geography, major contributing sectors and pollutants, and environmental impacts. Environmental impact analyses of manufacturing are insufficient without encompassing the extended physical boundary of waste management. Our analysis reveals that off-site manufacturing wastewater disposal occurs disproportionately in populations residing in census tracts already negatively impacted by environmental hazards (impacted populations, or IPs) with 44% of transfers and 55% of Risk Screening Environmental Indicators (RSEI) Hazard going to these areas, compared to their national share of 37%. Disposal hazard is concentrated in a small number of populations, with a Gini coefficient of 0.99 for RSEI Hazard. Four manufacturing sub-sectors are significant generators: Chemicals, Fabricated Metals, Primary Metals, and Transportation Equipment, with Chemicals off-site wastewater disposal largely occurring in IPs. For individual contaminants, chromium compounds and chromium represent more than 85% of the hazard but less than 10% of transfers. We explore transfer distances and waste generation and disposal hotspots, finding that the Midwest hosts a disproportionate share of off-site wastewater disposal. Further, RSEI Hazard steeply rises at shorter distances and plateaus over distances >500 miles, revealing opportunities to reduce hazard by reducing 20–500-mile transfers. Our findings strongly support targeted mitigation strategies like process substitutions, control technologies, on-site recycling and treatment, and minimizing transfer distances. This article met the requirements for a gold-gold JIE data openness badge described at http://jie.click/badges.

Fuchs, Heidi [Lawrence Berkeley National Laborator↗

DOE OSTI2021

Results for the First Quarter Calendar Year 2021 Tank 50 Salt Solution Sample

Tank 50 aqueous waste is analyzed on a quarterly basis and the results are compared to the Waste Acceptance Criteria (WAC) of the Z-Area Saltstone Production Facility (SPF). The information from this characterization will be used by Savannah River Remediation (SRR) for the transfer of aqueous waste from Tank 50 to SPF, where the waste will be treated and disposed in the Saltstone Disposal Facility. This Technical Report compares results, where applicable, to SPF WAC Limits and Targets. A memorandum reporting the average Cs-137 value and comparison to WAC Limits has been previously issued.

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