Executive Summary

INTRODUCTION

The Committee on Electrometallurgical Techniques for DOE Spent Fuel Treatment was formed to evaluate the technical viability of electrometallurgical technology as a method for treating U.S. Department of Energy (DOE)1 spent nuclear fuel (SNF).2 Over the course of the committee’s operating life, this charge has remained constant. Within the framework of this overall charge, the scope of the committee’s work—as defined by its statements of task—has evolved in response to further requests from DOE, as well as technical accomplishments and regulatory and legal considerations. As part of its task, the committee has provided periodic assessments of Argonne National Laboratory’s (ANL’s) R&D program on the electrometallurgical technology.

In 1995, ANL proposed the use of electrometallurgical technology for treatment of all spent nuclear fuel in the DOE inventory.3 Treatment would convert the fuel to components suitable for waste disposal as well as separate out any material that might be of use in future DOE operations. Electrometallurgical technology was suggested as a means to produce the same waste forms for all of the spent fuels in the DOE inventory, thus providing substantial cost savings for qualification of these waste materials for disposal in a geologic repository.

Electrometallurgical technology (EMT) consists of electrorefining the reactor fuel in an electrochemical cell. The fuel, in metallic form, is selectively dissolved at the anode while nearly pure uranium metal is deposited at the cathode, leaving fission products, fuel cladding material, plutonium, and other transuranic elements partially at the anode and partially in the molten salt electrolyte. Thus the fuel is separated into three components: metallic uranium, a metallic waste form from the anode, and a highly radioactive salt mixture that subsequently can be converted to a ceramic waste form. A key step in the ANL’s proposal was treatment of all the Experimental Breeder Reactor-II (EBR-II) spent fuel as a demonstration of the technology. As work progressed on the EBR-II spent fuel, the committee’s technical evaluation of electrometallurgical technology became increasingly focused on the demonstration project—which provided the primary source of data on which the committee could base its assessments.

1  

Acronyms and abbreviations are defined in Appendix E.

2  

DOE spent nuclear fuel refers to such fuels accumulated within the DOE complex; commercial production fuels are not included.

3  

Argonne National Laboratory, Proposal for Development of Electrometallurgical Technology for Treatment of DOE Spent Nuclear Fuel, Argonne National Laboratory, Argonne, IL, 1995.



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Electrometallurgical Techniques for DOE Spent Fuel Treatment: Final Report Executive Summary INTRODUCTION The Committee on Electrometallurgical Techniques for DOE Spent Fuel Treatment was formed to evaluate the technical viability of electrometallurgical technology as a method for treating U.S. Department of Energy (DOE)1 spent nuclear fuel (SNF).2 Over the course of the committee’s operating life, this charge has remained constant. Within the framework of this overall charge, the scope of the committee’s work—as defined by its statements of task—has evolved in response to further requests from DOE, as well as technical accomplishments and regulatory and legal considerations. As part of its task, the committee has provided periodic assessments of Argonne National Laboratory’s (ANL’s) R&D program on the electrometallurgical technology. In 1995, ANL proposed the use of electrometallurgical technology for treatment of all spent nuclear fuel in the DOE inventory.3 Treatment would convert the fuel to components suitable for waste disposal as well as separate out any material that might be of use in future DOE operations. Electrometallurgical technology was suggested as a means to produce the same waste forms for all of the spent fuels in the DOE inventory, thus providing substantial cost savings for qualification of these waste materials for disposal in a geologic repository. Electrometallurgical technology (EMT) consists of electrorefining the reactor fuel in an electrochemical cell. The fuel, in metallic form, is selectively dissolved at the anode while nearly pure uranium metal is deposited at the cathode, leaving fission products, fuel cladding material, plutonium, and other transuranic elements partially at the anode and partially in the molten salt electrolyte. Thus the fuel is separated into three components: metallic uranium, a metallic waste form from the anode, and a highly radioactive salt mixture that subsequently can be converted to a ceramic waste form. A key step in the ANL’s proposal was treatment of all the Experimental Breeder Reactor-II (EBR-II) spent fuel as a demonstration of the technology. As work progressed on the EBR-II spent fuel, the committee’s technical evaluation of electrometallurgical technology became increasingly focused on the demonstration project—which provided the primary source of data on which the committee could base its assessments. 1   Acronyms and abbreviations are defined in Appendix E. 2   DOE spent nuclear fuel refers to such fuels accumulated within the DOE complex; commercial production fuels are not included. 3   Argonne National Laboratory, Proposal for Development of Electrometallurgical Technology for Treatment of DOE Spent Nuclear Fuel, Argonne National Laboratory, Argonne, IL, 1995.

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Electrometallurgical Techniques for DOE Spent Fuel Treatment: Final Report A total of approximately 2,000 metric tons of SNF—broadly classified as production fuels, special fuels, or naval fuels—has accumulated throughout the DOE complex.4 The EMT process developed by ANL and originally proposed for the treatment of all DOE SNF is potentially applicable to a fairly wide variety of spent fuel types besides the EBR-II used by Argonne in its development and demonstration of the technology. For example, Fermi-1 blanket fuel and Fast Flux Test Facility sodium-bonded fuel are SNFs that can potentially be treated using EMT. DOE initially proposed that the EBR-II driver fuel and at least half of the blanket fuel be treated via the process.5 For most fuels, such as oxides, the fuel would first have to be converted to a suitable metallic form before the electrorefining could be applied. Electrometallurgical technology for treatment of DOE spent fuel evolved from ANL’s work on the Advanced Liquid-Metal Reactor Integral Fast Reactor (ALMR/IFR).6 With the termination of the ALMR/IFR project, this process, with some modification, served as the basis for ANL’s January 1995 proposal, which included the use of the electrometallurgical process for the treatment of EBR-II spent nuclear fuel. The proposal was accepted by DOE and was to include treatment of both reactor driver fuel and uranium blanket material. The present committee as part of its task was asked to evaluate the ongoing work on electrometallurgical technology at ANL. During its second year of operation, the committee was asked to evaluate the scientific and technological issues associated with extending ANL’s electrometallurgical program to handle plutonium, in the event that DOE might pursue an electrometallurgical treatment option for the disposition of excess weapons plutonium (WPu). The committee concluded that disposition of WPu would involve different feeds for use in SNF processing, raising several concerns with respect to electrometallurgical processing, zeolite loading, and waste form performance. THE ELECTROMETALLURGICAL PROCESS AT ANL The electrometallurgical treatment process for spent fuel at ANL consists of several distinct steps: chopping the fuel elements, electrorefining the driver and blanket fuel, removing entrained salt from uranium electrodeposits and consolidating dendritic deposits in a cathode processor, casting separately into ingots the uranium metal from the cathode and the metal residue from the anode, and, finally, mixing, heating, and pressing the salt electrolyte with zeolite to form a ceramic waste. The electrorefining step is the heart of the EMT process. The fuel element choppers are pneumatic punch presses that have been modified with blades for shearing driver and blanket fuel elements into segments for loading into the anode compartments of the Mark-IV electrorefiner (for driver fuel) and the Mark-V electrorefiner (for blanket fuel) developed at ANL. As part of ANL’s demonstration project criterion that required a blanket throughput rate of 150 kg per month sustained for 1 month,7 the blanket element chopper was used to process 3.5 blanket fuel assemblies or 66 blanket fuel elements, for a total of 164.4 kg of uranium. The Mark-IV electrorefiner has an overall anode batch size of 16 kg and was designed for processing driver fuel. The efficiency of the overall electrorefining operation is enhanced by using a second cathode inserted into melt through the fourth port. The cadmium pool in the bottom of the Mark-IV catches and dissolves any of the uranium deposits that either fall off the cathode or are scraped off the cathode by the scrapers during the electro-deposition process. During the repeatability phase of the demonstration project, the Mark-IV electrorefiner was used to treat 12 driver assemblies at an average rate of 24 kg of uranium per month over a 3-month period compared to the target of 16 kg (~4 driver assemblies) per month over a 3-month period.8 4   Argonne National Laboratory, Proposal for Development of Electrometallurgical Technology for Treatment of DOE Spent Nuclear Fuel, Argonne National Laboratory, Argonne, IL, 1995. 5   Argonne National Laboratory, Proposal for Development of Electrometallurgical Technology for Treatment of DOE Spent Nuclear Fuel, Argonne National Laboratory, Idaho Falls, ID, 1995. 6   National Research Council, Nuclear Wastes: Technologies for Separations and Transmutation, National Academy Press, Washington, D.C., 1996, pp. 27-28, 43, 155-158. 7   The full success criteria for the demonstration project, along with the goals to meet them, are included in Chapter 6. 8   R.D. Mariani, D. Vaden, B.R. Westphal, D.V. Laug, S.S. Cunningham, S.X. Li, T.A. Johnson, J.R. Krsul, and M.J. Lambregts, Process Description for Driver Fuel Treatment Operations, NT Technical Memorandum No. 111, Argonne National Laboratory, Argonne, IL, 1999.

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Electrometallurgical Techniques for DOE Spent Fuel Treatment: Final Report The second electrorefiner, the Mark-V, was designed for processing EBR-II blanket fuel, which is present in larger quantities and with lower enrichment than driver fuel. The basic difference between the Mark-IV and Mark-V is the design of the anodes and cathodes, which allowed significant increase in throughput. The material that collects in the product collector basket consists of uranium and approximately 20 wt % salt. Each anode-cathode module (ACM) in the Mark-V is capable of producing about 87 to 100 kg of uranium per month. Over the course of a 30-day process repeatability operation, the Mark-V was able to process the equivalent of 4.3 blanket assemblies (206 kg uranium (U) per month). The purpose of the cathode processor is twofold: to remove entrained salt (and any cadmium) from the uranium electrodeposits by evaporation and to consolidate dendritic deposits. The casting furnace provides a means to reduce the 235U enrichment of the driver fuel product from the cathode processor by the addition of depleted uranium and to further consolidate the uranium product. The operating parameters associated with the casting furnace include crucible coating, temperature control, and pressure control. Fabrication of Waste Forms Following the electrorefining operations the stainless-steel cladding hulls are left in the anode basket, along with the noble metal fission products, some actinides, and adhering salt electrolyte. The uranium content is about 4 wt %; zirconium (Zr) metal is added to improve performance properties and to produce a lower-melting-point alloy. The material in the anode basket is placed in the cathode processor and heated to 1100 °C to distill the salt. The charge from the cathode processor is placed in an yttrium oxide crucible, is melted at approximately 1600 °C in the casting furnace in an argon (Ar) atmosphere, and then is cooled in the crucible and cast into ingots. The ingot constitutes the metal waste form (MWF). The ceramic waste form (CWF) has been developed to immobilize the active fission products (alkalis, alkaline earths, and rare earths) and transuranic elements of the electrolyte. The CWF is produced in a batch process by mixing and blending the waste salt, periodically removed from the electrorefiner, with zeolite 4A at 500 °C to occlude the waste-loaded salt within the cages of the zeolite crystal lattice. Salt-loaded zeolite is mixed with a borosilicate glass and consolidated at high temperature (850 to 900 °C) and pressure (14,500 to 25,000 psi) in a hot isostatic press (HIP) to make the final waste form. Pressureless sintering was investigated at ANL and may provide advantages over the HIP process during fabrication by giving a safer and easier pathway to volumetric scale-up of waste form fabrication. Further investigation of this potential capability, including the CWF produced using this process, is needed and is being pursued by ANL. Recommendation: Studies to compare the type, abundance, and radionuclide inventory of minor and trace phases between ceramic waste forms produced by pressureless sintering versus the HIP process should be given high priority in the post-demonstration phase. Alternatives to Electrometallurgical Technology for Treatment of DOE SNF In two of its reports the committee, as part of its fulfillment of its tasks, evaluated alternatives to electrometallurgical technology:9,10 direct disposal, glass material oxidation and dissolution, melt and dilute, PUREX, 9   National Research Council, An Assessment of Continued R&D into an Electrometallurgical Approach for Treating DOE Spent Nuclear Fuel, National Academy Press, Washington, D.C., 1995. In this report the committee considered spent fuel treatment alternatives to EMT within the context of all DOE SNF. 10   National Research Council, Electrometallurgical Techniques for DOE Spent Fuel Treatment: Spring 1998 Status Report on Argonne National Laboratory’s R&D Activity, National Academy Press, Washington, D.C., 1998. In this report the committee considered spent fuel treatment alternatives within the context of EBR-II SNF.

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Electrometallurgical Techniques for DOE Spent Fuel Treatment: Final Report chloride volatility, and plasma arc. While each of these technologies had merit, only PUREX was considered sufficiently well advanced to be a reasonable alternative to electrometallurgical treatment of spent fuel. However, the committee noted in its seventh report that public concern regarding transport of EBR-II spent fuel to the PUREX facility at the Savannah River site (SRS), combined with the expected shutdown of the PUREX canyons at SRS, argues against this alternative.11 WASTE STREAMS PRODUCED BY THE EMT PROCESS Waste Form Qualification The DOE, through its Office of Civilian Radioactive Waste Management (DOE-RW) and in conjunction with the development of final waste acceptance criteria to be based on Environmental Protection Agency/U.S. Nuclear Regulatory Commission regulations, is assessing the viability of permanent disposal of SNF in a deep geologic repository at Yucca Mountain, Nevada.12 The performance and compatibility of the ANL waste forms must be assessed within this system context of overall repository safety. The committee found that ANL’s waste qualification strategy is appropriately based on guidance provided in the memorandum of agreement (MOA) between DOE-RW and DOE’s Office of Environmental Management (DOE-EM). Waste Acceptance Product Specifications Data collected during the demonstration project provide information supporting issuance of an environmental impact statement (EIS) regarding continued application of the EMT process to the remaining inventory of EBR-II spent fuel. Thus, ANL has oriented its current activities to provide evidence of successful compliance with demonstration criteria.13 Preliminary testing and modeling of the performance of EMT waste forms under repository conditions were also initiated during the demonstration project. ANL’s waste acceptance product specifications (WAPS)14,15 are patterned after the quality assurance protocols used for Defense Program High-Level Waste (DHLW) borosilicate glass.16 The committee observes, however, that DHLW borosilicate glass has not received final qualification and acceptance for geologic disposal by DOE-RW. The committee understands that DOE is preparing waste acceptance criteria, including guidance on long-term waste form performance testing and qualification. This new document may modify the actual waste-acceptance strategies and waste-acceptance criteria that ANL’s EMT program is currently following. These final criteria will influence long-term testing of the EMT metal waste form and the ceramic waste form, HLW waste forms intended for final disposition in a geologic repository. 11   National Research Council, Electrometallurgical Techniques for DOE Spent Fuel Treatment: Spring 1998 Status Report on Argonne National Laboratory’s R&D Activity, National Academy Press, Washington, D.C., 1998, p. 20. 12   Office of Civilian Radioactive Waste Management, Viability Assessment of a Repository at Yucca Mountain, DOE/RW-0508, Department of Energy, Washington, D.C., 1998. 13   National Research Council, Electrometallurgical Techniques for DOE Spent Fuel Treatment: Fall 1996 Status Report on Argonne National Laboratory’s R&D Activity, National Academy Press, Washington, D.C., 1997. Also see Chapter 6 of this report. 14   T.P. O’Holleran, R.W. Benedict, and S.G.. Johnson, Waste Form Qualification Strategy for the Metal and Ceramic Waste Forms from Electrometallurgical Treatment of Spent Nuclear Fuel, NT Technical Memorandum No. 115, Argonne National Laboratory, Argonne, IL, 1999. 15   T.P. O’Holleran, D.P. Abraham, J.P. Ackerman, K.M. Goff, S.G. Johnson, and D.D. Keiser, Waste Acceptance Product Specifications for the Waste Forms from Electrometallurgical Treatment of Spent Nuclear Fuel, NT Technical Memorandum No. 116, Argonne National Laboratory, Argonne, IL, 1999. 16   Westinghouse Savannah River Company, DWPF Waste Acceptance Reference Manual (U), WSRC-IM-93-45, Savannah River Site, Aiken, SC, 1993.

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Electrometallurgical Techniques for DOE Spent Fuel Treatment: Final Report Metal Waste Forms The MWF test plan consists of attribute tests, characterization tests, accelerated tests, and service condition tests. Good progress to date seems to have been achieved in the identification of the various phases of stainless steel-15 zirconium (SS-15Zr)-type materials.17 The characterization tests that have been terminated showed either no corrosive attack or only minor tarnish. ANL plans a total of 856 tests as necessary to achieve the goals of the project. In the view of the committee this seems an excessive number. Final corrosion tests from the demonstration project showed corrosion rates that were very low, and no correlation of elemental leaching with alloy composition was found. Tests have not yet been completed with added uranium. Electrochemical corrosion testing data presented at the end of the demonstration showed corrosion rates of the MWF alloys in J-13 (simulated Yucca Mountain well water) and in solutions of pH = 2, 4, and 10 that were low and similar to those of SS316 and alloy C22. Finding: Some of the corrosion products, which may sequester radionuclides, might remain on the sample surface and might therefore not be detected by solution analysis. Recommendation: Surface analysis by X-ray photoelectron spectroscopy (XPS) or Auger electron spectroscopy (AES) should be continued in the post-demonstration phase for selected samples to determine the chemical composition of passivating films and/or corrosion products. According to ANL personnel in presentations to the committee,18 galvanic corrosion tests according to ASTM G71 have indicated that enhanced corrosion of SS-15Zr due to galvanic coupling of the MWF with the inner lining of the waste form container (assumed to be alloy C22) is not likely to be significant. Vapor hydration tests found that corrosion rates were greatly accelerated by exposure to steam. The chemical nature of the corrosion products is under investigation. The effect of radiation on corrosion behavior has been discussed only briefly in presentations to the committee by ANL personnel. The toxicity characteristic leaching procedure (TCLP) test data suggest that the MWF passes the TCLP test.19 Recommendation: In the post-demonstration phase, ANL personnel should subject a few carefully selected samples to additional evaluation by surface analysis to determine the chemical composition of the corrosion products. Recommendation: ANL personnel should concentrate on a few key samples, expose them at higher temperatures and chloride concentrations, and obtain electrochemical and surface analysis data. Waste form degradation/radionuclide release models have been established that are an integral part of ANL’s waste form repository performance assessment effort and will be used for predicting the long-term corrosion behavior of the MWF. Ceramic Waste Form To support repository qualification of the CWF, ANL developed a protocol and conducted a variety of tests and analyses relevant to the WAPS requirements. Detailed results and conclusions are contained in ANL’s 17   Material balance estimates for the MWF are given in Table 4.1. 18   Presentations by Daniel Abraham and Dennis D. Keiser, Jr., to the committee, ANL-W, July 21, 1999. 19   Presentation by Dennis D. Keiser, Jr., to the committee, ANL-W, July 21, 1999.

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Electrometallurgical Techniques for DOE Spent Fuel Treatment: Final Report Ceramic Waste Form Handbook.20 The CWF is a multiphase, nonhomogeneous composite consisting of approximately 75% sodalite, 25% borosilicate glass, and up to 5% other minor phases, e.g., aluminosilicates, rare-earth silicates, oxides, and halite (NaCl).21 The CWF repository qualification program is based on the adaptation of models and test protocols developed for DHLW borosilicate glass. Accelerated alpha damage testing was carried out on simulated CWF doped with 0.2 to 2.5 wt % 238Pu or 239Pu. Initial results from this ongoing study show no significant degradation of the waste after 6 months at relatively low doses.22 Recommendation: The electrometallurgical technology program should continue to investigate and evaluate in the post-demonstration period whether the test protocols and conceptual models developed for monolithic single-phase borosilicate glass can adequately represent the behavior of the nonhomogeneous multiphase CWF. A variety of tests that monitor corrosion behavior were conducted by ANL to achieve a basic understanding of the processes that control dissolution of the CWF. Dissolution tests on the CWF over a 6-month period indicated that the CWF dissolves at a rate equal to or less than that of reference high-level waste borosilicate glass. The minor component actinides and rare earths form phases separate from the sodalite and glass phases. The actinides occur as nano-size (colloidal) crystal inclusions associated with the glass or the glass/sodalite grain boundaries. Finding: It is possible that some of these colloidal-sized crystal inclusions may be leached from the grain boundaries and that some may become colloidal suspensions with mobility much greater than expected from their solubility.23 Several mechanical and physical properties of the CWF were determined: cracking factor, thermal stability, fracture toughness, and density. The mechanical and physical properties of the CWF are comparable to or better than those of borosilicate high-level waste glass. Good product consistency is achieved using the specified demonstration HIP process parameters. Finding: The physical and mechanical behavior of the CWF under repository conditions should be comparable to that of borosilicate high-level waste glass. Waste form performance has been modeled at ANL to predict the environmental impact of ANL’s ceramic and metal waste forms on the proposed repository at Yucca Mountain. The model must be refined and verified with experimental data. The committee found that the demonstration project success criteria (listed with the specific goals to meet them in Chapter 6) regarding the CWF have been met, although it is recognized that further data collection and 20   W.L. Ebert, D.W. Esh, S.M. Frank, K.M. Goff, M.C. Hash, S.G. Johnson, M.A. Lewis, L.R. Morss, T.L. Moschetti, T.P. O’Holleran, M.K. Richmann, W.P. Riley, Jr., L.J. Simpson, W. Sinkler, M.L. Stanley, C.D. Tatko, D.J. Wronkiewicz, J.P. Ackerman, K.A. Arbesman, K.J. Bateman, T.J. Battisti, D.G. Cummings, T. DiSanto, M.L. Gougar, K.L. Hirsche, S.E. Kaps, L. Leibowitz, J.S. Luo, M. Noy, H. Retzer, M.F. Simpson, D. Sun, A.R. Warren, and V.N. Zyryanov, Ceramic Waste Form Handbook, NT Technical Memorandum No. 119, Argonne National Laboratory, Argonne, IL, 1999. 21   Material balance estimates for the CWF are given in Table 4.1. 22   Presentation by S.G. Johnson and L.R. Morss to the committee, ANL-W, July 21, 1999. 23   For a study on the potential impact of actinides on repository performance, see A.B. Kersting; D.W. Efurd, D.L. Finnegan, D.J. Rokop, D.K. Smith, and J.L. Thompson, “Migration of Plutonium in Ground Water at the Nevada Test,” Nature, Vol. 397, 1999, pp. 56-59.

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Electrometallurgical Techniques for DOE Spent Fuel Treatment: Final Report analysis must be carried out in the post-demonstration period to support a final decision on CWF acceptability for repository disposal. Finding: The committee sees no significant barriers to successful demonstration of an acceptable CWF, although full testing will extend beyond the demonstration time frame. Recovered Uranium Material As part of the EMT process, uranium metal is recovered at the cathode in the electrorefiner. After separation from the electrolyte it is cast as uranium metal ingots. During the cathode processing step and/or the casting step, natural or depleted uranium (DU) is added to the highly enriched uranium (HEU) derived from the EBR-II driver fuel. The disposition options for recovered uranium material are constrained by several DOE programmatic decisions and environmental impact statements. The depleted uranium recovered from treatment of the EBR-II blanket fuel is currently limited to indefinite storage or disposal as a transuranic (TRU) waste. Finding: The current alternatives for disposition of uranium recovered from EBR-II fuel by electrorefining are limited to indefinite storage or speculative schemes for disposal. Recommendation: The DOE should evaluate and select among these existing options for the disposition of recovered uranium in a timely manner so that the overall impacts of the EMT approach can be assessed. POST-DEMONSTRATION ACTIVITIES If DOE chooses to use the EMT process to treat sodium-bonded SNFs in the DOE inventory, or any other spent fuels,24 ANL must complete all the activities required to qualify both the metal and ceramic waste baseline forms for repository disposal. ANL-E must also provide ongoing technical support to operations at ANL-W, and ANL-W must complete the required facility modifications and qualify the new, larger-scale equipment needed to handle the increased volume of fuel. These constitute a minimum set of post-demonstration activities. Post-demonstration qualification testing of EMT-produced waste forms must focus on the long-term rate of dissolution of the waste-form matrix, formation of radioactive element solubility-limiting solids, and potential formation of radionuclide-bearing colloids. Recommendation: In its post-demonstration activities, ANL should reevaluate the appropriateness and applicability of its overall model to address the dissolution behavior and the multiphase nature of the EMT waste forms, especially the CWF. Associated test protocols, including that for the current product consistency test (PCT), should also be reevaluated. A previous NRC report25 criticized leach rate as a measure of the long-term performance of waste forms under expected repository conditions. The post-demonstration evaluation of the long-term performance of EMT waste forms, especially the CWF, under repository conditions must address this aspect of solubility limits for radioelements. 24   Office of Civilian Radioactive Waste Management, A Roadmap for Developing Accelerator Transmutation of Waste (ATW) Technology: A Report to Congress, DOE/RW-0519, U.S. Department of Energy, Washington, D.C., 1999. 25   National Research Council, A Study of the Isolation System for Geologic Disposal of Radioactive Waste, National Academy Press, Washington, D.C., 1983.

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Electrometallurgical Techniques for DOE Spent Fuel Treatment: Final Report Recommendation: In the post-demonstration period, ANL should supplement and refine its current ASTM-based test protocols for waste form dissolution with respect to the technical perspectives on the long-term performance of the waste forms in geologic repositories, as described in the NRC’s 1983 report by the Waste Isolation System Panel (WISP).26 There is considerable concern regarding the potential for rapid migration of significant quantities of radionuclides, especially Pu, at Yucca Mountain via colloidal transport.27 The potential for formation and transport of radionuclide-bearing colloids should be specifically addressed in post-demonstration analysis and evaluation of EMT waste forms. The committee observes that there may be alternative, nontesting approaches to assessing the acceptability of EMT waste forms for geologic disposal and that the merits of these alternatives would have to be technically evaluated by the DOE and by other independent peer reviews. Recommendation: The eventual DOE waste acceptance criteria for geologic disposal should take into account available technical assessments. These waste acceptance criteria should be independently reviewed. Should DOE decide to treat the remaining sodium-bonded spent fuel inventory, continuing efforts would be required to increase the capacity of some process equipment and to modify the facilities at ANL. Recommendation: If the DOE decides to treat the remaining sodium-bonded spent fuel inventory and the waste form qualification efforts are successful, the required equipment upgrades and facility modifications should be adequately funded to ensure that treatment can be completed in a reasonable time and at a reasonable cost. The use of pressureless sintering to produce the ceramic waste form can offer distinct advantages over the baseline HIP process. The potential advantages include a higher throughput per square foot of process space, increased safety, and reduced costs. Recommendation: If pressureless sintering were to be used in place of the HIP process to produce the EMT ceramic waste form, waste form qualification studies would have to be conducted to determine its suitability for producing a waste form intended for deposit in a geologic repository. In the post-demonstration period, continued development of a high-throughput electrorefiner (HTER), particularly if it could be cost-effectively developed and implemented in a timely fashion, could offer the advantage of considerably reducing the time required to treat the remaining sodium-bonded fuel. There are at least two options for increasing throughput up through the electrorefiner step in the EMT process. The first is continued development and implementation of a HTER (e.g., the 25-inch HTER under development at ANL-E) with a uranium deposition rate significantly exceeding that of the current Mark-V design. The second option is to simply double the current electrorefiner deposition rate by adding a second Mark-V electrorefiner to the Ar cell at ANL-W. 26   National Research Council, A Study of the Isolation System for Geologic Disposal of Radioactive Wastes, National Academy Press, Washington, D.C., 1983. 27   Office of Civilian Radioactive Waste Management, Viability Assessment of a Repository at Yucca Mountain, DOE/RW-0508, U.S. Department of Energy, Washington, D.C., 1998.

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Electrometallurgical Techniques for DOE Spent Fuel Treatment: Final Report Recommendation: Continued development of a HTER should be evaluated in the context of the cost and time required for its development and implementation relative to the cost reduction that could be achieved by increasing the electrorefiner throughput by adding a second Mark-V and completing the inventory operations in the shorter time period. ANL-E is pursuing the development of a production-scale zeolite column to increase the loading of the electrorefiner salt to about 3 wt % plutonium by running the salt through a column composed of zeolite. The use of the zeolite column could provide enhanced extraction and immobilization of fission products and actinides relative to a batch process, but a number of significant technical challenges remain. The removal of water during the early stages of elution might prove to be an intractable problem that could prevent the successful development of a zeolite column compatible with the EMT process. Finding: The volume of sodium-bonded spent fuel waste generated using the “throw away salt” option, where a portion of the plutonium and fission-product-contaminated salt is mixed directly with zeolite and glass particles for waste disposal, is such a small fraction of the total waste destined for geologic disposal that waste volume reduction resulting from the use of the zeolite column would not have a significant impact on the overall waste disposal problem. Recommendation: Continued development of the zeolite column should not be considered a high priority unless a compelling argument can be made that its development and implementation would significantly reduce waste disposal costs or associated costs of EMT treatment of the DOE sodium-bonded spent fuel inventory. For EMT to be used to treat oxide fuels, a head-end step is required to convert the oxide to metal. ANL-E has been pursuing the use of lithium metal as a reducing agent in molten LiCl salts to effect this conversion. The committee concluded that the state of development of the lithium reduction head-end treatment step is fairly mature, and if it were allowed to go to completion, the DOE would have an additional option for treating uranium oxide spent nuclear fuel. Recommendation: If the DOE wants an additional option besides PUREX for treating uranium oxide spent nuclear fuel, it should seriously consider continued development and implementation of the lithium reduction step as a head-end process to EMT. ELECTROMETALLURGICAL TECHNOLOGY DEMONSTRATION PROJECT SUCCESS CRITERIA The criteria proposed by ANL in 1998 for the demonstration project were similar in scope to those recommended by the committee in 1995 but smaller in scale in order to conform to the revised environmental assessment. The four criteria address the process, the waste streams, and the safety of the electrometallurgical demonstration project. Finding: The committee finds that ANL has met all of the criteria developed for judging the success of its electrometallurgical demonstration project. Finding: The committee finds no technical barriers to the use of electrometallurgical technology to process the remainder of the EBR-II fuel.

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Electrometallurgical Techniques for DOE Spent Fuel Treatment: Final Report The EBR-II demonstration project has shown that the electrometallurgical technique can be used to treat sodium-bonded spent nuclear fuel. The major hurdle that remains is qualification of the waste forms from this processing. The total quantity of EBR-II spent nuclear fuel is relatively small, particularly in comparison to the total DOE spent fuel inventory, so even if qualification of the waste form were to prove impossible, the quantity of these materials that had been produced would be modest. The committee has found no significant technical barriers to the use of electrometallurgical technology to treat EBR-II spent fuel, and EMT therefore represents a potentially viable technology for DOE spent nuclear fuel treatment. However, before using EMT for processing other spent fuels in the DOE inventory, which would generate much larger amounts of these wastes than were produced in ANL’s demonstration project, it would be necessary for these waste forms to receive the acceptance qualification.