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Overview of HEW Challenges at DOE Sites The DOE has approxi mately 400 mi l l ion l iters (1 00 mi l l ion gal ions) of I iquid H LW stored in underground tanks and approximately 4,000 cubic meters of solid HLW stored in bins. The current DOE estimate of the cost of converting these liquid and solid wastes into stable forms for shipment to a geological repository exceeds $50 billion to be spent over several decades (DOE, 2000~. The most important HLW sites in the United States are the follow- ng: · the Hanford Site, near Richland, Washington, · the Idaho National Engineering and Environmental Laboratory (I N EEL), near Idaho Fal Is, · the Savannah River Site (SRS), near Aiken, South Carolina, and · the WestValley Demonstration Project (WVDP), near Buffalo, New York. All of the HLW sites listed above are briefly described later in this chapter. Sidebar 2.1 shows a simplified diagram of one of the processes that produced HLW during chemical recovery of plutonium and uranium from spent fuel and target materials for the production of nuclear weapons. Most of the HLW is a multiphase mixture of solids and liquids stored in underground tanks. Some has been retrieved and solidified and is stored in underground bins or surface vaults. Processing and handling of these HLW forms generate other radioactive wastes man- aged as transuranic or low-level waste (LLW, see Sidebar 2.2), depend- ing on their characteristics. The reclassification of HLW residues to LLW or transuranic takes place according to DOE's procedures to determine O v e r v i e w o f H L W C h a I I e n 9 e s a t D O E S i t e s
SIDEBAR 2.1 THE PLUTONIUM AND URANIUM EXTRACTION (PUREX) PROCESS _m.~ Ed | [d |~ |~] | ~~ x. 0~ ~ ~ ~ ~ ~ ~~ ~~ em. ~F ~ ~ ~ ~ I:. O~s atop _ ~ _ . :: _ ~ _E1~1 ~ : . . my ,... ^.. 11. it [~ It|| jiTd{~ ~11 iatlel~.~. 1~ - Nll.iTd{~ mill i.~I.I} nI [I .~ UO 2+X : to hi 0~32 ~ ._ lit The PUREX process, shown above, purified and separated both uranium and plutonium from defense spent nuclear fuel. It was first used at the Hanford Site in 1952 and subsequently at the other HEW sites.The first step consists of removing the cladding of spent fuel (fuel preparation) to expose fuel rods. In the second step, spent fuel is dissolved in nitric acid.The third and key step is the separation and recovery of uranium and plutonium from the fission products. It is done in a continuous counter- current solvent extraction process using tri-N-butyl phosphate in a hydrocarbon diluent (such as kerosene). Plutonium and uranium are extracted in the organic solvent while fission products and other impurities remain dissolved in the highly acidic aqueous nitric solution stored in tanks and clas- sified as HLW.The fourth step consists of separating plutonium from uranium by reducing plutonium to the organic-insoluble, trivalent state while uranium remains in the organic phase in the oxidized form. Both plutonium and uranium are then converted to solid oxide or metal. Except at the INEEL, the H ~ G H - L E V E E W A S T E 14
waste was neutralized with sodium hydroxide prior to storage in the tank farms. A major advantage of the PUREX process over the earlier processes used (bismuth phosphate and REDOX processes, see the glossary in Appendix G) is that nitric acid is the only salting agent used in the aqueous phase; there- fore HLW has a lower salt content than waste from other processes. Other advantages of PUREX are reduction of the waste volume, greater flexibility in process conditions, reduction of fire and other haz- ards, and reduction of costs. Processes based on PUREX have been adopted for nearly all fuel repro- cessing throughout the world. SOURCE: Adapted from Benedict et al. (1981 ) and from JEER (1996). whether the waste can be treated as waste incidental to reprocessing (WIR) described in Chapter 6, Sidebar 6.1. The committee has not con- sidered issues with these other wastes in order to focus its attention on the retrieval and processing of HLW from tanks and bins. HLW Management Strategy at DOE Sites The current DOE plans to treat and dispose of HLW face many tech- nical uncertainties. Many of the planned treatment activities are first-of- a-kind efforts involving highly radioactive liquid and solid wastes. Moreover, the waste streams are fundamentally different in character, both among the HLW sites and within the sites, because waste streams were generated from processing different kinds of nuclear fuel and were treated and stored differently. Consequently, each HLW stream presents different challenges and may require specific processing and immobi- lization strategies. However, the overall HLW management strategy applied by DOE is the same throughout all HLW sites. The generic HLW remediation flow sheet is shown i n Figu re 2.1 and comprises the "backbone" of th is report. This flow sheet applies mainly to HLW in the tanks. In the case of HLW stored at the INEEL in a calcined form, some details are slightly different, as described later in this chapter. The HLW management strat- egy consists of different steps, described as follows. Characterization of waste in the tanks is performed to ensure the continued safe storage of the waste and to facilitate the following HLW management operations. At a given point, HLW must be retrieved from the storage tanks and bins and processed for ultimate disposal in a repository. The process of removing waste from the tanks and transport- ing it to the processing facil ities is Cal led retrieval. After retrieval, the HLW proceeds to the pretreatment phase, to separate constituents that O v e r v i e w o f H L W C h a I I e n 9 e s a t D O E S i t e s
SIDEBAR 2.2 NUCLEAR AND RADIOACTIVE WASTE TERMS A. High-level waste: The radioactive by-product generated from processing irradiated fuel to sepa- rate usable plutonium and other isotopes for weapons, research, and new fuel; and irradiated fuel itself if destined for direct disposal. B. Transuranic waste: Waste that contains alpha-emitting transuranic elements with half-lives greater than 20 years whose combined activity level is at least 100 nanocuries (3,700 becquerels) per gram of waste at the time of assay. C. Low-level waste: All radioactive waste not classified as HLW, transuranic waste, spent nuclear fuel, or the by-product material (see definition F.3.b below) generated by the processing of ores for extraction of natural uranium or thorium. D. Mixed waste: Waste that contains both radioactive material regulated under the Atomic Energy Act and hazardous chemical compounds regulated under the Resource Conservation and Recovery Act, such as mercury, polychlorinated biphenyls, or organic solvents. Spent nuclear fuel: Nuclear fuel that has been withdrawn from a nuclear reactor following irradi- ation, has undergone at least one year's decay since being used as a source of energy in a power reactor, and has not been chemically separated into its constituent elements by reprocessing. Spent nuclear fuel, through fission and neutron activation and decay, contains multiple radioac- tive elements with varying chemical and radiological properties. Spent fuel includes the special nuclear material, byproduct material, source material, and other radioactive materials associated with fuel assemblies. Nuclear material (Atomic Energy Act material): Nuclear material generally includes (1 ) source material, natural uranium, thorium, or any other material that, alone or in combination, is deter- mined to be a potential fuel for a fission reactor; (2) special nuclear material, fissile material such as uranium enriched in uranium-235, or plutonium that can undergo fission; and (3) byproduct material, meaning (a) any radioactive material (except special nuclear material) yielded in or made radioactive by exposure to the radiation incident to the process of producing or utilizing special nuclear material, and (b) the tailings or wastes produced by the extraction or concentra- tion of uranium or thorium from any material processed primarily for its source material content. G. Radioactive material: Radioactive material includes all of the nuclear material defined above. Radioactive material also includes naturally occurring radioactive elements such as isotopes of radium as well as material made radioactive by processes other than nuclear fission. SOURCE: Adapted from DOE (2000). are radioactive, hazardous, or detrimental tothe immobilization step from the bulk of the waste. After pretreatment, HLW is ready for immo- bilization in a borosilicate glass matrix. This process is called "vitrifica- tion." The HLW glass canisters produced are stored at the sites until H ~ G H - L E V E E W A S T E 16
High- Level Waste F \ 'I Retrieval Safety | ~: Characterization I Characterization Transport & Pretreatment _1 Immobilization _ Disposal in - - I Repository LLW Stream ]~ ~ Tank Closure ~ Monitoring New Science & Technology Process Path they can be transported to a HLW repository.] Once the waste is removed from the tanks to the maximum extent that is technically and economically practical, the next step is tank closure. The retrieval of HLW must be adequate to ensure that residues can be declared as inci- dental to reprocessing and can be disposed of as LLW (see Sidebar 6.1~. Tank closure consists of characterizing the residues, filling the tanks with cementitious material, and leaving the tanks in place. After closure, monitoring activities must take place to identify any release of residual waste into the environment. HEW Generation and Cleanup Strategies at DOE Sites The first HLW was generated at the Hanford Site from reprocessing reactor fuel for the production of plutonium used in nuclear weapons. The Yucca Mountain site in Nevada is currently under consideration as an HLW repository. O v e r v i e w o f H L W C h a I I e n 9 e s a t D O E S i t e s - FIGURE 2. 1 Generic waste remediation flow sheet used throughout this report. NOTE: LLW = low-level waste
Subsequently, H LW was generated at three other sites, I isted on p. 1 1 . Most of the HLW was produced during World War 11 and the ensuing Cold War.2 The generation of HLW and the cleanup strategy for each site are briefly described in the following sections. The Hanford Site The first recovery of plutonium at the Hanford Site began in late 1944 using the bismuth phosphate separation process (Gephart and Lu ndgren, 1 998~. Th is process recovered pi uton i u m, but not u ran i u m, from the spent fuel and produced large quantities of waste. Later efforts to recover uranium from the bismuth phosphate waste further changed the nature of HLW produced. The higher-activity liquid wastes from bis- muth phosphate reprocessing were neutralized chemically to reduce their corrosiveness and stored in carbon steel underground tanks. The first successful solvent extraction process for plutonium and ura- nium recovery in continuous plant operation was the REDOX (REDuction and OXidation) process, which began at Hanford in 1952. The REDOX process generated a lower volume of waste, which was chemically neutralized and stored in carbon steel underground tanks. In early 1956, a new solvent extraction process, PUREX (Plutonium and Uranium Recovery by Extraction), came into use at Hanford (Sidebar 2.1~. PUREX used a different organic solvent and nitric acid. PUREX wastes were highly radioactive, self-boiling, and were also neutralized and stored in carbon steel underground tanks. Therefore, the tank wastes at Hanford include many different chemical compositions and physical characteristics, since different processes for plutonium recovery (as well as other operations) were developed and applied during the operational I ife of th is site. Table 2.1 presents a si mpl if fed i nventory of the wastes at the Hanford Site.3 Only one percent of the mass of waste material is radioactive, but this is enough to make the tank contents highly danger- ous.4 Most of the radioactive species can be separated and immobilized for disposal as solid HLW; the residues remaining can be treated as other forms of waste. 2The HLW inventory may still grow slightly at the SRS because waste from deactivation and decommissioning operations is stored in HLW tanks. A small quantity of HLW is produced also from the reprocessing of part of damaged research reactor fuel and of the fuel "blanket" from the Experimental Breeder Reactor 11. For more details on research reactor spent fuel, see (NRC, 1998). 3HLW from the other sites has a different and less complex composition com- pared to the Hanford waste. 4The radiation field inside the tanks can be as high as 10,000 red per hour (100 gray per hour). H ~ G H - L E V E E W A S T E
TABLE 2.1 Total Waste inventory Based on Process Know~ecge for Aid Hanforc Tanks. Only some of the radionu- c~ides present are listed, and the inventory does not take into account the Decay in radioactivity since ~ 996. Although water is not listed, it represents a significant fraction of HEW (>200,000,000 kilograms). NOTE: ~ curie 3.7 x 103° becquere~s. = Total Inventory Total activity Half-Life Species (103 kilograms) (103 curies) (years) Na+ Al3+ Fe3+ Cr3+ Bi3+ Zr as ZrO(OH)2 pb2+ Nix+ Ca2+ K+ TOC NO3- NO2- CO32- po43- SO42- Si as SiO32- F- 49,170 8,372 1,872 994 492 214 279 183 620 521 1,818 50,430 1 4,093 4,836 3,795 3,226 662 575 Cl- 1,099 Sr-90/Y-90 0.44 123,203 29.12/(64 hours) Tc-99 1.94 33 21 3,000 Cs-137/Ba-137m 0.54 91,966 30/(2.6 min) Np-237 0.20 0.141 2,140,000 Pu-239 0.77 47 24,1 30 Pu-Total 0.77 225.59 NOTE:TOC = total organic carbon.The half-life of a radionuclide is the time after which half of atoms on average have disinte- grated. SOURCE: Agnew et al., Appendix E (1997). In addition to the evolution of the waste characteristics, the designs of the waste storage tanks evolved as well. The earliest tanks were sin- gle-shell tanks (SSTs) of carbon steel. Reliance on a single shell has resulted in leakage of waste from some of the SSTs into the subsurface soil. Of the 177 tanks at Hanford, 67 SSTs are known or suspected to have leaked. The total quantity of leaked waste is estimated to be between 3.4 and 6.7 million liters containing between 0.45 and 1.8 O v e r v i e w o f H LW C h a I I e n 9 e s a t D O E S i t e s 19
million curies (1.7 x 10~6 and 6.7 x 10~6 becquerels), although there are large uncertainties in these estimates (Waite, 1991; ERDA, 1975; Agnew et al., 1997~. The later double-shell tanks (DSTs) consisted of a first tank surrounded by a secondary containment tank. Figure 2.2 shows a diagram of a SST and a DST. The external shell of the DST pro- vides an additional barrier to leakage of waste into the subsurface soil. There has been some leakage into the annulus space lying between the shell liners of some DSTs. All the free liquid from SSTs has been trans- ferred to DSTs, thereby greatly diminishing the potential impact of future leaks. The risk from leakage in the future is expected to be asso- ciated almost entirely with retrieval of waste from the SSTs, depending on the retrieval method used. The vulnerability to leakage into the sub- surface soil is an important reason for closure of the Hanford tanks. A summary of the key facts and figures about the tank wastes at the Hanford Site is presented in Table 2.2. In its record of decision of 1997, DOE adopted a phased approach to tank waste management at the Hanford Site (DOE, 1997~. Under Phase 1, which will last approximately until 2018, 10 percent of the tank waste and 20 to 25 percent of the radioactivity is slated for retrieval and immobilization on a pilot scale for demonstration purpos- es. After val idation of the processes, DOE wi 11 implement Phase 11, the full-scale production phase, which will last approximately 30 years. F/GURE2.2 Diagram ofa SST Figure 2.3 shows DOE's baseline plan for HEW management in (left) and aDST(right) atthe Hanford. HanfordSite.SOURCE:DOE- So far, the waste has been characterized to the extent feasible for its Tanks Focus Area. heterogeneous contents. Waste retrieval operations have just begun. H ~ G H - L E V E E VV A 5 T E ~ A ~
TABLE 2.2 Facts anc Figures about the Hanforc Site Type of information Data Surface of the site 1,450 square kilometers (560 square miles) Number of HLW tanks 177 Volume of HLW 208 million liters (54 million gallons) Radioactivity About 200 million curies (7.4 x 1048 becquerels) Characteristic of HLW Alkaline waste (pH approximately 12), saltcake, viscous liquid, sludge, and highly non homogeneous waste Types of tanks Carbon steel, SST and DST Waste sluicing by pumping liquid to entrain the sludge is the technique chosen for retrieval of the waste from the tanks. According to the plan, after retrieval, the waste will be sent to a pretreatment facility to sepa- rate the solids, handled as HLW, from the liquid phase, which will be decontaminated further to remove the hazardous constituents.5 The solids-liquid separation (SLS) process is followed by sludge leaching of the separated solids to dissolve non-radioactive components. Cesium, strontium, transuranic elements, and technetium (as necessary) will be removed from the I iqu id phase and sent to the H LW vitrification faci I ity 5Further details on pretreatment and immobilization processes are given in Chapters 4 and 5. - I _| Solids-Liquid I Solid it= | Sluicing . 1.- · Radionuclide Removal Sl edge l l Washing ~ Solid LLW Vitrification = Liquid HLW Vitrification ·h~ HLW Interim Storage O v e r v i e w o f H L W C h a I I e n 9 e s a t D O E S i t e s FIGURE 2.3 Based on infor- mation gathered from DOE, this figure represents a sim- plified view of the baseline plan for waste management at the Hanford Site. Closed Tanks · ·~ _= LLW Disposal Vaults Off-Site Transport
along with the HLW solids. The HLW glass canisters will be stored tem- porarily on site for eventual disposal in a geological repository. According to the 1994 revision of the Tri-Party Agreement (TPA) between DOE, the U.S. Environmental Protection Agency (EPA), and the Washington State Department of Ecology, LLW wil I also be immobi- lized in glass and stored on site (TPA, 1998~. The Idaho National Engineering and Environmental Laboratory Between 1 953 and 1 992, the I N EEL reprocessed spent nuclear fuel mainly for recovery of the fissile isotope uranium-235. In the reprocess- ing operation, spent nuclear fuel and its cladding materials (aluminum, zirconium, stainless steel, and graphite) were dissolved in highly acidic solutions (nitric acid, hydrofluoric acid, and sulfuric acid were used) and Graphite was burned. INEEL's waste type, containing uranic and O , ,, , O . . . . . . . .. . . . . . . transuranic Isotopes generated tram the reprocessing at spent fuel, IS unique within DOE HLW sites in that it was stored while still in highly acidic form. Table 2.3 presents some general information on the wastes at the IN EEL. The majority of INEEL's waste has been calcined to a granular solid (ceramic), which is considered an interim storage form by the State of Idaho. The calcine is stored in partially buried stainless steel bins (designed to last 500 years) grouped inside concrete vaults. Calcine waste can be processed further to convert it into a more con- solidated long-term waste form. In addition, the INEEL has some liquid sodium-bearing waste (SBW) to be immobilized. This waste is mixed transuranic waste but, because of its radioactivity and large volume, it TABLE 2.3 Facts anc Figures about the KNEEL Type of information Data Surface of the site 2,300 square kilometers (890 square miles) Number of HLW containers 1 1 tanks, 7 calcine vaults Volume of HLW About 5.3 million liters (1.4 million gallons) of liquid waste and 3.8 million liters (1 million gallons) of calcine Radioactivity 520,000 curies of radioactivity (1.9 x 1 on becquerels) in liquid form and 24 million of curies (8.9 x 1017 becquerels) of radioactivity in calcined form Characteristic of HLW Very acidic (pH approximately 0), sludges and viscous liquid, calcine solids Types of tanks Stainless steel, single shell H I G H - L E V E L VV A 5 T E ~ A ~
is stored in underground tanks and managed as H LW (N RC, 1999b). DOE has recently decided (Huntoon, 2000) to seek direct vitrification as the preferred alternative for SBW. The preferred alternative for INEEL's calcine is to continue work on characterization, pretreatment, and vitri- fication options to be ready for shipment to a disposal facility by the end of 2035. Immobilized calcine and SEW will be stored temporarily on site awaiting eventual disposal in a geological repository. Figure 2.4 shows the proposed baseline plans for the management of HLW calcine and liquid at INEEL. DOE is still studying this strategy and in 1999 was advised by the NRC on the different possibilities (NRC, 1999b). The Savannah River Site Plutonium and tritium for nuclear weapons were produced at the SRS near Aiken, South Carolina, by reprocessing spent fuel and targets. The PUREX process was used from the time the SRS opened in 1952; as FIGURE 2.4 Simplified base- line plan for the manage- ment of liquid and calcined HLW at INEEL. Question marks represent alternative treatments for calcine. The final decision is still pending in Hanford, acidic H LW generated by PU REX was neutral ized for storage in carbon-steel tanks. SRS also produced plutonium-238 and uranium- (Huntoon,~OOO).Formore processes other than standard PU REX. The details, see NRC (1999b). 233 and recovered them by r HLW Tanks Calcine Bins ? ? Evaporator Liquid Dissolver Vitrification Facility Grout Plant LLW Disposal (On-Site or Off-Site) HLW Interim Storage O v e r v i e w o f H L W C h a I I e n 9 e s a t D O E S i t e s Off-Site Transport
waste tanks contain HLW in the form of sludge, concentrated super- nate, and saltcake. The sludge, of thin peanut butter-like consistency, contains 67 percent of the radioactivity and represents 9 volume per- cent; the supernate and the saltcake represent 91 volume percent and 3 3 percent of the red inactivity. HLW at the S RS is stored i n 51 tan ks, 2 of which have already been closed. Table 2.4 presents some of the prin- cipal data on the tank wastes at the SRS. Figure 2.5 shows the baseline pi an for waste management at th is site. Radioactive vitrification operations at the SRS began in 1996, and approximately 1,000 HLW glass logs have been produced to date. The sludge is retrieved from different tanks and mixed in pretreatment batches to lower (blend down) the concentration of components that might add to the complexity of the vitrification step (more details are given in Chapter 5~. The sludge is also pretreated by caustic extended sludge washing to remove aluminum from HLW. Aluminum is known to increase the viscosity of the molten glass, which complicates pouring the glass melt into the canisters and also considerably increases the total volume of immobilized HLW. The SRS is still in the process of selecting a preferred method to separate cesium, strontium, and actinides from the saltcake and supernate, after the previous process (large-scale in-tank precipitation) was abandoned because of technical difficulties (NRC, 2000b). An NRC panel is currently advising DOE on viable alternatives (NRC, 2001 a). The leached sludge solids are then combined with cesium, strontium, technetium, and transuranic ele- ments from the liquid decontamination step, and sent to the vitrification pi ant, the Defense Waste Processi ng Faci I ity (DWPF). The decontam i- nated LLW filtrate is then grouted (i.e., immobilized by cementation). The HLW glass canisters produced are temporarily stored on site wait- ing for eventual disposal in a geological repository. TABLE 2.4 Facts anc Figures about the SRS Type of Information Data Surface of the site 800 square kilometers (300 square miles) Number of HLW tanks 51 (2 tanks already closed) Volume 125 million liters (33.4 million gallons) Radioactivity 240 million curies (8.9 x 1048 becquerels) in the sludge and 180 million curies (4.4 x 1049 becquerels) distributed between the supernate and the saltcake Characteristic of HLW Alkaline (pH approximately 14) Types of tanks The carbon steel tanks are differentiated into Type I, II, III, and IV tanks, according to their characteristics. Of the 51, only 27 tanks (type ill) are double shell tanks H I G H - L E V E L VV A s T E ~ A ~ 24
Salt ._.. Sludge l Cs, Sr, and TRU remove I · | Extended Sludge | Slurry Raffinate Precipitate Saltstone Grout Plant LLW Vaults HLW Vitrification HLW Interim Storage HLW Glass Logs 1 Failed Melters The West Valley Demonstration Project The WVDP is situated near Buffalo, New York. The State of New York owns this site, not DOE. However, by the terms of the WVDP Act of 1980, DOE is responsible for retrieving the HLW on site, removing it for disposal, and decommissioning all parts of the site used for defense- related tasks. The U.S. Congress appropriates funds to DOE for this work, and New York State shares the costs. Because the WVDP has almost completed its HLW cleanup operations, the recommendations in this report are not as relevant to this site as they are for the Hanford, Savannah River, and INEEL sites. However, the committee gathered information on the WVDP because of the "operational experience" acqu i red at th is site. From 1966 to 1972, commercial spent nuclear fuel was reprocessed at the WVDP to recover uranium and plutonium by a licensed, com- mercial fuel reprocessing plant. The reprocessed reactor fuel included a substantial quantity of typical light water reactor (LWR) fuel, a single thorium-based core from an LWR, and a substantial amount of low-bur- nup defense spent fuel from the Hanford N Reactor, provided as a base load to support this first commercial reprocessing plant. The N Reactor fuel and the typical LWR fuel were reprocessed to recover uranium and plutonium using the PUREX process; the thorium-based core was reprocessed to recover uranium-233 and uranium-235. The single tank farm at the WVDP contains two carbon steel tanks (each with a capacity of 2.4 million liters), one for storage of alkaline waste and one spare, along with two stainless steel tanks (each with a capacity of 1 10,000 liters), one for storage of the acid waste from the Off-Site Transport FIGURE 2.5 Simplified base- line plans for management of the waste at the SRS. The Defense Waste Processing Facility (DWPF) is the vitrifi- cation plant at the site. The waste supernate is the main source of cesium, strontium, and transuranic elements (Cs, Sr and TRW). O v e r v i e w o f H L W C h a I I e n 9 e s a t D O E S i t e s
TABLE 2.5 Facts anc Figures about the WVDP Type of Information Data Surface of the site Number of HLW tanks Volume of HLW (initial conditions) Radioactivity Characteristic Types of tanks 0.3 square miles (0.8 square kilometers) 2 carbon steel tanks of 2.4 million liters (660,000 gallons), 2 stainless steel tanks of 1 10,000 liters (30,000 gallons) 2.4 million liters (0.66 million gallons), after blending a small amount of acid waste into the alkaline waste 0.6 million curies (2.2 x 1 on becquerels) Alkaline waste (pH approximately 14) Double shell, carbon steel thorium-based core and one spare. Table 2.5 presents some of the prin- cipal data on the tank wastes at the WVDP. All of the waste was blend- ed together in one of the larger storage tanks. All of the tanks were inte- grated into a recycling pretreatment system. As a result, there was a sin- gle waste composition to process. Figure 2.6 shows a simplified baseline plan for the management of HLW at the WVDP. The supernatant liquid was removed and pretreated with zeolite, a granular aluminosilicate ion-exchange material, to remove the radioactive cesium and strontium, yielding a separate LLW salt solution. The sludge remaining at the bottom of the tanks was FIGURE 2.6 Simplified base- washed, and the I iqu ids were treated with zeal ite. The remai n i ng H LW, line plans for the WVDP consisting of the washed sludge and the zeolite used for pretreatment, basedoninformationpro- was vitrified in borosilicate glass using a slurry-fed Joule melter of a videdby DOE design similar to that employed at the SRS. The LLW was further evapo- Washed sludge ~ Supernate Supernate LLW fig Program ~: Zeolite HLW Glass HLW Vitrification Interim Storage Possible On-Site I Relocation Off-Site Transport H ~ G H - L E V E E W A S T E 26
rated and blended with cement for interim storage on site. To date, the WVDP has almost completed6 retrieval and vitrification of its HEW. The 255 glass canisters produced are stored temporarily on site, awaiting eventual disposal in a geological repository. The WVDP is now perform- ing final flushing of the tanks to prepare for shutdown of the waste melter. 6More than 99 percent of the alpha emitters and more th beta and gamma emitters have been retrieved to date. a in 97 percent of the O v e r v i e w o f H L W C h a I I e n 9 e s a t D O E S i t e s 27
