APPENDIX C

HIGH-LEVEL WASTE AT THE SAVANNAH RIVER SITE

During and immediately following the Second World War, the U.S. Government established large industrial complexes at several sites across the United States to develop, manufacture, and test nuclear weapons. One of these complexes was established in 1950 at the Savannah River Site (SRS) to produce isotopes, mainly plutonium and tritium, for defense purposes. The site is located adjacent to the Savannah River near the Georgia-South Carolina border and the city of Augusta, Georgia, and comprises an area of about 800 square kilometers (~300 square miles).

The SRS was host to an extensive complex of facilities that included fuel and target fabrication plants, nuclear reactors, chemical processing plants, underground storage tanks, and waste processing and immobilization facilities. Plutonium and tritium were produced by irradiating specially prepared metal targets in the nuclear reactors at the site. After irradiation, the targets were transferred to canyon facilities, where they were processed chemically to recover these radionuclides. This processing resulted in the production of large amounts of highly radioactive liquid waste, known as high-level waste (HLW), that, after treatment with caustic, is being stored in two underground tank farms at the site.

TANK WASTE PROCESSING

DOE has the responsibility for waste management at SRS and has implemented a program to stabilize this HLW and close the tank farms. The tank waste processing system at SRS comprises the major components; (a) waste concentration and storage, (b) radionuclide immobilization, (c) extended sludge processing, (d) salt processing, and (e) salt disposal.

Waste Concentration and Storage

The high-level waste resulting from operations in the chemical processing canyons is currently being stored in 48 underground carbon-steel tanks. The tanks range in size from about 3 million to 5 million liters (750,000 to 1.3 million gallons). The HLW was made alkaline with sodium hydroxide (NaOH) and formed a caustic sludge before being transferred to the tanks to reduce corrosion of the carbon steel primary containment. Consequently, the waste has a high pH (>14) and a high salt (especially sodium) content.

Approximately 400 million liters (100 million gallons) of HLW were produced at SRS since operations began in the 1950s, but this volume has been reduced to about 130 million liters (34 million gallons) by removal of excess water through evaporator processing operations. About 10 percent of the waste by volume is in the form of a water-insoluble precipitate, or sludge, that contains most of the actinides (i.e., uranium as well as transuranic elements) and strontium-90. This sludge was formed by natural settling and by precipitation when NaOH was added to the waste. The remaining waste consists of solid sodium salts (saltcake) and an aqueous solution (saturated with sodium salts) called supernate (which contains approximately 95 percent of the cesium in the tank waste, as well as minor amounts of actinides). The saltcake, produced by crystallization after the alkaline waste was processed through evaporators to reduce the volume of material, will dissolve when additional water is added during waste processing. The saltcake



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EVALUATION OF CRITERIA FOR SELECTING A SALT PROCESSING ALTERNATIVE FOR HIGH-LEVEL WASTE AT THE SAVANNAH RIVER SITE: INTERIM REPORT APPENDIX C HIGH-LEVEL WASTE AT THE SAVANNAH RIVER SITE During and immediately following the Second World War, the U.S. Government established large industrial complexes at several sites across the United States to develop, manufacture, and test nuclear weapons. One of these complexes was established in 1950 at the Savannah River Site (SRS) to produce isotopes, mainly plutonium and tritium, for defense purposes. The site is located adjacent to the Savannah River near the Georgia-South Carolina border and the city of Augusta, Georgia, and comprises an area of about 800 square kilometers (~300 square miles). The SRS was host to an extensive complex of facilities that included fuel and target fabrication plants, nuclear reactors, chemical processing plants, underground storage tanks, and waste processing and immobilization facilities. Plutonium and tritium were produced by irradiating specially prepared metal targets in the nuclear reactors at the site. After irradiation, the targets were transferred to canyon facilities, where they were processed chemically to recover these radionuclides. This processing resulted in the production of large amounts of highly radioactive liquid waste, known as high-level waste (HLW), that, after treatment with caustic, is being stored in two underground tank farms at the site. TANK WASTE PROCESSING DOE has the responsibility for waste management at SRS and has implemented a program to stabilize this HLW and close the tank farms. The tank waste processing system at SRS comprises the major components; (a) waste concentration and storage, (b) radionuclide immobilization, (c) extended sludge processing, (d) salt processing, and (e) salt disposal. Waste Concentration and Storage The high-level waste resulting from operations in the chemical processing canyons is currently being stored in 48 underground carbon-steel tanks. The tanks range in size from about 3 million to 5 million liters (750,000 to 1.3 million gallons). The HLW was made alkaline with sodium hydroxide (NaOH) and formed a caustic sludge before being transferred to the tanks to reduce corrosion of the carbon steel primary containment. Consequently, the waste has a high pH (>14) and a high salt (especially sodium) content. Approximately 400 million liters (100 million gallons) of HLW were produced at SRS since operations began in the 1950s, but this volume has been reduced to about 130 million liters (34 million gallons) by removal of excess water through evaporator processing operations. About 10 percent of the waste by volume is in the form of a water-insoluble precipitate, or sludge, that contains most of the actinides (i.e., uranium as well as transuranic elements) and strontium-90. This sludge was formed by natural settling and by precipitation when NaOH was added to the waste. The remaining waste consists of solid sodium salts (saltcake) and an aqueous solution (saturated with sodium salts) called supernate (which contains approximately 95 percent of the cesium in the tank waste, as well as minor amounts of actinides). The saltcake, produced by crystallization after the alkaline waste was processed through evaporators to reduce the volume of material, will dissolve when additional water is added during waste processing. The saltcake

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EVALUATION OF CRITERIA FOR SELECTING A SALT PROCESSING ALTERNATIVE FOR HIGH-LEVEL WASTE AT THE SAVANNAH RIVER SITE: INTERIM REPORT and sludge contain substantial quantities of supernate within their mass; this interstitial supernate corresponds to about half of the total supernate in the tanks. Radionuclide Immobilization The Defense Waste Processing Facility (DWPF) was constructed to immobilize radioactive waste in borosilicate glass for eventual shipment to and disposal in a geological repository. The glass-making process is referred to as vitrification. This glass is produced by combining the processed HLW (the processing operations are discussed below) with specially formulated glass frit and melting the mixture at about 1150 °C. The molten glass is then poured into cylindrical stainless steel canisters, allowed to cool, and sealed. The DWPF canisters are about 60 centimeters (2 feet) in diameter and about 300 centimeters (10 feet) in length and contain about 1,800 kilograms (4,000 pounds) of glass. About 700 canisters have been produced to date4, and SRS estimates that a total of about 6,000 canisters would be produced by 2026, when the tank waste processing program is planned to be completed. These canisters are to be stored at the site until a permanent geological repository is opened and ready to receive them. Extended Sludge Processing Extended sludge processing is being used to prepare the sludge portion of the tank waste for processing into glass. The sludge is removed from the tanks by hydraulic slurrying and washed to remove aluminum and soluble salts, both of which can interfere with the glass-making process. The washed sludge is transferred to the DWPF for further processing before being incorporated into glass. Sludge processing would result in immobilization in glass of nearly all of the strontium and actinides from the tanks. Salt Processing Salt processing would be used to remove much of the radionuclides from the HLW salt for eventual vitrification. The salt is to be redissolved and transferred out of the tanks. It would then be mixed with a sorbent to remove any remaining actinides (mainly uranium and plutonium) and strontium. The currently planned sorbent is monosodium titanate (MST). The solution will then be subjected to another (and as-yet undetermined) process to remove cesium. The separated actinides, strontium, and cesium would be washed to remove soluble salts and sent to the DWPF for immobilization. Salt Disposal. A variety of secondary waste streams are formed during the processing operations described above. Some of these waste streams are recycled back to the tanks, some are recycled within the various processing operations, and yet other wastes are treated and stabilized for burial. Most notably, the “decontaminated” salt supernate (i.e., the solutions remaining after actinide, strontium, and cesium removal) would be disposed of onsite in a waste form known as Saltstone. The residual solutions are classified as “incidental waste” from the processing of HLW. Saltstone is created by mixing the residual salt solutions with fly ash, slag, and Portland 4   Since this appendix was originally published, over 300 additional canisters have been produced.

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EVALUATION OF CRITERIA FOR SELECTING A SALT PROCESSING ALTERNATIVE FOR HIGH-LEVEL WASTE AT THE SAVANNAH RIVER SITE: INTERIM REPORT cement to create a grout slurry. This slurry is then poured into concrete vaults, where it cures (solidifies) and is eventually covered with soil. The Saltstone contains small quantities of some radionuclides. CESIUM REMOVAL PROBLEM As noted above, SRS planned to remove actinides, strontium, and cesium from the salt solutions in two processing steps. First, actinides and strontium were to be removed by mixing the salt solutions with MST, resulting in the sorption of actinides and strontium. The product of this reaction could be removed from the salt solutions by filtration for subsequent processing and immobilization. Subsequently, the removal of cesium from the salt solutions would be accomplished by a yet-to-be-chosen process from among precipitation, ion exchange or solvent extraction processes. In the late 1970s and the 1980s, SRS developed a process for removing cesium from salt solutions through a precipitation reaction involving sodium tetraphenylborate (NaTPB) and cesium to form cesium TPB (CsTPB): SRS refers to this process as “In-Tank Precipitation.” The NaTPB was to be added directly to a large waste tank to produce a cesium-bearing precipitate. SRS undertook an ITP pilot project in 1983 to demonstrate proof of principle. The process removed cesium from the salt solution, but it also resulted in the generation of flammable benzene from radiolytic reactions and possibly from catalytic reactions with trace metals in the waste. In September 1995, SRS initiated ITP processing operations in a tank that contained about 1.7 million liters (450,000 gallons) of salt solutions. The operations were halted after about 3 months because of higher-than-expected rates of benzene generation. SRS staff then initiated a research program to develop a better understanding of the mechanisms of benzene generation and release. They also considered possible design changes to handle the benzene during processing operations and catalyst poisoning strategies. In 1996, the Defense Nuclear Facility Safety Board (DFNSB) issued Recommendation 96-1, urging DOE to halt all further testing and to begin an investigative effort to understand the mechanisms of benzene formation and release. Investigations by SRS in 1997 uncovered the possible role of metal catalysts in the benzene formation process. SRS concluded, however, that both safety and production requirements could not be met, which led to the suspension of operations altogether in early 1998. At the time of suspension, SRS had spent almost a half billion dollars to develop and implement the ITP process. In March 1998, Westinghouse Savannah River Company (WSRC) formed a systems engineering team to identify alternatives to the ITP process for separating cesium. This team began by undertaking a literature and patent screening procedure to identify currently known processes, followed by a system of analyses by panels of experts to reduce the number of alternative processes to four.Strontium/Actinide Removal by MST In all four of the final candidate processes for cesium separation, prior removal of strontium and actinides is viewed by SRS as a requisite process. At present, the use of MST is the method of choice. Some technical uncertainties remain to be resolved, of which the major ones are the kinetics of sorption on MST and the amount of titanate acceptable for proper quality of the vitrified waste form.

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EVALUATION OF CRITERIA FOR SELECTING A SALT PROCESSING ALTERNATIVE FOR HIGH-LEVEL WASTE AT THE SAVANNAH RIVER SITE: INTERIM REPORT Tetraphenylborate Precipitation Process The ITP developed by WSRC removes cesium from HLW supernates by precipitation with tetraphenylborate ion, [B(C6H5)4]- (TPB). Sodium TPB is a reagent used for analyzing for the potassium ion based on the insolubility of potassium TPB (KTPB). The 200-fold lower solubility of cesium TPB (CsTPB) can provide decontamination factors (DF) from the salt as high as 105 to 106and the mixed CsTPB/KTPB precipitate is typically in a form that is easily filtered. On the average, the SRS HLW in the waste tanks contains sodium ions (approximately 5 molar), potassium ions (approximately 0.03 molar), and cesium ions (approximately 0.00025 molar). HLW treatment, including the removal of cesium-137, involves separation of selected radioactive components and their subsequent immobilization in a borosilicate glass at the DWPF. To prevent organic material from being fed to the DWPF melters, the CsTPB/KTPB precipitate must be treated to remove more than 90 percent of the phenyl (C6H5) groups bound to the boron. Thus, a precipitate hydrolysis process (PHP) was developed to hydrolyze the TPB using formic acid in the presence of a copper catalyst. The hydrolysis products are benzene, which is removed by evaporation and incineration, and an aqueous solution containing137Cs+, B(OH)3, and K+ ions. An attractive feature of TPB is its susceptibility to catalytic decomposition downstream. Crystalline Silicotitanate Ion Exchange Ion exchange has been in commercial use for over 100 years to remove ions from aqueous solutions, e.g., to make deionized water. In most applications the separated ions are eluted from the ion exchange material, e.g., using a dilute acid, the eluted ions are concentrated, and the ion exchanger is reused over and over. Although this technology is well established, ion exchange for cesium removal from high-level waste at SRS and other DOE sites poses challenges. The ion exchange material must withstand both high alkalinity and high radiation fields and must be very selective for cesium in the presence of much higher concentrations of the chemically related sodium and potassium ions. A promising material for use by SRS to remove cesium is crystalline silicotitanate (CST), developed by Sandia National Laboratory and Texas A&M University, based on work performed on amorphous hydrous titanium oxide in the 1960s and 1970s at Sandia. CST has received considerable attention because of its promise as an ion exchange material for nuclear waste applications. The material has a high selectivity for Cs+in salt solutions over a large portion of the pH range from acidic to basic solution, and exhibits high stability to radiation as well. CST is also unusual in that cesium is difficult to remove from the material (i.e., it is nonelutable and the CST cannot be reused). As a result, CST must be incorporated into the HLW stream along with the radionuclides, and the stability of borosilicate glass with higher concentrations of titanium is an issue that must be addressed. Caustic Side Solvent Extraction A typical solvent extraction process includes four steps. First, a feed stream is contacted with a solvent that is virtually insoluble in the stream. During this contact, one or more components of the stream transfer to the solvent, while other components do not. The loaded solvent, scrubbed to remove minor contaminants and leaving relatively clean solvent plus the

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EVALUATION OF CRITERIA FOR SELECTING A SALT PROCESSING ALTERNATIVE FOR HIGH-LEVEL WASTE AT THE SAVANNAH RIVER SITE: INTERIM REPORT component(s) to be finally recovered, is sent to a stripping operation where the component(s) to be recovered is removed. The stripped solvent may then go to a solvent-recovery step, in which it is cleaned prior to returning to the first step. In such a process, very high removals of extracted components often can be attained. Solvent extraction has had a long history of successful use in the nuclear industry for such operations as spent fuel reprocessing and plutonium recovery. This history includes long periods of time in which solvents of various organic species have been exposed to high-radiation fields without experiencing catastrophic degradation rates. Solvent extraction operations usually consist of selectively transferring components from an aqueous, acidic stream into the organic stream. A second aqueous stream of somewhat different composition is often used to strip the solvent and concentrate the extract. For the SRS application, the solvent extraction process must remove approximately 99.998 percent of the cesium (a decontamination factor, or DF, of 50,000) from an aqueous, tank-waste feed stream. The raffinate aqueous stream, thus purified of cesium, would be sent to the SRS Saltstone Facility, and the extract, concentrated in cesium by about an order of magnitude is sent to the DWPF. Direct Disposal in Grout Direct disposal of the tank waste following removal of strontium and actinides is very similar to the Saltstone process that was to have been used to dispose of the salt solutions from ITP operations as low-level incidental waste. Although it is a rather mature technology and has already been demonstrated at the site for less radioactive salt solutions, the degree of retention of cesium may not satisfy regulatory requirements.