3 Alternative EBR-II Spent Fuel Treatment Technologies

To address the task of possible alternative technologies to EMT for the treatment of EBR-II spent fuel, the committee was briefed by a number of experts in the field of nuclear waste treatment at the March 16-17, 1998, committee meeting (Appendix D). These experts were from a broad spectrum of facilities that deal with nuclear waste issues, including Idaho National Engineering and Environmental Laboratory (INEEL), Oak Ridge National Laboratory (ORNL), Savannah River Site (SRS), and the Pacific Northwest National Laboratory (PNL). The presenters were requested to give an overview of the alternative treatment technology in which they had expertise, and then to discuss the technology as a possible treatment option for EBR-II spent fuel. The committee then was able to question the speakers about details of their processes.

DIRECT DISPOSAL1

In the direct disposal, or the high-integrity canister (HIC) option, disposal of spent nuclear fuel (SNF) in a geological repository takes place with little or no treatment or conditioning. This design uses a thick outer steel barrier, or overpack, over a thinner container constructed from a highly corrosion resistant alloy. The outer layer provides not only initial corrosion protection but also cathodic protection of the inner container should the outer layer be breached after prolonged exposure. The inner corrosion-resistant container may be affected by localized corrosion such as pitting or crevice corrosion. Such containment-based strategies for geological disposal of radioactive material are predicated on the concept that radioactive decay of radionuclides during the containment period will lower their inventories, hence lower the release of these radionuclides when containment eventually fails. For example, if a container lasts for a period that is ten times longer than the half-life of a given radionuclide, then the inventory of that radionuclide will decrease by a factor of 1024 during that containment period.

The acceptability of an extended-containment strategy, such as the HIC, depends on a number of factors, such as the time scale adopted for safety assessment of a geological repository and the chemical controls on radionuclide release.

With respect to the time scale appropriate for safety analyses, the National Research Council's Committee on Technical Bases for Yucca Mountain Standards states in its report:2 “We recommend calculation of the maximum risks of radiation releases whenever they occur as long as the geologic characteristics of the repository environment do not change significantly. The time scale for long-term geologic processes at Yucca Mountain is on the order of approximately one million years.” It is also noted by the presenter, however, that recent federal legislation has considered reducing the time period of safety assessment to 10,000 years or less.

DOE's own safety analyses of a potential repository, Yucca Mountain (TSPA-95, TRW, 1995), show that the important dose-contributing nuclides are 36Cl, 99Tc, 129I, 135Cs, and 237Np. All of these radionuclides have half-lives greater than 200,000 years. To significantly mitigate releases of these radionuclides, containment times for a HIC package would need to be on the order of one million years, or greater. The committee is not aware of any credible evidence for asserting the longevity of any containment material for such a period of time in a deep geological repository.

In addition, for solubility-limited radionuclides, such as Np-237 (half-life of 2.14 million years), it was stated by the presenter that reductions in inventory by extended containment may not impact peak dose-release rates. This is because such peak rates are controlled by an intensive variable (solubility)

1  

Based in part on information presented to the committee by Howard Eckert (DOE) and Eric Shaber (INEEL), March 16, 1998, Washington, D.C.

2  

Technical Bases for Yucca Mountain Standards, National Research Council, National Academy Press, Washington, D.C., 1995, pp. 71-72.



The National Academies | 500 Fifth St. N.W. | Washington, D.C. 20001
Copyright © National Academy of Sciences. All rights reserved.
Terms of Use and Privacy Statement



Below are the first 10 and last 10 pages of uncorrected machine-read text (when available) of this chapter, followed by the top 30 algorithmically extracted key phrases from the chapter as a whole.
Intended to provide our own search engines and external engines with highly rich, chapter-representative searchable text on the opening pages of each chapter. Because it is UNCORRECTED material, please consider the following text as a useful but insufficient proxy for the authoritative book pages.

Do not use for reproduction, copying, pasting, or reading; exclusively for search engines.

OCR for page 13
ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY 3 Alternative EBR-II Spent Fuel Treatment Technologies To address the task of possible alternative technologies to EMT for the treatment of EBR-II spent fuel, the committee was briefed by a number of experts in the field of nuclear waste treatment at the March 16-17, 1998, committee meeting (Appendix D). These experts were from a broad spectrum of facilities that deal with nuclear waste issues, including Idaho National Engineering and Environmental Laboratory (INEEL), Oak Ridge National Laboratory (ORNL), Savannah River Site (SRS), and the Pacific Northwest National Laboratory (PNL). The presenters were requested to give an overview of the alternative treatment technology in which they had expertise, and then to discuss the technology as a possible treatment option for EBR-II spent fuel. The committee then was able to question the speakers about details of their processes. DIRECT DISPOSAL1 In the direct disposal, or the high-integrity canister (HIC) option, disposal of spent nuclear fuel (SNF) in a geological repository takes place with little or no treatment or conditioning. This design uses a thick outer steel barrier, or overpack, over a thinner container constructed from a highly corrosion resistant alloy. The outer layer provides not only initial corrosion protection but also cathodic protection of the inner container should the outer layer be breached after prolonged exposure. The inner corrosion-resistant container may be affected by localized corrosion such as pitting or crevice corrosion. Such containment-based strategies for geological disposal of radioactive material are predicated on the concept that radioactive decay of radionuclides during the containment period will lower their inventories, hence lower the release of these radionuclides when containment eventually fails. For example, if a container lasts for a period that is ten times longer than the half-life of a given radionuclide, then the inventory of that radionuclide will decrease by a factor of 1024 during that containment period. The acceptability of an extended-containment strategy, such as the HIC, depends on a number of factors, such as the time scale adopted for safety assessment of a geological repository and the chemical controls on radionuclide release. With respect to the time scale appropriate for safety analyses, the National Research Council's Committee on Technical Bases for Yucca Mountain Standards states in its report:2 “We recommend calculation of the maximum risks of radiation releases whenever they occur as long as the geologic characteristics of the repository environment do not change significantly. The time scale for long-term geologic processes at Yucca Mountain is on the order of approximately one million years.” It is also noted by the presenter, however, that recent federal legislation has considered reducing the time period of safety assessment to 10,000 years or less. DOE's own safety analyses of a potential repository, Yucca Mountain (TSPA-95, TRW, 1995), show that the important dose-contributing nuclides are 36Cl, 99Tc, 129I, 135Cs, and 237Np. All of these radionuclides have half-lives greater than 200,000 years. To significantly mitigate releases of these radionuclides, containment times for a HIC package would need to be on the order of one million years, or greater. The committee is not aware of any credible evidence for asserting the longevity of any containment material for such a period of time in a deep geological repository. In addition, for solubility-limited radionuclides, such as Np-237 (half-life of 2.14 million years), it was stated by the presenter that reductions in inventory by extended containment may not impact peak dose-release rates. This is because such peak rates are controlled by an intensive variable (solubility) 1   Based in part on information presented to the committee by Howard Eckert (DOE) and Eric Shaber (INEEL), March 16, 1998, Washington, D.C. 2   Technical Bases for Yucca Mountain Standards, National Research Council, National Academy Press, Washington, D.C., 1995, pp. 71-72.

OCR for page 13
ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY rather than an extensive variable (inventory) of the system. A similar insensitivity on repository performance has been identified in analyses of partitioning and transmutation of SNF; reduction in Np-237 inventory of >99.9 percent is needed to cause a reduction in the peak release rate of this solubility-limited nuclide.3 Furthermore, extended containment would not alleviate concerns regarding the chemical reactivity of metallic-uranium SNF in an underground repository. If a HIC option were pursued, the first preparation step would be to vacuum-dry the SNF to remove free water. Next, the SNF would be packaged in a standard-size canister fabricated from C-22 or possibly some other highly corrosion-resistant material such as titanium (Ti) Grade-12. C-22 (UNSN06022) is a Ni-Cr-Mo alloy, which is highly corrosion resistant due to its high (22%) Cr and high (13%) Mo content. Recent laboratory studies in solutions simulating environments expected in potential underground, high-level nuclear waste repositories have suggested that C-22 and Ti Grade-12 are immune to localized corrosion in these environments.4 However, it is extremely difficult to make reliable long-term predictions concerning the susceptibility to localized corrosion based on the electrochemical polarization technique used in this study which evaluates the probability that pitting will occur. Nevertheless, the design of this system should provide exceptional corrosion resistance since it combines the corrosion allowance of the outer container and its eventual role as sacrificial anode with the exceptional corrosion resistance of the inner container. As an option, neutron poison inserts could be added within the canister to help mitigate potential nuclear criticality concerns presented by untreated, highly enriched SNF, including EBR-II. The high-integrity canister would then be placed in dry storage at an appropriate location. Prior to emplacement in a geological repository, the SNF-containing HIC could be placed into an overpack designed to promote containment under repository conditions. Variations of this direct disposal concept include co-disposal of SNF canisters with canisters of reprocessed high-level waste (HLW) borosilicate glass within a single overpack. After waste emplacement, the encapsulated SNF (e.g., EBR-II) would be isolated from direct contact by water or steam for the lifetime of the overpack and the HIC. This containment period will depend on a variety of environmental factors, such as water infiltration rate, temperature, and future climatological conditions. Certainly a minimum containment time of 1,000 years might be envisioned. Containment times several times longer than this 1,000-year period might be postulated, if substantiated by future “service condition” tests on overpack and HIC materials. Reliable predictions of HIC lifetimes to tens of thousands of years or greater are, however, unsupported by any data known to the committee. GLASS MATERIAL OXIDATION AND DISSOLUTION SYSTEM5 The glass material oxidation and dissolution system (GMODS) was described by the presenter as a general-purpose process for treating miscellaneous fissile materials including EBR-II SNF.6 The basic concept of GMODS is to add unprocessed SNF and a sacrificial oxide to a glass melter at a temperature of 800-1,000 ºC. The solvent is a lead-borate glass (2PbO•B2O3). The oxides in the SNF dissolve in the molten glass. The PbO serves as the oxidant for the SNF metal components, e.g., uranium, plutonium, zirconium, etc., converting them in situ to metal oxides, which dissolve in the glass, and metallic lead. Metals more noble than Cu are not oxidized and remain dissolved in the molten lead. The lead metal 3   Nuclear Wastes: Technologies for Separation and Transmutation, National Research Council, National Academy Press, Washington, D.C., 1996. 4   A.K. Roy, D.L. Fleming, and B.Y. Lum, “Localized Corrosion Behavior of Candidate Nuclear Waste Package Container Materials,” p. 54 in Materials Performance, National Association of Corrosion Engineers, Houston, TX, March 1998. 5   Based in part on information presented to the committee by Charles W. Forsberg (ORNL), March 16, 1998, Washington, D.C. 6   For a representative reference to the GMODS process, see C. W. Forsberg, E.C. Beahm, G.W. Parker, and K.R. Elam, “Conversion of Radioactive and Hazardous Chemical Wastes into Borosilicate Glass Using the Glass Material Oxidation and Dissolution System,” Waste Management, Vol. 16, 1996, p. 615.

OCR for page 13
ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY (with dissolved metals) separates from the molten glass and sinks to the bottom of the melter. The molten metallic lead is then separated from the molten glass, reoxidized to PbO, and recycled to the melter. Optionally, the dissolved noble metals can be separated and recovered from the Pb metal. Carbon is added to reduce the excess PbO to lead metal. Lead-borate melt is converted to borate fusion melt (B2O3 with dissolved oxides). Due to the powerful dissolution and oxidation properties of the 2PbO•B2O3 melt, containment is a concern, and a water-cooled, cold-wall, induction-heated (or skull) melter must be used. Three options were proposed depending on how the GMODS process would operate. The first option would lead directly to the production of a borosilicate waste form into which most of the components of the SNF would be incorporated (i.e., most fission products, uranium, plutonium, and minor actinides, and cladding materials as metal oxides distributed in the glass matrix). The presenter reported his expectation that this waste form would be suitable for disposal in a repository with little or no further development or qualification as a waste form. A second option would be to use the borate fusion melt product from GMODS as feed to the Savannah River PUREX process so that uranium (primarily highly enriched uranium) and plutonium could be recovered. The third option would be to couple GMODS directly to the Defense Waste Processing Facility at Savannah River where the borate fusion melt would be fed, without further processing, directly into the existing glass melter. The GMODS process is at a very early stage of development. Studies to date include: Thermodynamic analysis showing chemical feasibility (chemical reaction studies of simulated components [depleted uranium and nonradioactive fission products] of SNF in the lead borate melt in a sacrificial high-density Al2O3 crucible [not cold-wall melter] establishing proof of principle for the melting/oxidation step), Analysis of proposed flow sheets, Identification of equipment options, and Identification of uncertainties. The GMODS process was claimed to be similar to processes practiced industrially: smelting of lead ore, lead recycle operations, and various cold-wall melting processes, including those under development for waste management purposes in Russia and France (hot prototype tests for production of HLW). At this stage of development, the overall process is conceptual, with the basic chemistry partially demonstrated and understood, and analogous industrial processes suggesting process feasibility. There are unresolved engineering issues with respect to B2O3 management/recycle, treatment of off-gas containing radionuclides, chemical reaction kinetics, criticality control, containment of melt glass and molten lead, and, in general, major engineering issues. Significant development work will be required to make GMODS an industrial process, including a full cold-pilot plant, and a hot-pilot plant. GMODS was claimed by the presenter to have a number of advantages. It is proposed as a general-purpose process that could accept a wide variety of SNFs, ranging from EBR-II spent fuels to plutonium and uranium scrap and residues. Further, GMODS was claimed to require little front-end treatment or characterization and the process produces borosilicate glass, a high-level waste form generally accepted by the nuclear community. MELT AND DILUTE7 The melt and dilute process is a modification of that proposed for treatment of Al-based SNF.8,9 The fuel elements would be chopped and melted at ~650-850 ºC, and then diluted by addition of depleted 7   Based in part on information presented to the committee by Harold Peacock (SRS), March 16, 1998, Washington, D.C.

OCR for page 13
ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY uranium and iron. This process was proposed by the presenter as being applicable to metallic, non-oxide fuels. No experimental work relevant to the EBR-II SNF has been carried out; i.e., this proposed process is only at the conceptual stage. An existing phase diagram of the U-Fe system was discussed by the presenter with an indication of how the proposed process would be carried out. Off-gases, including sodium from the EBR-II drivers, Cs, Tc, I, and Kr, would require collection, treatment, and disposal. The metallic ingot resulting from treatment of the SNF would be dried and sealed into an inert gas-filled canister for interim storage. Advantages claimed by the presenter for the process are stated as being “simple and flexible, ” and the dilution with depleted uranium would reduce criticality concerns. If the metal ingots resulting from the process are to be sent ultimately to a repository, a waste form qualification would be required. No cost estimates were provided that would be directly applicable to the proposed process, although estimates for the Al-fuel melt and dilute techniques are included in a 1996 DOE report.10 THE PUREX PROCESS AND SAVANNAH RIVER SITE FACILITIES11 The PUREX process is a counter-current solvent extraction method, which has been used extensively throughout the world since 1954 to separate and purify uranium and plutonium from fission-product-containing SNF and irradiated uranium targets. The tri-n-butylphospate solvent is dissolved in an inert organic diluent such as n-paraffin oil; e.g., dodecane. As utilized at SRS, a head-end step is employed to dissolve SNF or irradiated targets. For aluminum-clad SNF or irradiated targets, the cladding is dissolved using an aqueous solution of sodium hydroxide and sodium nitrate, after which the uranium/plutonium core material is dissolved in an aqueous solution of nitric acid. Stainless steel clad SNF would require use of mechanical decladding, electrolytic dissolution, or chop/leach of the SNF after which the uranium/plutonium core would be dissolved in aqueous nitric acid. After SNF or irradiated target dissolution, the resulting aqueous nitric acid solution containing uranium, plutonium, and fission products undergoes feed clarification and pH adjustment as well as addition of gelatin for silica removal. The clarified aqueous solution is treated via the PUREX process utilizing 16 stages of centrifugal contractors and separators to produce an aqueous HLW that contains the bulk of the fission products, americium, and neptunium; a product stream that contains the recovered plutonium; and a product stream that contains the recovered uranium. The plutonium- and uranium-containing streams each undergo a second cycle of solvent washing to further separate the residual fission products and actinides from the plutonium and uranium. The presenter stated that an overall decontamination factor of one million is achieved typically for separation of plutonium and uranium from fission products, americium, and neptunium.12 The first SRS production-scale PUREX plant (221-F Canyon) went hot in November 1954 and was used for recovery of weapons-grade plutonium from irradiated uranium target elements. The system was operated and maintained remotely. Currently, the facility is in good condition and is being operated 8   Technical Strategy for the Treatment, Packaging, and Disposal of Aluminum-Base Spent Nuclear Fuel, a report of the Research Reactor Spent Nuclear Fuel Task Team, Department of Energy, Office of Spent Fuel Management, Vol. 1, June 1996, page 38. 9   Research Reactor Aluminum Spent Fuel, National Research Council, National Academy Press, Washington, D.C., 1998. In this report, the recommendation was made that, for treatment of Al-based spent nuclear fuel, “melt and dilute treatment is worth pursuing despite the additional development and infrastructure requirements because it allows more control over waste form composition and performance characteristics.” 10   Technical Strategy for Treatment, Packaging, and Disposal of Aluminum-Base Spent Nuclear Fuel, a report of the Research Reactor Spent Nuclear Fuel Task Team, Department of Energy, Office of Spent Fuel Management, Vol. 1, June 1996, p. 59. 11   Based in part on information presented to the committee by Malcom McKibben (SRS), March 16, 1998, Washington, D.C. 12   Based in part on information presented to the committee by Malcom McKibben (SRS), March 16, 1998, Washington, D.C.

OCR for page 13
ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY to dispose of legacy DOE aluminum-clad SNF from throughout the United States. Present system throughput is 40 metric tons of uranium per month, about 22 percent of the historic throughput. Use of the SRS PUREX facility for treatment of EBR-II driver and blanket elements remaining after the ongoing EMT demonstration (about 43 metric tons of heavy metal) would require addressing successfully three significant considerations: The sodium-bonded EBR-II SNF is generally incompatible with the aqueous nitric acid solutions employed in the head end of the SRS facility. Mechanical decladding and sodium removal was done by Atomics International several years ago for a small number of EBR-II blanket elements; the resulting declad blanket elements are in storage at SRS. Treatment of EBR-II fuel in the SRS facility would require either hiring an outside organization to remove the sodium or developing a sodium removal approach to use at the SRS facility. The presence of stainless-steel cladding on the EBR-II driver and blanket elements would require significant modifications or additions to the existing head end of the SRS PUREX facility. This would require installation of equipment necessary for employment of one of three possible approaches (mechanical decladding, electrolytic dissolution, or chop/leach). If electrolytic dissolution of the EBR-II elements were employed, the presence of zirconium (5 weight percent) in the driver elements would result in forming unacceptably large quantities of insoluble sludge during electrolytic dissolution of the driver elements. Electrolytic dissolution has been employed successfully in the SRS facility for feed materials that did not contain appreciable amounts of zirconium. The presenter stated that if one assumes that the above three considerations could be addressed successfully in order to treat EBR-II SNF, it was estimated by the presenter that a period of about three months would be required (excluding time required for SNF pretreatment) for processing of EBR-II driver and blanket elements remaining after the EMT demonstration. This time period results largely from the requirement for successive shutdowns for fissile material accountability control. In addition, a period of about 1 year would be required for final purification (B-Line finishing) of plutonium and uranium products. The resulting product streams would be high-purity uranium trioxide and plutonium metal, which would be sent to the surplus plutonium program. In addition, HLW would be produced in the form of about 40 borosilicate glass canisters as well as a large volume of low-level waste (LLW) in concrete. These product and waste forms are typical of those currently being addressed at the SRS. SRS staff estimate that operation with EBR-II fuel could begin approximately in January 2001. Estimated costs prepared by the SRS staff include about $160 million for pretreatment head-end equipment (does not include costs for EBR-II element decladding and sodium removal), $25 million for operation of 221-F Canyon for 3 months, $60 million for operation of B-Line for 6 months for product finishing, and $4 million for converting resulting HLW and LLW to final waste forms. The cost for plutonium disposal was not estimated. CHLORIDE VOLATILITY13 The proposed process uses the differences in volatilities of chloride compounds to separate the constituents of spent nuclear fuels.14 The unit operations consist of: (1) a high-temperature chlorination step that operates at approximately 1500 ºC and converts metallic fuel and cladding materials to gaseous chloride compounds, (2) a molten zinc chloride bed that removes the TRU chlorides and most of the fission products (FP) and operates at approximately 400 º C, (3) a series of fluidized beds and condensers operating at successively lower temperatures to condense zirconium tetrachloride, uranium hexachloride, 13   Based in part on information presented to the committee by Jerry Christian (INEEL), March 17, 1998, Washington, D.C. 14   For a representative reference describing the chloride volatility process, see Olson A.L. (ed.), “Potential Dispositioning Flowsheets for ICPP SNF and Wastes,” INEL-95/0534, November 1995.

OCR for page 13
ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY and stannous tetrachloride, and (4) a zinc chloride regeneration/recycle process. The TRU and FP chloride are then converted to either fluorides or oxides for final disposal. A proposed flowsheet was developed at INEEL for processing naval fuel. The proposed application to EBR II fuel is an extrapolation of the naval fuel process scheme. Theoretical chloride volatilities have been used to postulate the equipment sizing and operating parameters. Because of the lack of any experimental basis, significant concerns exist about the distribution of chloride compounds for multivalent elements such as uranium and plutonium. These concerns, in turn, lead to potential uncertainties in separation capabilities and overall flowsheet performance. The final waste form is also questionable. The use of halides, either fluorides or chlorides, for the TRU and FP elements raises questions about the use of a glass or vitrified waste form. The proposal to use boric acid at 1000 ºC to convert the chlorides to oxides allays some of these concerns. DIRECT PLASMA ARC-VITREOUS CERAMIC PROCESS15 In this process, components of SNF's are melted and oxidized, with the help of an oxygen lance, in a rotating furnace containing molten ceramic materials at a temperature of 1600 ºC or higher.16 A direct current plasma torch that creates a high-temperature plasma between the hollow torch electrode and the rotating furnace (i.e., the external electrode) supplies the energy required in the process. This process begins by placing the starting ceramic material in the furnace and striking the arc. As the temperature rises, the starting material melts to form a molten pool that is kept in the furnace by the centrifugal force of rotation. Next, the SNF, which has been cut into small pieces (less than 6 inches), is added, melted, oxidized, and homogenized in the melt. When this step is complete, the speed of rotation is reduced to allow the molten vitreous ceramic to pour by gravity flow through an axial hole at the bottom of the furnace directly into a canister. Plasma systems have been used for melting metals since 1980, and about 25 units are in industrial operation today. Plasma systems have been used by the Swiss for treating medical and industrial wastes and the plasma arc centrifugal (PAC) system has been demonstrated for treating low-level radioactive wastes. PAC technology has also been demonstrated successfully for processing small-caliber and hand-held pyrotechnic, smoke, and dye ordnance. The scale of most plasma arc systems operated to date is probably larger than could be used with DOE SNF and other DOE fissile-containing material. Since the plasma arc process is a high-temperature process, care must be taken to provide treatment of off-gases produced in the process. In particular, particulates and volatile and/or semi-volatile elements (and oxides) such as Cs, Tc, Se, and Ru would escape from the melts and would have to be trapped. The presenter claims that the off-gas system developed for plasma arc processing is in full compliance with the Environmental Protection Agency's (EPA) emission requirements. However, the plasma arc process would have to be modified (i.e., a dry off-gas system) for the treatment of SNF so that all materials and filters used in the operation could be recycled to the melter to minimize secondary waste. The advantages claimed by the presenter of plasma processing are as follows: Its simplicity is that it is a single-step process that converts SNF and associated wastes into a vitreous ceramic waste form. The waste form is capable of high loading (up to 90 percent), and thus, the total volume of waste would be reduced. 15   Based in part on information presented to the committee by Xiangdong Feng (PNL), March 17, 1998, Washington, D.C. 16   For a representative reference describing the plasma arc process, see Gray, W.J. and R.T. Ines (eds.), Scientific Basis for Nuclear Waste Management XX, Materials Research Society, Pittsburgh, Pa., 1996, pp. 25-32.

OCR for page 13
ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY The necessity to perform costly characterization would be greatly reduced compared to alternate technologies. Minimal handling and pretreatment would be needed with the exception of reducing the size of materials to be compatible with the feed system. A durable and safe, monolithic waste form suitable for long-term storage would be produced. The presenter also claimed that this process would be compatible with processing a wide range of SNF types of varying fuel compositions in various degrees of degradation (metallic, oxide, carbide, hydride, sodium-bearing, aluminum, zirconium). This process was also claimed by the presenter to be suitable for processing a range of different cladding materials (stainless steel, zircaloy, aluminum), miscellaneous Pu residues, and other contaminated wastes streams such as sludge, soil, metals fragments, metal containers, and concrete grits. The vitreous waste form resulting from this process would have to undergo extensive qualification testing. A test plan appropriate for this type of material would have to be first developed and then shown to be at least equivalent in performance to the acceptance tests developed for borosilicate glass. No work on qualifying this waste form has been performed to date. Other questions that would have to be addressed to treat SNF, including the EBR-II fuel remaining after the demonstration at ANL-W, would include feasibility, safety, and reliability of remote hot-cell operation; criticality control including the potential release of cooling water from the torch; process throughput and batch limits; off-gas treatment for high-temperature plasma arc process relative to containment of high levels of fission products such as 137Cs and 99Tc; and qualification of the waste form. With regard specifically to EBR-II fuels, extensive work would have to be performed to ensure that the plasma arc process would be compatible with safely processing potentially pyrophoric uranium and sodium metal. Also, because EBR II fuel is sodium-bonded to the stainless-steel cladding, studies would need to be conducted under the required operating conditions to determine the loss of sodium by volatilization. COMMITTEE ANALYSIS Direct Disposal From a materials and design standpoint, overpack technologies appear to be technically sound. However, the long-term durability of the proposed overpack container has not been demonstrated or documented. Without such a demonstration of extended containment, the ability of the HIC disposal concept to meet the stated safety standards proposed by the National Research Council is unknown.17 At the present time, direct emplacement of EBR-II SNF is precluded by DOE policy concerning acceptance of RCRA-designated mixed waste (which contains both hazardous and radioactive waste). Because of the presence of both metallic uranium and sodium, EBR-II SNF is categorized as an RCRA hazardous waste that is potentially both pyrophoric and reactive. Glass Material Oxidation and Dissolution System The GMODS process would be an attractive general approach that could be employed if it were successfully developed. However, considering the time period and cost necessary for development of the GMODS relative to the level of maturity of the EMT process, GMODS does not appear to be a viable alternative for processing only EBR-II SNF unless the process would also be applied to other DOE 17   Technical Bases for Yucca Mountain Standards, National Research Council, National Academy Press, Washington, D.C., 1995, pp. 71-72.

OCR for page 13
ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY SNFs and miscellaneous fissile materials. Melt and Dilute No experimental work relevant to the processing of EBR-II SNF has been carried out. No development schedule or cost estimates were presented; therefore the committee does not have sufficient information to evaluate melt and dilute as a viable alternative to EMT. PUREX The PUREX process is well developed and has been used to treat SNF and irradiated uranium for over 40 years. The development of a versatile head-end process to handle mechanical decladding, sodium removal, and zirconium sludge formation for EBR-II SNF for the SRS PUREX facility does not seem justified solely for the purpose of treating the relatively small quantity of EBR-II fuel that will remain after completion of the EMT demonstration. However, DOE would need to consider in the broader context whether developing a versatile head-end treatment step for the SRS PUREX facility is an attractive means for treating large quantities of DOE SNF (e.g., Zr or zircaloy-clad fuels such as N-reactor fuel) and other fissile-containing materials. A significant issue for treating EBR-II fuel at SRS by PUREX relates to public concerns about transportation of the fuel from the current storage site at ANL-W to SRS.18 A final consideration is that at the present time the PUREX canyons at SRS are due to be shut down and would not be available for this work according to Administration Policy. Chloride Volatility As proposed, this technology is applicable to a very limited range of fuel types. A reducing agent, such as carbon monoxide, would be needed in the chlorination step to prevent the formation of oxychloride compounds even for metallic fuel forms. The behavior of some of the constituents of EBR-II fuel, such as metallic sodium, may also limit the suitability of this process due to the accumulation of considerable amounts of sodium chloride during the chlorination step. Other unknown quantities, including the behavior of stainless steel, in the chlorination step further decrease the potential applicability of this process to stainless steel-clad SNF (e.g., EBR-II fuel). At present, considering also cost and schedule, the chloride volatility process is not competitive with the current EMT process. In addition, the committee is aware that a significant amount of work on chloride volatility processes was conducted during the 1960s at ANL and ORNL.19 However, this earlier work does not suggest to the committee that this approach is an attractive alternative to EMT. Plasma Arc Although plasma arc processing has been used successfully to treat nonradioactive and low-level radioactive wastes, significant research, development, and demonstration would be needed to process SNF because of the much higher fission product and fissile material content. With regard specifically to EBR-II fuels, extensive work would have to be performed to ensure that the plasma arc process would be compatible with safely processing potentially pyrophoric uranium and volatile and reactive sodium metal. Unresolved safety issues at present preclude consideration of plasma arc processing as a viable alternative to the EMT process. 18   For a discussion of transportation issues regarding spent fuel and other nuclear wastes, see Nuclear Wastes Technologies for Separations and Transmutation, National Research Council, National Academy Press, Washington, D.C., 1996, pp. 102-103. 19   Gens, T.A. “Chloride Volatility Processing of Nuclear Fuels,” Chem. Eng. Prog. Symposium Series, Nuclear Engineering, Part X (47) 60, (1964), pp. 37-47.

OCR for page 13
ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY FINDINGS All of the processes evaluated, with the exception of the PUREX process, are at an early stage of development. Given the time and cost required to develop these processes as alternatives to electrometallurgical treatment, none of these alternative processes could be implemented and validated without significant further R &D and “hot” demonstrations. This would result in increased costs and delays in processing of the EBR-II SNF relative to ANL's proposed schedule for complete processing of EBR-II fuel. Although the PUREX process is well developed, the development of a versatile head-end process to handle mechanical decladding, sodium removal, and zirconium sludge formation for EBR-II SNF for the SRS PUREX facility would require a significant investment. This investment would have to be weighed against the cost of treating, by the EMT process, the relatively small quantity of EBR-II fuel that will remain after the completion of the EMT demonstration. Although overpack (direct disposal) does not require processing, direct emplacement of EBR-II SNF is presently precluded by DOE policy concerning acceptance of RCRA-designated mixed waste.

OCR for page 13
ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY This page in the original is blank.