A backfill placed between a nuclear waste canister and the host geology of a nuclear waste repository can impede the migration of water through the waste package and retard the movement of radionuclides into the geologic formation. Hydraulic conductivities and swelling pressures are being determined as functions of the density of the compacted backfill, temperature, radiation dose, hydraulic head and the chemical composition of the permeating fluid. Bentonite clays and bentonite/sand mixtures have received initial emphasis. Sodium bentonite and calcium bentonite samples compacted to a dry density of 2.1 g/cm3 had hydraulic conductivities in the range of 10−12 to 10−13 cm/s. In addition, batch distribution ratios (Rd) for Sr, Cs, Am, Np, I, U and Tc have been measured for a number of candidate backfill materials. Both initial permeability and sorption studies have used a synthetic basaltic ground water.

The process known as Catalyzed Electrochemical Plutonium Oxide Dissolution (CEPOD) has been shown effective for removing plutonium from a variety of residues and solids. This process involves the electrochemical oxidation of PuO{sub 2} (and other Pu species) to (PuO{sub 2}){sup 2+}, and dissolution of the latter species in the anode solution (anolyte). Silver is used to transfer charge from the electrodes to the solid Pu oxide. Ag (1) is oxidized at the anode to Ag(II) and carried by the solution to the plutonium oxide solids, where the silver and oxide undergo a redox reaction that converts Pu(IV) to Pu(VI), and Ag(II) to Ag(I). Other metal ions [such as Ce(IV) and Co(III)] may also be used for this charge transfer, but have been found to be less effective than silver. The same process may be used to destroy various organic materials (such as paper and wood, oil and fuels, and synthetic polymer materials) by complete oxidation to CO{sub 2} or H{sub 2}O, for example. Upon completion of a CEPOD dissolver run, the anolyte may be processed to remove solution species of interest (i.e., Pu), or the anolyte may be recycled, or disposed. Because silver is a Resource Conservation and Recovery Act (RCRA) land ban material, it must be removed from waste streams. Preliminary experiments, completed in FY 1991, demonstrated a simple, effective technique for silver removal from solutions. Ascorbic acid (C{sub 6}H{sub 8}O{sub 6}) Was Used to reduce silver ion to metallic silver, which precipitates from solution. The process was demonstrated effective on a bench scale using samples of actual CEPOD anolyte. Further experiments, in FY 1993, optimized these parameters and demonstrated the effectiveness of the technique on CEPOD anolyte on a larger, process scale (liters of solution). This report describes both the preliminary bench-scale experiments and the more recent process-scale experiments. The results are also compared to electro-deposition, another method of silver ion removal.

Catalyzed Electrolytic Plutonium Oxide Dissolution (CEPOD) was first demonstrated at PNL in early 1974 in work funded by EXXON Corporation. That work was aimed at dissolution of Pu-containing residues remaining in mixed-oxide reactor fuels dissolution and was first publicly disclosed in 1981. The process dissolves PuO{sub 2} in an anolyte containing small (catalytic) amounts of elements that form kinetically fast, strongly oxidizing ions. These are continuously regenerated at the anode. Catalysts used, in their oxidized form, include Ag{sup 2+}, Ce{sup 4+}, Co{sup 3+}, and AmO{sub 2}{sup 2+}. This paper reviews the chemistry involved in CEPOD and the results of its application to the dissolution of the Pu content of a variety of PuO{sub 2}-containing materials such as off-standard oxide, fuels dissolution residues, incinerator ash, contaminated soils, and other scrapes or wastes. Results are presented for both laboratory-scale and plant-scale dissolvers. Spin-off applications such as decontamination of metallic surfaces and destruction of organics are discussed. 27 refs., 14 figs.

Yttrium-90 is a daughter of {sup 90}Sr, a fission product with about a 30-year half-life, that decays through emission of a {beta} particle to give stable {sup 90}Zr. Among radionuclides that have been used to label antibodies for therapeutic purposes, {sup 90}Y has received much attention owing to its favorable half-life of 64.1 hours and its predictable chemical properties. The procedure that we have used to produce multicurie quantities of {sup 90}Y is similar to that developed at Oak Ridge National Lab.

A study focusing on identification and recommendation of the most applicable existing processes for (aqueous) head-end treatment of plutonium residues has recently been concluded. Four U.S. Department of Energy Defense Program sites were represented in the working group: Los Alamos National Laboratory (Los Alamos), the Hanford Site, Rocky Flats Plant (Rocky Flats), and Savannah River Plant (Savannah River), with further participation by Pacific Northwest Laboratory and Savannah River Laboratory. This report describes the results of the study and is intended to document existing processes, categorize these processes by applicability, and recommend processes for specific types of residue feeds. Three principle areas of plutonium residue (aqueous) processing were considered in the study: Feed pretreatment, including comminution/size reduction, screening, sorting, and oxidation; dissolution in aqueous solution using chemical/mechanical and electrochemical systems; claification via gravity settling/decantation, centrifugation, and filteration. Results are discussed.

Dissolution of plutonium (PU) from high-fired plutonium dioxide (PuO/sub 2/) and the leaching of Pu from scrap or wastes have often posed difficult process problems. Traditionally, dissolution has been accomplished using 12 M HNO/sub 3/-0.18 M HF at boiling temperatures. This work discusses the development of electrochemical dissolution of PuO/sub 2/ in compartmented cells. Small amounts of oxidation catalysts are used to carry electrons from the solid PuO/sub 2/ to the anode surface, producing PuO/sub 2//sup 2 +/ solution. PuO/sub 2/ that has been fired at temperatures to 1700/sup 0/C can be dissolved at 25/sup 0/C to concentrations exceeding 400 g Pu/L using the catalyzed electrochemical plutonium oxide dissolution (CEPOD) process. The silver ion is used as the redox catalyst. The dissolution rates are 5 to 15 times faster than the rates obtained using the current process (12 M HNO/sub 3/-0.18 M HF), and the corrosive fluoride ion is avoided.

The nuclear industry involves a number of operations. Uranium ore must first be mined and the uranium recovered from the ore, purified, and concentrated. After the uranium has been enriched and fabricated into fuel elements, it is placed in nuclear reactors where it produces energy, fission products, and transmutation products. Finally, if the fuel cycle is completed, the uranium and useful transmutation products are recovered and separated from each other as well as from the fission products. The uranium may be recycled or used elsewhere, while most of the fission products become waste. Ion exchange finds use in nearly every part of the nuclear fuel cycle; these uses are the subject of this paper.

The Nuclear Waste Vitrification Project was conducted to demonstrate the vitrification of high-level liquid waste (HLLW) generated during the reprocessing of spent fuel discharged from an operating light water reactor.Six pressurized water reactor fuel assemblies, containing 2.3 tU, were processed for the generation of HLLW. A conventional Purex-type process was used for the first cycle so that the HLLW generated would be typical of the nitric acid, fission product waste stream from the first extraction cycle of a commercial plant. Uranium and nonradioactive chemicals, normally added to the HLLW by back-cycling of waste from second and third solvent-extraction cycles, were added to the dilute HLLW to produce a waste composition typical of the HLLW from a commercial plant.Uranium and nonradioactive chemicals, normally added to the HLLW by back-cycling of waste from second and third solvent-extraction cycles, were added to the dilute HLLW to produce a waste composition typical of the HLLW from a commercial plant. The waste was then concentrated tenfold to provide feed for solidification by the spray calciner/in-can melting process. During calcination, the liquid waste was pumped at a rate of 10 to 15 ℓ/h to the calciner vessel, which was heated to 750°C. The powdered calcine fell into a stainless steel canister, which was maintained at 1050°C; this canister was attached directly to the bottom of the calciner. Glass-forming chemicals were metered into the canister simultaneously with the calcine. After the materials melted, the canister was cooled to produce vitreous glass. Two 20.3-cm-diam × 244-cm-high canisters containing glass were produced.

A backfill barrier, emplaced between the containerized waste and the host rock, can both protect the other engineered barriers and act as a primary barrier to the release of radionuclides from the waste package. Attributes that a backfill should provide in order to carry out its required function have been identified. Primary attributes are those that have a direct effect upon the release and transport of radionuclides from the waste package. Supportive attributes do not directly affect radionuclide release but are necessary to support the primary attributes. The primary attributes, in order of importance, are: minimize (retard or exclude) the migration of ground water between the host rock and the waste canister system; retard the migration of selected chemical species (corrosive species and radionuclides) in the ground water; control the Eh and pH of the ground water within the waste-package environment. The supportive attributes are: self-seal any cracks or discontinuities in the backfill or interfacing host geology; retain performance properties at all repository temperatures; retain peformance properties during and after receiving repository levels of gamma radiation; conduct heat from the canister system to the host geology; retain mechanical properties and provide resistance to applied mechanical forces; retain morphological stability and compatibility with structural barriers and with the host geology for required period of time. Screening and selection of candidate backfill materials has resulted in a preliminary list of materials for testing. Primary emphasis has been placed on sodium and calcium bentonites and zeolites used in conjunction with quartz sand or crushed host rock. Preliminary laboratory studies have concentrated on permeability, sorption, swelling pressure, and compaction properties of candidate backfill materials.

Sequential anion and cation exchange processes have been used for the final purification of 241Am recovered during the reprocessing of aged plutonium metallurgical scrap. Plutonium was removed by absorption on Dowex 1, X-3.5 (30–50 mesh) anion exchange resin from 6.5–7.5 M HNO3 feed solution. Following a water dilution to 0.75–1.0 M HNO3, americium was absorbed on Dowex 50W, X-8 (50–100 mesh) cation exchange resin. Final purification was accomplished by elution of the absorbed band down 3 to 4 successive beds of the same resin, preloaded with Zn2+, with an NH4OH buffered chelating agent. The recovery of mixed 241Am−243Am from power reactor reprocessing waste has been demonstrated. Solvent extraction was used to recover a HNO3 solution of mixed lanthanides and actinides from waste generated by the reprocessing of 13.5 tons of Shippingport Power Reactor blanket fuel. Sequential cation exchange band-displacement processes were then used to separate americium and curium from the lanthanides and then to separate ∼60 g of 244Cm from 1000 g of mixed 241Am-243Am.

Abstract An efficient extraction process that does not utilize halides or organic solvents has been developed for the recovery and purification of 210po. Polonium-210, produced in bismuth metal by neutron irradiation, is extracted from molten bismuth metal into molten NaOH in an unique 3-compartment contactor under an inert atmosphere. At a temperature of 450±25°C, and at a NaOH/Bi weight ratio of 0.044, five successive 60-minute extractions remove >96% of the 210po. Following phase separation and freezing, additional purification steps include dissolution of the solidified NaOH in HN03, recovery of 210po from this solution by MnO2 carrier precipitation, dissolution of the precipitate by H2O2 in HNO3, and, finally, electrodeposition of 210po onto platinum gauze.