Appendix C
ANL Monthly Highlights of the Electrometallurgical Treatment Program
As a mechanism for keeping the committee informed of its progress in R&D, ANL has prepared brief monthly progress reports. These “Monthly Highlights” of the Electrometallurgical Treatment Program are submitted to DOE, which in turn provides them to the National Research Council for distribution to the committee. The period covered by the monthly reports reprinted in this appendix extends from May 1998 to October 1998.
Electrometallurgical Treatment Program Technical Highlights for May 1998
The electrometallurgical program has two components: the EBR-II Spent Fuel Treatment Demonstration project and its adaptation to a variety of DOE spent fuel types. The monthly highlights are divided between the main Work Breakdown Structure (WBS) elements for the project plus two additional tasks: treatment of oxide spent fuels and treatment of aluminum-based fuels. The technical highlights provide an overall picture of the program.
WBS 1.0 Treatment Operations
Electrometallurgical treatment technology will convert the highly enriched uranium and the reactive bond sodium in EBR-II fuel into low enriched uranium product, ceramic waste and metal waste. This work element involves the demonstration equipment operations in the Fuel Conditioning Facility (FCF) and the Hot Fuel Examination Facility (HFEF). The process steps will be operated in an integrated manner to demonstrate the economic and technical feasibility of the process with spent irradiated fuel.
WBS 1.1 Driver Treatment
The demonstration will treat 100 driver assemblies so that the processes can be demonstrated in an integrated system and fission product loading in the Mark-IV (Mk-IV) electrorefiner will reach three weight percent. At the end of May, 64 driver assemblies had been chopped and 60 driver assemblies had been introduced to the electrorefiner.
During May, six driver assemblies were chopped. The electrorefiner processed four driver assemblies. This batch was operated with the dual anode serial cathodes (see March Highlights) operating sequence. Two cathodes were produced from direct transport operating mode and two cathodes were produced from deposition mode. Twenty kilograms uranium was recovered and the entire batch operating time was 17 days. This rate is approximately twice the rate that is needed to be demonstrated.
The cathode processor converted two batches of electrorefiner cathodes (22 kg uranium metal) into uranium ingots. One of these batches investigated the use of an aerosol zirconia coating method for the process crucible. The aerosol coating is easier to apply remotely; however, the resultant coating does not perform as well as the standard spray coating method. This run confirmed the previous observations. One batch of cathode processor ingots was converted to a 40.5 kg low enriched uranium product.
WBS 1.2 Blanket Treatment
The EBR-11 blanket assemblies contain 47 kg of uranium each and will demonstrate high throughput rates in the Mark-V (Mk-V) electrorefiner. The internal readiness assessment and the DOE review for blanket operations was completed during the first week in May. A new, improved concentric anode cathode module was received and installed in-cell. The new anode-cathode module (ACM) changes the direction of rotation so that the uranium is removed from the cathode tube where a large area is available for settling into the product collection basket. During installation, mechanical alignment problems were encountered and special remotely operated tooling was fabricated to fix the problem. The improved ACM was ready to start
testing at the end of the month. Due to these problems, operations with irradiated blankets are forecast for late July.
WBS 1.3 Metal Waste
Three batches (two assemblies each) of irradiated cladding from driver treatment operations will be converted into typical metal waste forms for waste qualification. Chemical analysis of the metal waste ingots shows good agreement with the predicted composition from cladding hull samples. The analysis of material phases has started.
WBS 1.4 Ceramic Waste Operation with Irradiated Materials
After 100 driver assemblies are treated in the Mk-IV electrorefiner, a portion of the salt will be transferred to the HFEF where the salt and fission products will be immobilized in ceramic waste samples. This activity is not scheduled to begin until February 1999.
WBS 1.5 Facility Operations
Two driver assemblies were received from the Radioactive Scrap and Waste Facility. At the end of the reporting period, 17 driver and 21 blanket assemblies (25 irradiated, 1 unirradiated) were stored in the FCF air cell. One unirradiated blanket assembly was transferred to the argon cell for initial start-up testing of the blanket equipment.
WBS 2.0 Equipment and Facility Modifications
This work element covers the engineering design, fabrication, assembly and testing activities that are required to implement new process equipment, equipment improvements or facility modifications that support operations or development activities.
The new, modified concentric ACM was shipped to Argonne-West. The V-mixer heaters were modified to correct the reliability problems that were identified during process qualifications. Since the V-mixer was not operational, the hot isostatic press (HIP) was transferred to the mock-up area for remote handling/operations qualification. After the remote qualification is completed, the HIP will process the materials from the V-mixer.
WBS 3.0 Treatment Process Development
The key step in electrometallurgical treatment of spent nuclear fuel is electrorefining to separate pure uranium from the spent fuel, thus reducing the volume of high level waste. Treatment of large quantities of spent fuel, such as the EBR-II blanket, required development of the high-throughput electrorefiner. The Mk-V high-throughput design known as the ACM has been installed in the FCF at ANL-W for demonstration operations with the EBR-II blanket fuel. One of the Mk-V ACMs was prepared for testing at ANL-E, in the J-118 laboratory. Testing of the unit outside the glovebox was done to check for proper fit-up and operation. Fabrication and/or modifications were completed on all of the components needed for testing the Mk-V ACM in the glovebox. Some minor fabrication errors were detected and corrected, and initial shake-down testing and operations are expected to begin in June.
WBS 4.0 Process Modeling and Analysis
These activities develop and apply models to improve the understanding of various process steps; to help in design of equipment and selection of process variables; to evaluate the data on performance of the engineering-scale equipment; to provide support in planning of test campaigns; and to conduct operations.
The anode-cathode modeling status report, which was a DOE milestone, was issued in May 1998. This report describes the computational fluid dynamics analysis that is assisting in the understanding of the new high throughout electrorefiner.
WBS 5.0 Metal Waste Treatment Development
The noble metal fission products and undissolved cladding hulls are immobilized into a stainless steel-zirconium alloy for geologic repository disposal. In support of waste qualification activities, small samples of the metal waste are being produced so they can be characterized to establish the performance. Also,
equipment is being developed and tested in various laboratories to support design efforts on a larger casting furnace for inventory operations.
To examine candidate metal waste form crucible materials, a salt distillation experiment (SD-1) was conducted using three small-scale prototype crucibles and LiCl-KCl-UCl3 salt. Two of the crucibles were (proprietary) ceramics (MNP) and the third was a HfO2-15 wt % Y2O3 ceramic (HYO). All of the ceramics were formed on base refractory metals (i.e., Ta and Nb) and fired to form dense materials. The salt contained 4 wt % U. The salt was dark purple (almost black) because of the presence of the UCl3. After the test the crucibles were examined visually. All of the crucibles had become discolored, but there was no visual evidence of salt interactions in the two MNP crucibles. All of the crucibles were apparently salt free and only minimal mass changes were measured (+0.1 to 0.4 g increase). The HYO crucible seems to have “blistered ” a bit during the test. The results will be further analyzed, but the following may be said at this time: 1) The MNP crucibles are resistant to UCl3 salt attack under the present test conditions, 2) The HYO crucible may not have long-term resistance to salt attack, and 3) The UC13 content of the distilled salt is much lower than that of the charge salt. The change in salt composition is most likely due to vapor phase reactions with the Y2O3 secondary crucible, but crucible interactions cannot be excluded as a possibility at this time. Future testing and analysis will focus on proving the durability of the MNP crucible with uranium and UCl 3. Future Differential Scanning Calorimetry tests and prototype crucible tests are pending, according to the WBS schedule.
WBS 6.0 Metal Waste Qualification Testing
The metal waste form attributes and fission product release mechanisms and rates are being quantified to support repository performance modeling.
Thermophysical properties of the metal waste form are being determined by the Thermo-Physical Research Laboratory (TPRL), West Lafayette, IN, as part of the metal waste form test plan. The test data are detailed in the TPRL report # 1973. Both the specific heat and the coefficient of thermal expansion relationships show a change in slope in the range of 540 °C to 550 °C, and these changes are probably caused by a magnetic transition in the ferrite phase. The properties of the stainless steel-20 wt % zirconium alloy are very similar to the reference composition: stainless steel-15 wt % zirconium.
WBS 7.0 Ceramic Waste Treatment Development
The electrolyte salt is periodically removed from the electrorefiner and passed through a waste treatment system to immobilize fission products and transuranium elements for disposal. The necessary processes, materials, and demonstration equipment are being developed and tested so these waste treatment processes can be demonstrated in the HFEF with salts from the Mk-IV electrorefiner.
The electrolyte salt can be passed through a zeolite column to remove fission products and transuranium elements, which will allow the salt to be recycled to the electrorefiner. A series of ion-exchange tests are being done with several salt compositions, including salts that contain only one representative fission product and salts that are representative of electrorefiner salt after processing 300 driver assemblies. In these tests, zeolite beads that were pre-loaded with LiCl-KCl eutectic salt were placed in a mesh basket, attached to a stirrer, and immersed in the salt for contact times ranging from 0.5 to 72 hours. After this contact time, the zeolite was crushed and washed to remove surface salt; then samples were submitted for analytical chemistry and X-ray analysis. Although the chemical analyses are not yet available, the X-ray analyses show that the zeolite structure gradually changes over a period of 0.5 to 4 hours. The resulting structure appears to be a second zeolite. This transition to the second zeolite appears to be related to the degree of fission product loading as well as to the time of exposure to the salt phase. These changes do not appear to be harmful to the structure or durability of the final sodalite waste form, but it is important to understand the behavior of the zeolite during the ion exchange process.
The ceramic waste form incorporates the waste chloride salt, most of the fission products, and the transuranic elements. The reference fabrication process for the ceramic waste form includes hot isostatic pressing (HIP) to consolidate the glass-sodalite composite. An ambient-pressure sintering process is being developed, as an alternative to HIP, that offers the advantages of simplicity and semi-continuous production. In the most recent series of sintering tests, glass loadings of 35, 40, 45, and 50 wt % were
mixed with salt-loaded zeolite and sintered in air at 850 °C. Under these conditions, the zeolite is converted to sodalite. Three-day leach tests were done with these test samples, and cesium release was used as the performance indicator. The cesium releases from sintered and HIPed samples were found to be statistically indistinguishable over this range of glass loadings. The total mass losses and chloride releases were also comparable to HIPed samples at glass loadings of 40 wt % and higher. While it is too early to make process changes, based on these meager data, the ambient-pressure sintering method appears to be feasible, and it offers significant improvements for the waste form production process.
WBS 8.0 Ceramic Waste Qualification Testing
The ceramic waste form is being characterized so that its performance in different repository conditions and scenarios can be assessed. This work characterizes hot uniaxial pressing samples and laboratory scale and demonstration scale samples from hot isostatic pressing.
The goal of this task is to evaluate the applicability of standard durability tests for qualification of the ceramic waste form for disposition in a high-level waste repository and to provide ceramic waste form behavior testing and associated activities in support of its qualification. A report on ceramic waste form testing methods was completed on schedule and distributed to comply with project milestones outlined in section 11.2 of the Work Breakdown Structure for the EBR-II spent fuel demonstration project. All of the non-radioactive reference ceramic waste form samples that were to be produced by the Ceramic Waste Form Development Group for the qualification testing have been made and transferred. All of these samples have been removed from the stainless steel canisters, characterized, and prepared for testing. The characterization of these materials included visual inspection, X-ray diffraction analysis, density measurements, accessible free salt measurements, and 3-day MCC-1 tests in demineralized water at 90 °C. The bulk of the PCT and MCC-1 tests that will have durations up to a year have been initiated. Tests with longer durations will be initiated next month.
To date, four ceramic waste form samples with Pu have been made by hot uniaxial pressing and delivered for characterization testing at ANL-E. Three- and 28-day MCC-1 tests with demineralized water at 90 °C were performed with the Pu-containing samples. A good correlation between sample density and corrosion behavior was observed. The data suggest that slight changes in the processing conditions used at ANL-W may increase the durability of the hot uniaxially pressed samples. It appears that the retention of Pu by the reference ceramic waste form will be better than that of the lanthanum-borosilicate (LaBS) glass, although not quite as good as the titanate based waste forms.
WBS 9.0 Repository Performance Assessment Modeling
A significant element in establishing the viability of electrometallurgical treatment technology is a defensible assessment which shows that the wastes to be generated from the process will perform acceptably when ultimately disposed in a geologic repository.
The models for both the ceramic and metal waste form continue to be developed.
WBS 10.0 Environmental and Safety Support Tasks
These tasks provide the necessary safety analysis support for the electrometallurgical treatment demonstration activities.
The new Final Safety Analysis Report (FSAR) for the HFEF was reviewed and approved by the Operations Division Safety Review Committee. This FSAR is being prepared to meet current DOE guidelines.
Treatment of Oxide Spent Fuels: Oxide spent fuels cannot be electrorefined directly but must first be converted to metals. Lithium metal is used as the reducing agent in molten LiCl, and the Li2O dissolves in the molten salt. The engineering-scale test, ES-8, designed to test the operation of a large scale electrowinning cell and to evaluate corrosion of the vessel material, was completed this month. Salt from a previous reduction experiment was used to demonstrate compatibility between the reduction and electrowinning steps. The cell consisted of a porous metal cathode to collect lithium and a platinum anode to evolve oxygen. The operating potential was 3.7 V. The progress of the electrowinning was measured by analyzing for the lithium oxide concentration in the salt and by measuring the oxygen concentration in the
off-gas stream. Two cathodes were used sequentially during the experiment, and both cathodes were found to be loaded with lithium. Although further analysis of the data from this experiment is required, ES-8 was successful in demonstrating that the electrowinning process can be scaled up to significant sizes.
Platinum has been used as the anode material in most of the work to date on lithium metal recycle; however, because of the high cost of platinum, many potential replacement anode materials have been examined. Electrochemical cell tests have recently been completed, using a commercial electrode material from the glass industry, called “Stannex.” This doped tin oxide material was shown to be stable in the cell environment, and it is a catalytically active material with good electrical conductivity. The standard-grade material that was tested had an electrical conductivity that was lower than desired; nevertheless, it functioned well in producing lithium metal while reducing the Li2O content in the salt from 2 wt % to 1 wt %. The high-conductivity grade material will be specified for the process. Based on these results, high-conductivity “Stannex” is now the material of choice for replacing platinum as the anode material for the lithium electrowinning process.
Treatment of Aluminum-Based Fuels: Demonstration of the feasibility of electrometallurgical treatment of aluminum alloy spent fuels, such as foreign and domestic research reactor fuels, has been done in laboratory-scale experiments. The key step in treatment of this fuel is electrorefining of the aluminum, which represents about 90% of the spent fuel volume, and which can, after electrorefining, be discarded as low-level waste. Good separation of aluminum from uranium was demonstrated in previous work. Preparation of the engineering-scale aluminum electrorefiner and the electrolyte salt for beginning aluminum electrorefining is being done by modifying a small storage well in the J-118 glovebox. The well is being converted to a furnace well for testing the engineering-scale aluminum electrorefiner. Meanwhile, a short-term fluxing experiment was conducted to examine the extent to which active metal and rare earth fission products can be extracted from the U-Al-Si fuel alloy. Samples of the salt phase were taken and submitted for chemical analysis. They are expected to contain essentially all of the active metal and rare earth fission product inventory. A more extensive series of tests is planned in conjunction with a literature study of rare-earth/aluminum solutions.
Electrometallurgical Treatment Program Technical Highlights for June 1998
The electrometallurgical program has two components: the EBR-11 Spent Fuel Treatment Demonstration project and its adaptation to a variety of DOE spent fuel types. The monthly highlights are divided between the main Work Breakdown Structure (WBS) elements for the project plus two additional tasks: treatment of oxide spent fuels and treatment of aluminum-based fuels. The technical highlights provide an overall picture of the program.
WBS 1.0 Treatment Operations
Electrometallurgical treatment technology will convert highly enriched uranium and the reactive bond sodium in the EBR-II fuel into low enriched uranium product, ceramic waste and metal waste. This work element involves the demonstration equipment operations in the Fuel Conditioning Facility (FCF) and the Hot Fuel Examination Facility (HFEF). The process steps will be operated in an integrated manner to demonstrate the economic and technical feasibility of the process with spent irradiated fuel.
WBS 1.1 Driver Treatment
The demonstration will treat 100 driver assemblies so that the processes can be demonstrated in an integrated system and fission product loading in the Mark-IV (Mk-IV) electrorefiner will reach three weight percent. At the end of June, 70 driver assemblies had been chopped and 68 driver assemblies had been introduced to the electrorefiner.
During June, six driver assemblies were chopped. Two electrorefiner batches processed eight driver assemblies. Both batches were operated with the dual anode serial cathodes (see March Highlights) operating sequence. The objective was to finalize the operating conditions for the repeatability demonstration that is scheduled for this fall. The first run produced three cathodes (8.5, 1.9, and 2.0 kg).
The smaller second and third cathodes were consistent with previous run data. A data evaluation indicated that throughput and cathode size could be optimized by controlling the cathode potential at a higher voltage. During June, the first two cathodes (8.5 and 7.0 kg) from the second batch were produced from the direct transport operations. The third cathode from this batch will be produced in July; however, the initial data indicates that the cathode size can be increased by controlling at higher currents with higher cathode voltages. The electrorefiner control software will be modified to support this mode of operation.
The cathode processor converted one batch of electrorefiner cathodes (93 kg uranium metal) into uranium ingots. Two batches of cathode processor ingots were converted to low enriched uranium products (40.4 kg and 30.4 kg).
WBS 1.2 Blanket Treatment
The EBR-11 blanket assemblies contain 47 kg of uranium each and will demonstrate high throughput rates in the Mark-V (Mk-V) electrorefiner. The improved Mk-V anode-cathode module (ACM) was loaded into chopped unirradiated sodium bonded fuel. The ACM was operated at various process parameters to find the best conditions where mechanical jamming can be avoided. At the end of June, the ACM was removed from the electrorefiner so that a visual examination could be completed before the facility maintenance period, which is scheduled for the first two weeks in July.
WBS 1.3 Metal Waste
Three batches (two assemblies each) of irradiated cladding from driver treatment operations will be converted into typical metal waste forms for waste qualification. The second typical metal waste ingot was cast from irradiated driver cladding hulls. This ingot weighed 3.8 kg and had a good visual appearance. This run investigated new thermal heat shields and the elimination of additional stainless steel that was used to establish a metal pool. A sampling plan for the two demonstration ingots has been developed and initial core samples have been taken of the first ingot.
WBS 1.4 Ceramic Waste Operation with Irradiated Materials
After 100 driver assemblies are treated in the Mk-IV electrorefiner, a portion of the salt will be transferred to the HFEF where the salt and fission products will be immobilized in ceramic waste samples. This activity is not scheduled to begin until February 1999.
WBS 1.5 Facility Operations
Two driver assemblies were received from the Radioactive Scrap and Waste Facility. At the end of the reporting period, 14 driver and 26 blanket assemblies (25 irradiated, 1 unirradiated) were stored in the FCF air cell. One can containing 69 kg uranium product was shipped from FCF to interim storage. A facility maintenance period is scheduled for the first two weeks of July and process operations will not be performed during this maintenance time.
WBS 2.0 Equipment and Facility Modifications
This work element covers the engineering design, fabrication, assembly and testing activities that are required to implement new process equipment, equipment improvements or facility modifications that support operations or development activities.
The remote qualification of the hot isostatic press (HIP) was completed and the equipment was returned to the engineering laboratory to support ceramic waste process qualification activities.
WBS 3.0 Treatment Process Development
The key step in electrometallurgical treatment of spent nuclear fuel is electrorefining to separate pure uranium from the spent fuel, thus reducing the volume of high level waste. Treatment of large quantities of spent fuel, such as the EBR-11 blanket, required development of the high-throughput electrorefiner. The Mark-V high-throughput electrorefiner design known as the anode-cathode module (ACM) has been installed in the Fuel Conditioning Facility (FCF) at ANL-W for demonstration of electrometallurgical treatment of EBR-II blanket fuel. One of the Mark-V ACMs was prepared for testing at ANL-E, in the J-118 laboratory. This ACM was placed into the electrorefiner well and heated to 500 °C, with no fuel in the anode baskets, to test its operation at temperature. The anode basket assembly was rotated at 20 rpm for
two days, and the mechanical systems were found to perform well under these conditions. The ACM was then loaded with simulated fuel (unirradiated N-reactor fuel) and operated with an initial current of 200 A and a cut-off voltage of 0.45 V. After 280 A-h, the current had to be reduced to 150 A to avoid the cut-off voltage. At about 2100 A-h net current, the anode rotation locked up. The anode was easily freed by manually reversing the rotation, then continuing the operation. After another ~100 A-h, excessive vibration of the anode assembly was noted, accompanied by spikes in the rotation motor current. The unit was shut down for examination, and it was found that a fairly large piece of uranium had broken away from the cathode wall, which probably accounted for the initial blocking of rotation. Also, a large number of cylinder-shaped product pieces (~1/8-in dia by ~1/4-in long) were observed in the collection basket, indicating a “cutting ” or “rolling” action of the scrapers against the deposited uranium. These tests and operations will continue, to assist understanding of the operation of the system in the FCF.
WBS 4.0 Process Modeling and Analysis
These activities develop and apply models to improve the understanding of various process steps; to help in design of equipment and selection of process variables; to evaluate the data on performance of the engineering-scale equipment; to provide support in planning of test campaigns; and to conduct operations.
The computer files for the mass tracking system were created to incorporate the analytical chemistry results. These files will be updated in the on-line system during the July maintenance outage.
WBS 5.0 Metal Waste Treatment Development
The noble metal fission products and undissolved cladding hulls are immobilized into a stainless steel-zirconium alloy for geologic repository disposal. In support of waste qualification activities, small samples of the metal waste are being produced so they can be characterized to establish the performance. Also, equipment is being developed and tested in various laboratories to support design efforts on a larger casting furnace for inventory operations.
A salt distillation experiment (SD-1) was conducted using three small-scale prototype crucibles and LiCl-KCl-UCl3 salt. Two of the crucibles were (proprietary) ceramics (MNP) and the third was a HfO2-15 wt % Y2O3 ceramic (HYO). All of the ceramics were formed on base refractory metals (i.e., Ta and Nb) and fired to form dense materials. The crucibles were each charged with ~4 g of salt and loaded into an yttria crucible in an induction furnace. The crucibles were heated under vacuum to 1200 °C for 2 hours and then cooled. A vapor trap was used to collect the evaporated salt; interestingly, the distilled salt was very white, indicating the absence of UCl3.
After the test, the crucibles were examined visually. All of the crucibles had become discolored, but there was no visual evidence of salt interactions in the two MNP crucibles. All of the crucibles were apparently salt free and only minimal mass changes were measured (+0. 1 to 0.4 g increase). The HYO crucible seems to have “blistered ” a bit during the test. The results will be further analyzed, but the following may be said at this time: the MNP crucibles are resistant to UCl3 salt attack under the present test conditions (for a single test), the HYO crucible may not have long-term resistance to salt attack, and the UCl3 content of the distilled salt is much lower than that of the charge salt.
The change in salt composition is most likely due to vapor phase reactions with the Y2O3 secondary crucible, but crucible interactions cannot be excluded as a possibility at this time. Future testing and analysis will focus on proving the durability of the MNP crucible with uranium and UCl 3. Future Differential Scanning Calorimetry tests and prototype crucible tests are pending, according to our WBS schedule.
WBS 6.0 Metal Waste Qualification Testing
The metal waste form attributes and fission product release mechanisms and rates are being quantified to support repository performance modeling.
Neutron diffraction analyses have been completed on a stainless steel-15 wt % zirconium-2 wt % Tc (SS-15Zr-2Tc) alloy. The alloy ingot was cast in an yttrium-oxide crucible in the Hot Fuel Examination Facility (HFEF), located at ANL-W, and sectioned for microscopy, spectroscopy, diffraction and chemical
analyses. A portion of the ingot was sent to the Intense Pulsed Neutron Source (IPNS) for examination by neutron diffraction.
The microstructure of the SS-15Zr-2Tc alloy was examined by scanning electron microscopy (SEM) and energy dispersive spectroscopy (EDS). The microstructure was very similar to that obtained from a SS-15Zr alloy without Tc. The SEM showed an intermetallic phase, Zr(Fe,Cr,Ni) 2+x, and an iron solid solution phase, ferrite in the alloy. No discrete Tc-containing phases were observed. The Tc appeared to be homogeneously distributed in the alloy and present in both the ferrite and intermetallic phases.
Five phases were identified in the SS-15Zr-2Tc alloy: 1) Ferrite, 2) Austenite, 3) the C36 Zr(Fe,Cr,Ni)2+x, Laves polytype, 4) the C15 Zr(Fe, Cr, Ni)2+x, Laves polytype, and 5) (Fe,Cr,Ni)23Zr6. These phases are the same as those observed in a SS-15Zr alloy without Tc. However, the lattice parameters of all phases in the SS- 15Zr alloy are altered by Tc addition; this is apparently due to the incorporation of Tc into all the alloy phases.
WBS 7.0 Ceramic Waste Treatment Development
The electrolyte salt is periodically removed from the electrorefiner and passed through a waste treatment system to immobilize fission products and transuranium elements for disposal. The necessary processes, materials and demonstration equipment are being developed and tested so these waste treatment processes can be demonstrated in the HFEF with salts from the Mk-IV electrorefiner.
The ceramic waste form resulting from electrometallurgical treatment of spent nuclear fuels incorporates the chloride salt, most of the fission products, and the transuranic elements. The reference fabrication process for the ceramic waste form includes hot isostatic pressing (HIPing) to consolidate the glass-sodalite composite. An ambient-pressure sintering process is being developed, as an alternative to HIPing, that offers the advantages of simplicity and the possibility of semi-continuous production. A series of experiments were completed that were aimed at scaling the sintered waste form up to 6-in diameter. One set of samples, 3-in and 6-in diameter, was made using 35 wt % glass, and a second 3-in-diameter sample was made using 50 wt % glass. The sintered samples made with 35 wt % glass were not fully dense; however, the sample containing 50 wt % glass was found to be fully densified. The fully dense sample was sectioned and polished for examination and for leach testing. These and other experimental results from ambient-pressure sintering tests continue to appear promising; however, considerably more work is needed to show that this waste form is comparable to the reference glass-sodalite waste form prepared by HIPing.
WBS 8.0 Ceramic Waste Qualification TestingThe ceramic waste form is being characterized so that its performance in different repository conditions and scenarios can be assessed. This work characterizes hot uniaxial pressing samples and laboratory scale and demonstration scale samples from hot isostatic pressing.
The goal of this task is to evaluate the applicability of standard durability tests for qualification of the ceramic waste form for disposition in a high-level waste repository and to provide ceramic waste form behavior testing and associated activities in support of its qualification. Initial analysis of some of the samples revealed the presence of regions enriched with rare earth elements and phases enriched with barium and strontium and other phases enriched with barium and cerium. These newly observed results indicate that some fission products may be immobilized in phases other than sodalite. Furthermore, for the first time, pure rare earth phases, crystalline phases with rare earth elements, and barium and calcium sulfate species have also been observed in the reference ceramic waste form samples. The sulfate is probably a component in the clay binder used to make the granulated zeolite 4A starting material. Some alteration phases, including sulfate, cesium, and iodide bearing phases, were observed as residue in voids. This suggests that volatile water from the glass phase is forming the voids and acting as a transport mechanism for some soluble species in the ceramic waste form. Additional investigations are being performed to quantify the structures and corrosion properties of the specific phases that retain alkaline and rare earth elements. The identification of radionuclide bearing phases has direct applicability to the criteria outlined in the Waste Acceptance System Requirements Document and will be needed for the development of a mechanistic corrosion model.
Samples from seven hot isostatically pressed ceramic waste form canisters made at ANL-W were received as part of a blind characterization testing investigation. The initial suite of tests includes visual inspection, accessible salt measurements, X-ray diffraction examination, seven-day PCT, and three-day MCC-1 tests on material from each canister. The purposes of these tests are to identify the physical characteristics of the ANL-W samples and to determine their initial short-term corrosion behavior. The results from the initial tests indicate that the accessible salt measurement, PCT, and MCC-1 test are all sensitive to small changes in process variables and can be used to provide information about sample consistency.
Uranium-containing ceramic waste form samples were produced for qualification testing. These samples are being used in tests to determine the effect of uranium on the corrosion behavior of the ceramic waste form. Initial results from short-term MCC-1 tests with uranium-containing ceramic waste form samples indicate that the uranium does not adversely affect the short-term corrosion behavior of the waste form. Analysis of the unreacted uranium-containing samples with scanning electron microscopy and electron dispersive spectroscopy indicated that the uranium tends to behave similarly to the rare-earth elements and is often distributed in rare-earth phases.
WBS 9.0 Repository Performance Assessment Modeling
A significant element in establishing the viability of electrometallurgical treatment technology is a defensible assessment which shows that the wastes to be generated from the process will perform acceptably when ultimately disposed in a geologic repository.
The models for both the ceramic and metal waste form continue to be developed.
WBS 10.0 Environmental and Safety Support Tasks
These tasks provide the necessary safety analysis support for the electrometallurgical treatment demonstration activities.
The new Final Safety Analysis Report (FSAR) for the HFEF was reviewed and approved by the ANL Nuclear Safety Review Committee. The FSAR is ready for transmittal to DOE for their approval.
Treatment of Oxide Spent Fuels: Oxide spent fuels cannot be electrorefined directly but must first be converted to metals. Lithium metal is used as the reducing agent in molten LiCl, and the Li2O dissolves in the molten salt. A series of parametric studies are being done to better understand the kinetics of the oxide reduction process. The parameters that are being examined include the lithium surface area to salt volume ratio and the oxide fuel surface area to salt volume ratio. The rate of reduction was found, in earlier experiments, to increase with increasing lithium surface to salt volume ratio. Another series of experiments was started to investigate the effect of increasing the fuel mass (and surface area) to salt volume ratio, while keeping all other parameters constant. Preliminary results suggest that the overall rate of reduction actually decreased with increasing fuel mass. This decrease in reduction rate is apparently due to the limited rate at which lithium can be supplied to the UO 2 surface. It also suggests that scale-up will require increased lithium surface area to assure a high rate of fuel reduction. Additional experiments are being done to complete this parametric study and to develop a better understanding of the reduction process.
Treatment of Aluminum-Based Fuels: Demonstration of the feasibility of electro-metallurgical treatment of aluminum alloy spent fuels, such as foreign and domestic research reactor fuels, has been done in laboratory-scale experiments. The key step in treatment of this fuel is electrorefining of the aluminum, which represents about 90% of the spent fuel volume, and which can, after electrorefining, be discarded as low-level waste. Preparation of the engineering-scale aluminum electrorefiner and the electrolyte salt for beginning aluminum electrorefining is being done by modifying a small storage well in the J-118 glovebox. The well is being converted to a furnace well for testing the engineering-scale aluminum electrorefiner. Meanwhile, a short-term fluxing experiment was conducted to examine the extent to which active metal and rare earth fission products can be extracted from the U-Al-Si fuel alloy. Samples of the flux salt are expected to contain essentially all of the active metal and rare earth fission product inventory. A more extensive series of tests is planned in conjunction with a literature study of rare-earth/aluminum solutions.
Electrometallurgical Treatment Program Technical Highlights for July 1998
The electrometallurgical program has two components: the EBR-II Spent Fuel Treatment Demonstration project and its adaptation to a variety of DOE spent fuel types. The monthly highlights are divided between the main Work Breakdown Structure (WBS) elements for the project plus two additional tasks: treatment of oxide spent fuels and treatment of aluminum-based fuels. The technical highlights provide an overall picture of the program.
WBS 1.0 Treatment OperationsElectrometallurgical treatment technology will convert the highly enriched uranium and the reactive bond sodium in EBR-II fuel into low enriched uranium product, ceramic waste and metal waste. This work element involves the demonstration equipment operations in the Fuel Conditioning Facility (FCF) and the Hot Fuel Examination Facility (HFEF). The process steps will be operated in an integrated manner to demonstrate the economic and technical feasibility of the process with spent irradiated fuel.
WBS 1.1 Driver Treatment
The demonstration will treat 100 driver assemblies so that the processes can be demonstrated in an integrated system and fission product loading in the Mark-IV (Mk-IV) electrorefiner will reach 3 wt %. At the end of July, 72 driver assemblies had been chopped and 70 driver assemblies had been introduced to the electrorefiner.
During July, two driver assemblies were chopped. The electrorefiner produced a cathode via a deposition run from the cadmium pool. This cathode completed the June batch (4 assemblies) by dual anode serial cathodes. This batch was a test of the proposed operating condition for the three month repeatability demonstration. At the end of July, a batch (2 assemblies) of driver fuel was being processed by alternating direct transport-deposition. This new proposed mode of operation is attempting to match the product output with the fuel input so the number of operations can be minimized and throughput is increased. Two experiments are planned to test this operating mode.
The cathode processor converted one batch of electrorefiner cathodes (12.7 kg uranium metal) into uranium ingots.
WBS 1.2 Blanket Treatment
The EBR-II blanket assemblies contain 47 kg of uranium each and will demonstrate high throughput rates in the Mark-V (Mk-V) electrorefiner. The improved Mk-V anode-cathode module (ACM), which was loaded with chopped unirradiated sodium bonded fuel, was removed from the Mk-V electrorefiner. The initial examination showed that the product collector had collected significant product (17.8 kg); however, the anode baskets could not be removed from the cathode tubes. Since electrotransport operations could continue during the facility maintenance period, the recent data from the high throughput development tests (see this month's WBS 3.0 highlights) was used to select operating conditions that would clean the dense uranium phase off the cathode tube. These operating conditions, which use a high current to strip the cathode deposits back to the anode basket, did successfully clean the cathode tubes and the unit was disassembled. The product collector contained 19.8 kg of product and only a small amount of uranium remained in the anode baskets. One complication was the cathode scrapers, which are beryllia, were broken in one region of the anode baskets. Even though the exact time of this breakage could not be identified, the process data indicate that it probably occurred in the early part of the run. This run showed the deposit on the cathode walls can be controlled by the right set of operating conditions. This positive result plus the development tests results demonstrate that blanket processing could be started.
WBS 1.3 Metal Waste
Three batches (two assemblies each) of irradiated cladding from driver treatment operations will be converted into typical metal waste forms for waste qualification. Preparations were started so the metal waste ingots could be core-drilled for waste qualification activities.
WBS 1.4 Ceramic Waste Operation with Irradiated Materials
After 100 driver assemblies are treated in the Mark-IV electrorefiner, a portion of the salt will be transferred to the HFEF where the salt and fission products will be immobilized in ceramic waste samples. This activity is not scheduled to begin until February 1999.
WBS 1.5 Facility Operations
Two driver assemblies were received from the Radioactive Scrap and Waste Facility. At the end of the reporting period, 14 driver and 26 blanket assemblies (25 irradiated, 1 unirradiated) were stored in the FCF air cell. The facility maintenance period was completed during the first two weeks of July and process operations were limited during this maintenance time.
WBS 2.0 Equipment and Facility Modifications
This work element covers the engineering design, fabrication, assembly and testing activities that are required to implement new process equipment, equipment improvements or facility modifications that support operations or development activities.
Modifications of the salt and zeolite grinder were started. These modifications will address remote handling issues that were identified during the glovebox testing.
WBS 3.0 Treatment Process Development
The key step in electrometallurgical treatment of spent nuclear fuel is electrorefining to separate pure uranium from the spent fuel, thus reducing the volume of high level waste. Treatment of large quantities of spent fuel, such as the EBR-II blanket, required development of the high-throughput electrorefiner. The Mark-V high-throughput electrorefiner design known as the anode-cathode module (ACM) was installed in the Fuel Conditioning Facility (FCF) at ANL-W for demonstration of electrometallurgical treatment of EBR-II blanket fuel. One of the Mark-V ACMs is being tested at ANL-E, in the J-118 laboratory. A stripper cathode was used to remove the dense uranium deposit from the cathode tube of the ACM. The stripper cathode consisted of several steel rods that were used to replace the anode baskets. The polarity of the electrorefiner was reversed so that the concentric tubes with their uranium deposit became the anodes, and the steel rods became cathodes. During operation, the steel rods were rotated at 40 rpm through the concentric passages, and a current of 400 A was passed for one hour. Post-test examination showed that all of the uranium was removed from the concentric tubes. Most of the uranium deposited on the steel rods was washed off and collected in the product collection basket. About 1.3 kg uranium was collected in the basket, while about 300 g remained on the steel rods. This test showed that the stripper cathode can be used to remove the dense uranium deposit from the cathode tubes, if necessary, and that the design and operating method used in this test were effective in performing this function.
The ACM has now operated continuously for over 5500 amp-hours, and it is estimated to have produced about 16 kg of uranium so far. This successful sustained operation was achieved using the following operating parameters: 1) No fuel was loaded into one of the baskets in each channel, 2) The anode rotation direction is reversed briefly after one hour of forward operation, and 3) The cathode deposit was stripped off using a 600-A reverse current for five minutes once each hour. This successful performance shows that operating conditions have been found for sustained operation of the Mark-V ACM. Additional tests will be made to determine which parameters are essential for sustained operation and to find any design changes that are needed to increase the uranium throughput rate.
WBS 4.0 Process Modeling and Analysis
These activities develop and apply models to improve the understanding of various process steps; to help in design of equipment and selection of process variables; to evaluate the data on performance of the engineering-scale equipment; to provide support in planning of test campaigns; and to conduct operations.
Mass tracking files were updated in the on-line system during the July maintenance outage.
WBS 5.0 Metal Waste Treatment Development
The noble metal fission products and undissolved cladding hulls are immobilized into a stainless steel-zirconium alloy for geologic repository disposal. In support of waste qualification activities, small samples
of the metal waste are being produced so they can be characterized to establish the performance. Also, equipment is being developed and tested in various laboratories to support design efforts on a larger casting furnace for inventory operations.
The spent fuel cladding hulls, zirconium, and noble metal fission products remain in the anode basket after separating the uranium by electrorefining. These cladding hulls are melted, together with residual zirconium and noble metal fission products, to make the metal waste form. A “product consistency test,” in the vein of that used for glass high-level waste forms, is needed to assure the acceptability of the metal waste form. High-temperature immersion (autoclave) tests were conducted at 200 °C for 28 days, as a possible quick test to serve this purpose. Although the weight losses (and gains) were small, all of the specimens exhibited some measurable surface corrosion. For example, the specimens having 15 to 20 wt % zirconium showed uniform corrosion and weight gains on the order of 0.1 mg. The specimens having 5 wt % zirconium showed some evidence of preferential attack and weight losses on the order of 0.3 mg. The chemical analyses of the leach solutions are not yet available; however, it is evident from these tests that differences in alloy composition result in differences in corrosion behavior. These results suggest that the 200 °C, 28-d immersion test may be useful as a product consistency test for the metal waste form.
WBS 6.0 Metal Waste Qualification Testing
The metal waste form attributes and fission product release mechanisms and rates are being quantified to support repository performance modeling. Metal waste samples containing technetium are being tested by the pulsed flow leach test, which provides an estimate of the forward reaction release rate. Currently, fourteen samples are being tested and the initial data indicates that technetium is released at a rate that is one thousand times less than the release from high level waste glasses. The technetium sample phase characterization and release rate data are being summarized in two memos that will fulfill the status report of the technetium behavior in metal waste milestones. This milestone is expected to be completed by the end of September. Additionally, eight samples containing uranium are being characterized by the pulsed flow leach test.
WBS 7.0 Ceramic Waste Treatment Development
The electrolyte salt is periodically removed from the electrorefiner and passed through a waste treatment system to immobilize fission products and transuranium elements for disposal. The necessary processes, materials and demonstration equipment are being developed and tested so these waste treatment processes can be demonstrated in the HFEF with salts from the Mk-IV electrorefiner.
The ceramic waste form resulting from electrometallurgical treatment of spent nuclear fuels incorporates the chloride salt, most of the fission products, and the transuranic elements. The reference process for fabrication of this ceramic waste form includes hot isostatic pressing (HIPing). Ambient-pressure sintering is being explored as an alternative to HIPing, because it provides the possibility of lower cost and much higher throughput rates. The sintering process is being scaled up to 3-in and 6-in diameter bodies that are fabricated on graphite-coated stainless steel setters. A series of these samples were sintered at 850 °C with hold times of 1, 2, 3, and 4 hours. At hold times of three hours and over, the waste forms achieved densities near theoretical and open porosities less than 1%. When sodalite is exposed to these high temperatures for long times, it is possible that it could convert to the non-zeolitic mineral, nepheline and, thus, lose its capacity to retain salt. X-ray diffraction analyses confirmed that even the 4-hour sintered samples remained as sodalite and had not converted to nepheline. While these results are too preliminary to justify changing the reference waste form fabrication method to ambient-pressure sintering, this alternative method continues to appear very promising. Work is continuing on this fabrication method as a high priority.
Granulated zeolite, the reference material for use in the demonstration, was first processed in the heated V-mixer. The free chloride value (measure of mixing and salt occlusion into the zeolite) obtained from this run, 0.011 weight percent, was the lowest to date. A V-mixer run to simulate processing material from a zeolite column was also performed. To support the production of ceramic waste form samples containing 238Pu for accelerated alpha decay studies, two runs to produce 238Pu loaded samples were made. Only one additional run is expected before operations with 238Pu commence.
WBS 8.0 Ceramic Waste Qualification Testing
The ceramic waste form is being characterized so that its performance in different repository conditions and scenarios can be assessed. This work characterizes hot uniaxial pressing samples and laboratory scale and demonstration scale samples from hot isostatic pressing.
The goal of this task is to evaluate the applicability of standard durability tests for qualification of the ceramic waste form for disposition in a high-level waste repository and to provide ceramic waste form behavior testing and associated activities in support of its qualification. Solution exchange corrosion tests were performed with the ceramic waste form and with its potential base constituents of glass, zeolite 5A, and sodalite. These solution exchange tests were performed at 90 °C by replacing 80 to 90% of the leachate with fresh demineralized water or Mg-Na-K-Cl brine after set time intervals. The results from these tests provide information about the intrinsic ability of the ceramic waste form and its constituent materials to retain waste components. The results from solution exchange tests with demineralized water, reported in a previous monthly report, indicated that radionuclides are retained in the pure zeolite and in the ceramic waste form to a greater degree than cations like Li, K, and Na, during exposure to low ionic strength solutions. Results from solution exchange tests with candidate ceramic waste forms with zeolite 5A and tests with the constituent materials alone were done to determine the corrosion behavior of each component in a Mg-Na-K-Cl brine. This brine was used to investigate the ion-exchange behavior of the ceramic waste form in the presence of high ionic strength solutions. The results indicate that in these high ionic strength solutions, ion exchange does occur between components of the brine and fission products in the zeolite. The degree of ion exchange is greater in tests with zeolite 5A than in tests with sodalite. Sodalite still retains Ce preferentially to the other fission products even in a brine solution.
WBS 9.0 Repository Performance Assessment Modeling
A significant element in establishing the viability of electrometallurgical treatment technology is a defensible assessment which shows that the wastes to be generated from the process will perform acceptably when ultimately disposed in a geological repository.
The analysis of the June 1998 WBS 9.0 DOE milestone, “Complete Waste Packaging Parametric Study with Thermal Loading,” was completed and the results are being compiled in a Waste Form Production report which will be issued as a draft by the end of August. The draft Yucca Mountain Project document, “Viability Assessment of a Repository at Yucca Mountain,” was reviewed and comments submitted to the National Spent Nuclear Fuel Program for inclusion in the overall DOE review of this document. Particular attention was given to identifying items relevant to placement of the FCF waste forms into the Yucca Mountain repository.
WBS 10.0 Environmental and Safety Support Tasks
These tasks provide the necessary safety analysis support for the electrometallurgical treatment demonstration activities.
Possible methods to simplify the Technical Safety Requirements for the Fuel Conditioning Facility were discussed by operating personnel and safety analysts. The results of these discussions will be documented by the end of September so this milestone was rescheduled.
Treatment of Oxide Spent Fuels: Oxide spent fuels cannot be electrorefined directly but must first be converted to metals. As part of the program to explore treatment of degraded spent nuclear fuels, and for the purpose of testing the ability of the electrometallurgical treatment process to handle finely divided uranium, some uranium hydride obtained from Ames Laboratory was decomposed by heating. The finely divided UH3 was placed in a platinum crucible and heated to 450 °C in a high-purity helium atmosphere glove-box furnace well. The resulting material was not the expected finely divided uranium powder but an easily crushed ash-like material having the same size and shape as the original charge. The material had a two-tone gray color. A sample was taken and crushed for X-ray diffraction (XRD) analysis, but the results are not yet available. The weight change of the material indicated that only half of the hydrogen had been released. Additional processing at 450 °C or higher may be necessary to achieve full decomposition of the hydride. Direct electrorefining, with hydrogen evolution, will be attempted with other samples of the UH3 to assess the feasibility of that approach.
Treatment of Aluminum-Based Fuels: Demonstration of the feasibility of electrometallurgical treatment of aluminum alloy spent fuels, such as foreign and domestic research reactor fuels, has been done in laboratory-scale experiments. The key step in treatment of this fuel is electrorefining of the aluminum, which represents about 90% of the spent fuel volume, and which can, after electrorefining, be discarded as low-level waste. Preparation of the engineering-scale aluminum electrorefiner is being done by converting a small storage well in the J-118 glovebox to a furnace well for testing the engineering-scale aluminum electrorefiner. All of the components needed for this test are near completion, and assembly in the glovebox is ready to begin.
Electrometallurgical Treatment Program Technical Highlights for August 1998
The electrometallurgical program has two components: the EBR-II Spent Fuel Treatment Demonstration project and its adaptation to a variety of DOE spent fuel types. The monthly highlights are divided between the main Work Breakdown Structure (WBS) elements for the project plus two additional tasks: treatment of oxide spent fuels and treatment of aluminum-based fuels. The technical highlights provide an overall picture of the program.
WBS 1.0 Treatment Operations
Electrometallurgical treatment technology will convert the highly enriched uranium and the reactive bond sodium in EBR-II fuel into low enriched uranium product, ceramic waste and metal waste. This work element involves the demonstration equipment operations in the Fuel Conditioning Facility (FCF) and the Hot Fuel Examination Facility (HFEF). The process steps will be operated in an integrated manner to demonstrate the economic and technical feasibility of the process with spent irradiated fuel.
WBS 1.1 Driver Treatment
The demonstration will treat 100 driver assemblies so that the processes can be demonstrated in an integrated system and fission product loading in the Mark-IV (Mk-IV) electrorefiner will reach three weight percent. At the end of August, 75 driver assemblies had been chopped and 72 driver assemblies had been introduced to the electrorefiner.
During August three driver assemblies were chopped. Two low-enriched uranium ingots weighing a total of 79.2 kg were produced in the casting furnace. One 15.7 kg ingot was produced in the cathode processor.
Two electrorefining experiments were performed with one four-assembly batch of fuel. The goal of these experiments was to increase the deposit mass by using a novel power supply configuration. The power supply was alternately cycled between direct transport (fuel dissolution baskets as anode) and deposition (cadmium pool as anode) during both of the experiments. The time-span and frequency in the deposition part of the cycle were varied to make two experiments. Repeated switching of the power supply was intended to emulate electrical shorting between fuel dissolution baskets and the electrorefiner vessel. Two previous experiments involving this kind of electrical shorting yielded deposits containing an amount of uranium comparable to the charge of uranium in the fuel dissolution baskets. The two August experiments produced deposits weighing 7.7 kg and 7.0 kg—less than the goal range of 9.5 kg to 10 kg. A third run, simple deposition from the cadmium pool, was executed to reduce the uranium metal content of the electrorefiner. The deposit weighed 4 kg, which was the goal.
WBS 1.2 Blanket Treatment
The EBR-II blanket assemblies contain 47 kg of uranium each and will demonstrate high throughput rates in the Mark-V (Mk-V) electrorefiner.
During August, the first experiment with irradiated blanket fuel was conducted in the Mk-V electrorefiner. Seven blanket fuel elements (17.5 kg uranium) were loaded into seven of the nine fuel baskets of the anode. One fuel basket was left empty in each of the anode channels. This loading was intended to assist dislodging the uranium deposit on the cathode tube walls during direct transport. A new stripping program was employed to prevent the buildup of a dense uranium phase on the cathode tube surfaces. The experimental conditions duplicated a successful test performed at Argonne East, with similar results.
Following the eight-day test, the uranium was substantively exhausted from the chopper fuel segments. The net weight of the hulls was 400 g (salt plus metal), and the net weight of the product collector was 13.3 kg (salt plus refined uranium metal). The net weight of the cathode tubes was 7.1 kg, and some uranium metal was found lodged up inside the cathode tube, below the anode baskets. The test was successful in that 17.5 kg of uranium was electrorefined with very little jamming of the anode-cathode. When jamming did occur, the uranium was readily recovered.
WBS 1.3 Metal Waste
Three batches (two assemblies each) of irradiated cladding from driver treatment operations will be converted into typical metal waste forms for waste qualification. In support of waste qualification activities, the metal waste ingots were sampled by core-drilling.
WBS 1.4 Ceramic Waste Operation with Irradiated Materials
After 100 driver assemblies are treated in the Mk-IV electrorefiner, a portion of the salt will be transferred to the HFEF where the salt and fission products will be immobilized in ceramic waste samples. This activity was originally scheduled to begin in February 1999; however, the process development testing (WBS 7.0) has encountered several problems. Currently the forecasted start of these activities is March 1999.
WBS 1.5 Facility Operations
Four driver assemblies were received from the Radioactive Scrap and Waste Facility. At the end of the reporting period, 14 driver and 26 blanket assemblies (25 irradiated, 1 unirradiated) were stored in the FCF air cell. Low enriched uranium product (120 kg) was shipped to the ZPPR storage facility. Currently a total of 422 kg of product are in interim storage with an additional 223 kg in FCF.
WBS 2.0 Equipment and Facility Modifications
This work element covers the engineering design, fabrication, assembly and testing activities that are required to implement new process equipment, equipment improvements or facility modifications that support operations or development activities.
Initial modifications of the salt and zeolite grinder were completed and initial testing shows the grinder can be operated to produce the coarser salt that will improve the V-mixer operations. The grinder was being reinstalled in the ceramic waste glovebox.
WBS 3.0 Treatment Process Development
The key step in electrometallurgical treatment of spent nuclear fuel is electrorefining to separate pure uranium from the spent fuel, thus reducing the volume of high level waste. Treatment of large quantities of spent fuel, such as the EBR-II blanket, required development of the high-throughput electrorefiner.
The high-throughput electrorefiner (HTER) was developed for rapid treatment of large quantities of spent fuel, specifically the EBR-II blanket fuel. One version of the HTER, known as the anode-cathode module (ACM), is now being used to treat the EBR-II blanket fuel in the FCF. The uranium product from the ACM is collected in a screen basket at the bottom, and it contains between 5 and 20 wt % salt. The upper 70% of the product is low in salt, and it is easily removed from the collection basket; however, the lower 30% contains about 20 wt % salt, and it is very difficult to recover without damaging the screens. The ACM that is being tested at ANL-E was used to solve this problem by heating the product to melt the salt. The collection basket was inverted, placed in the hot gas space in the electrorefiner, and heated for about six hours to melt the salt phase. The ~5 kg of uranium product remaining in the basket dropped from the collection basket into a receiver crucible. After this operation, nearly all of the uranium product had been removed from the collection basket, and the basket was free of damage and ready for the next test. This procedure can now be incorporated into operation of the ACM being used to treat the EBR-II blanket fuel in the FCF.
WBS 4.0 Process Modeling and Analysis
These activities develop and apply models to improve the understanding of various process steps; to help in design of equipment and selection of process variables; to evaluate the data on performance of the engineering-scale equipment; to provide support in planning of test campaigns; and to conduct operations.
The Nuclear Technology report on the ASIDE code for radioactive decay has been completed. The ASIDE code remedied shortcomings in the ORIGEN code and its derivatives, including mass conservation for all isotopes and rapid execution time. The ASIDE code has been implemented in the MTG. An evaluation of the necessity for performing radioactive decay of material for every processing step in FCF has also been completed, with the conclusion that only performing the radioactive decay of the product at the completion of processing before shipping from the FCF would provide adequate accuracy. An evaluation of the effect on substitution of measured data is being done.
WBS 5.0 Metal Waste Treatment Development
The noble metal fission products and undissolved cladding hulls are immobilized into a stainless steel-zirconium alloy for geologic repository disposal. In support of waste qualification activities, small samples of the metal waste are being produced so they can be characterized to establish the performance. Also, equipment is being developed and tested in various laboratories to support design efforts on a larger casting furnace for inventory operations.
The spent fuel cladding hulls, zirconium, and noble metal fission products remain in the anode basket after separating the uranium by electrorefining. These cladding hulls are melted, together with residual zirconium and noble metal fission products, to make the metal waste form. Any residual uranium not transported during electrorefining remains in the anode basket and becomes part of the metal waste form. Neutron diffraction analyses were completed on a SS-15Zr-5U alloy that had been cast in the Hot Fuel Examination Facility at ANL-W. The lattice changes that occurred as a result of the presence of uranium suggest that uranium is present only in the intermetallic phases of the alloy. Small lattice contractions were observed in the ferrite and austenite phases, suggesting a minor depletion of chromium from these phases. Lattice expansions in the Laves intermetallic phases may be due to substitution of uranium atoms at the Fe sites of the ZrFe2 lattice, whereas the lattice contraction of the Fe23Zr6 intermetallic phase suggests U substitution at the Zr sites. The quantities of ferrite and austenite in the alloy are comparable whether uranium is present or not. Significant differences are observed, however, in the quantities of intermetallic phases. The amounts of C15 and Fe23Zr6 are larger and the amount of C36 is smaller in the alloy that contains uranium. None of these relatively minor changes in the structure of the metal waste form is expected to affect the corrosion rate nor its performance in the geologic repository.
WBS 6.0 Metal Waste Qualification Testing
The metal waste form attributes and fission product release mechanisms and rates are being quantified to support repository performance modeling. A total of 29 spiked metal waste form samples are currently undergoing long term immersion testing with four of these started in August. These tests will yield forward reaction rates for the release of technetium and uranium in simulated ground water. The results to date have been very encouraging; i.e., very little release has been observed at immersion times exceeding 500 days. Samples from two actual metal waste ingots cast in FCF have been obtained. A battery of tests will be performed over the next several months on these samples. The tests include elemental analysis, immersion tests, microscopy and the EPA Toxic Characteristic Leaching Procedure (TCLP).
WBS 7.0 Ceramic Waste Treatment Development
The electrolyte salt is periodically removed from the electrorefiner and passes through a waste treatment system to immobilize fission products and transuranium elements for disposal. The necessary processes, materials and demonstration equipment are being developed and tested so these waste treatment processes can be demonstrated in the HFEF with salts from the Mk-IV electrorefiner.
The ceramic waste form resulting from electrometallurgical treatment of spent nuclear fuels incorporates the chloride salt, most of the fission products, and the transuranic elements. The electrolyte salt that contains fission products and transuranic elements is absorbed into the zeolite structure in preparation for making the ceramic waste form. This absorption process is done with a blender that mixes the zeolite
powder and salt at 500 °C. The time required to achieve effective blending has recently been systematically studied, at blending times ranging from 4 to 20 hours, in an attempt to determine the minimum time needed for blending. After blending, the mixtures were sampled and the samples were washed in water to determine the amount of free chloride remaining outside the zeolite structure. These test showed that a blending time of 11 hours is required to reach a free chloride level of 0.05% and 16 hours is required to reach the lowest level achieved, 0.02 wt %. The bulk of the salt was absorbed quickly at the beginning of the test; however, absorption of the last few tenths of percent salt required about 2/3 of the total blending time.
In an attempt to more accurately measure the amount of salt released during leach testing of the ceramic waste form, tests were done in which the initial washing time for leach samples was reduced from the standard 20 min to 2 min. The 20-min washings were carried out by changing the ethanol wash solution every 5 min for 4 wash cycles. There were no ethanol changes for the 2-min wash. All samples were leach tested at 90 °C for three days in deionized water. It was expected that the chloride release would be greater for the samples washed 2 min than for those washed 20 min; however, this was not the case for all the samples. Five of the seven samples showed more chloride release for the 20-min washed samples. From a statistical point of view, the samples washed for 20 min prior to testing showed only a slightly lower chloride release than those samples washed for only 2 min. The shorter wash cycle will likely be adopted to provide a more conservative evaluation of chloride release from the ceramic waste form.
Tests were initiated using the reference, as-procured, granulated zeolite from an outside vendor (UOP) in the demonstration-scale equipment. Problems were encountered with segregation of salt and zeolite in the blending operation in the V-mixer. The reason for the segregation was determined to be the different particle sizes of the salt and zeolite. Modification to the salt grinding operation are being implemented to remedy this problem.
In order to determine product quality and consistency in production operations, plans have been developed to run small-scale samples simultaneously with HIP cans. The small samples, called witness tubes, will be analyzed to assess the quality of the large products without destructive analysis of the ceramic in the HIP cans. The first of these witness tubes was processed with a demonstration-scale HIP can. Analysis is ongoing to determine the validity of this method to assess product quality.
For the accelerated alpha decay tests, the final samples containing 239Pu were produced in preparation for the work with 238Pu. The 238Pu operations will begin in September.
WBS 8.0 Ceramic Waste Qualification Testing
The ceramic waste form is being characterized so that its performance in different repository conditions and scenarios can be assessed. This work characterizes hot uniaxial pressing samples and laboratory scale and demonstration scale samples from hot isostatic pressing.
The goal of this task is to evaluate the applicability of standard durability tests for qualification of the ceramic waste form for disposition in a high-level waste repository and to provide ceramic waste form behavior testing. Solution exchange corrosion tests were performed with the ceramic waste form and with its potential base constituents of glass, zeolite 5A, and sodalite. These solution exchange tests were performed at 90 °C by replacing 80% to 90% of the leachate with fresh demineralized water of Mg-Na-K-Cl brine after set time intervals. The results from these tests provide information about the intrinsic ability of the ceramic waste form and its constituent materials to retain waste components. The results from solution exchange tests with demineralized water indicated that radionuclides are retained in the pure zeolite and in the ceramic waste form to a greater degree than cations like Li, K, and Na during exposure to low ionic strength solutions. A Mg-Na-K-Cl brine was used to investigate the ion-exchange behavior of the ceramic waste form in the presence of high ionic strength solutions. The results indicate that in high ionic strength solutions some ion exchange occurs between components of the brine and fission products in the zeolites. The degree of ion exchange is greater in tests with zeolite 5A than in tests with sodalite. Sodalite still retains Ce preferentially to the other fission products even in a brine solution.
Over 300 planned PCT and MCC-1 tests with reference ceramic waste form samples and tests with its constituent materials have been initiated. The samples used in these tests include reference ceramic waste form samples with non-radioactive fission products, uranium-containing reference ceramic waste form samples with non-radioactive fission products, hot isostatically pressed granulated salt-loaded sodalite samples without the glass 57 binder, hot isostatically pressed glass 57 samples, glass 57 frit as received, and seven hot isostatically pressed ceramic waste form samples produced at Argonne West. These PCT and MCC-1 tests will be performed with durations of 1 day to greater than 1 year; in three leachants, demineralized water, EJ-13 (a simulated groundwater), and brine; and at 40 °C and 90 °C. All of the planned PCT and MCC-1 tests with these materials have been initiated and all the tests with durations less than or equal to 28 d have been terminated. The suite of tests that were initiated was identified in the Work Breakdown Structure for the EBR-II Spent Fuel Demonstration Project. Results from tests with durations of 182 days or less should be available by June 1999. Longer-term tests will be terminated according to schedule and the results reported once the analyses are complete.
To determine the effects that vessels made with different materials may have on the corrosion tests, 14 tests were performed that followed the product consistency tests procedure A (PCT-A) with reference ceramic waste form samples and Teflon, Type 304 stainless steel, and titanium vessels. While the concentration of elements in the leachates were similar for tests performed in different types of vessels, significant differences were observed for some elements that adsorbed on to the vessel walls. These adsorbed species were removed during acid strips of the vessels. The normalized concentrations of the alkaline earth and rare earth elements were significantly higher from acid strips of metal vessels than from acid strips of the Teflon vessels. These results illustrate the importance of performing acid strips to completely account for the release of all elements from the ceramic waste form and suggest that corrosion processes may be affected by the vessel material.
WBS 9.0 Repository Performance Assessment ModelingA significant element in establishing the viability of electrometallurgical treatment technology is a defensible assessment which shows that the waste to be generated from the process will perform acceptably when ultimately deposited in a geologic repository.
The draft National Spent Nuclear Fuel Program (NSNFP) document, Preliminary Design Specification for Department of Energy Standardized Spent Nuclear Fuel Canisters, was reviewed and comments sent to the NSNFP. Particular attention was given to identifying items relevant to placement of the FCF waste forms into the Yucca Mountain repository.
WBS 10.0 Environmental and Safety Support Tasks
These tasks provide the necessary safety analysis support for the electrometallurgical treatment demonstration activities.
Treatment of Oxide Spent Fuels: Oxide spent fuels cannot be electrorefined directly but must first be converted to metals. Lithium metal is used as the reducing agent in molten LiCl, and the Li2O dissolves in the molten salt. As part of the program to explore treatment of degraded spent nuclear fuels, small-scale electrorefining experiments were performed with simulated EBR-II fuel segments. This fuel had been exposed to boiling water for a total of 300 hours, to simulate severely degraded spent fuel. The fuel segments were loaded into anode baskets, placed in the electrorefining apparatus, and a constant potential of 0.4 volts was applied. The cell current that was achieved was much lower than previous experiments, ranging from 0.04 to 0.06 A/cm2. Because this cell current was so low, another similar experiment was done at a higher cell potential of 0.8 V. The cell current was higher in the second experiment. It ranged from an initial value of 0.29 A/cm2, and declined to near zero at the end. A total charge of about 2.2 A-h was passed in about 10 hours. The salt-coated deposit contained about 49% salt and 6.6 g uranium metal. The cladding hulls were found to be completely empty, indicating that oxide or hydride coatings, caused by degradation of the spent fuel, will inhibit the rate of electrorefining; however, degraded fuel can be successfully treated for removal of uranium and preparation for repository disposal.
A series of electrowinning experiments is underway to test the use of tin oxide anodes to replace the expensive platinum. Several technical issues have been identified that pertain to use of tine oxide anodes:
1) the SnO2 anode, like the platinum, will react with lithium metal, thus care must be taken to avoid this interaction, 2) the SnO2 has a higher electrical resistivity than platinum; therefore, the cell voltage must be higher, and 3) because a high cell voltage is needed, extra care is required to assure that decomposition of LiCl and production of chlorine gas is avoided. These conditions were satisfactory met in an experiment in which the concentration of Li 2O in LiCl was reduced from 1.64 wt % to 0.58 wt %, and the current efficiency was 59%. Under the conditions of this experiment, the SnO2 anode has proven to be practical and chemically stable. The SnO2 electrodes are commercially available and much less expensive than platinum.
Treatment of Aluminum-Based Fuels: Demonstration of the feasibility of electrometallurgical treatment of aluminum alloy spent fuels, such as foreign and domestic research reactor fuels, has been done in laboratory-scale experiments. The key step in treatment of this fuel is electrorefining of the aluminum, which represents about 90% of the spent fuel volume, and which can, after electrorefining, be discarded as low-level waste. Preparation of the engineering-scale aluminum electrorefiner is being done by converting a small storage well in the J-118 glovebox to a furnace well for testing the engineering-scale aluminum electrorefiner. All of the components needed for this test are near completion, and assembly in the glovebox is under way.
Electrometallurgical Treatment Program Technical Highlights for September and October 1998
The electrometallurgical program has two components: the EBR-II Spent Fuel Treatment Demonstration project and its adaptation to a variety of DOE spent fuel types. The monthly highlights are divided between the main Work Breakdown Structure (WBS) elements for the project plus two additional tasks: treatment of oxide spent fuels and treatment of aluminum-based fuels. The technical highlights provide an overall picture of the program.
WBS 1.0 Treatment Operations
Electrometallurgical treatment technology will convert the highly enriched uranium and the reactive bond sodium in EBR-II fuel into low enriched uranium product, ceramic waste and metal waste. This work element involves the demonstration equipment operations in the Fuel Conditioning Facility (FCF) and the Hot Fuel Examination Facility (HFEF). The process steps will be operated in an integrated manner to demonstrate the economic and technical feasibility of the process with spent irradiated fuel.
Over the last three months, three operational events have occurred in the FCF which, collectively, have resulted in a significant impact on the demonstration program schedule for fuel and blanket conditioning. The following is a brief description of each event and its subsequent impact on the demonstration program.
On August 19, during the repair of a seal tube for a manipulator, contamination escaped from the temporary radiological containment system which had been set up for the maintenance. The release of contamination into the facility resulted in external contamination of 11 personnel and four of these personnel received uptakes, the largest of which was 23 nanocuries of cesium-137. The release also resulted in contamination of some areas of both the main operating floor and basement areas.
Facility work was restricted until decontamination of the facility was completed on August 24. Further maintenance on manipulator seal tubes and other similar maintenance and operational activities were suspended until the investigation of the event was completed, corrective actions identified and the corrective actions implemented. The suspension of these activities has not resulted in significant impact on the demonstration program, but the decontamination effort did have an impact.
During maintenance on an uninterruptible power system on September 14 it was discovered that the facility condition of “Argon Cell Secure Mode” had not been established as required by the maintenance procedure. The problem was that not all “material at risk” had been stored in the appropriate defense-in-depth containers. Although this was not a violation of the Technical Safety Specifications, the event was considered extremely serious and all in-cell activities were terminated by direction of Argonne management until corrective actions could be identified and implemented.
On September 22, during a transfer of the equipment transfer platen to the argon cell transfer lock, the platen was unseated from its support ram. This event physically shut down transfers into and out of the argon cell and, thus, shut down the processing activities in the argon cell. Additionally, in view of the seriousness of the two previous events, the curtailment of processing activities in the argon cell instituted on September 14 was extended to cover all discretionary operational and maintenance activities in the FCF. This suspension of activities in FCF was rescinded on September 29; however, since the continuation of processing requires continuous availability of the transfer system to move equipment, fuel elements into the argon cell and to transfer other material in and out of the cell as necessary to support processing operations, the demonstration program was physically shut down until the platen and transfer system could be returned to service.
The platen, which weighs in excess of 4000 lb, and the rest of the transfer system are located in a contaminated area in the basement of FCF. The weight of the platen and the location of all components of concern contributed to a lengthy recovery to a full operational status. The platen was returned to the transfer cart on October 6. The control software for this system was upgraded, modified and installed, and the transfer system control and operation were checked out. Electrical and hydraulic system problems identified during the checkout were completed on November 2, and the transfer system was returned to service.
These events resulted in a two month loss of operating time. The demonstration is now projected to be complete in September 1999. The main concern is in the processing of blanket fuel, which is not projected to be complete until late August. The project is currently investigating ways to make up the lost time.
WBS 1.1 Driver Treatment
The demonstration will treat 100 driver assemblies so that the processes can be demonstrated in an integrated system and fission product loading in the Mark-IV (Mk-IV) electrorefiner will reach three weight percent. At the end of October, 76 driver assemblies had been chopped and 72 driver assemblies had been introduced to the electrorefiner.
Driver treatment was suspended because of facility shutdown. One cathode product ingot was produced from material harvested from the Mk-IV anode-cathode module (ACM).
WBS 1.2 Blanket Treatment
The EBR-II blanket assemblies contain 47 kg of uranium each and will demonstrate high throughput rates in the Mark-V (Mk-V) electrorefiner.
Blanket treatment experiments were suspended during facility shutdown.
WBS 1.3 Metal Waste
Three batches (two assemblies each) of irradiated cladding from driver treatment operations will be converted into typical metal waste forms for waste qualification.
WBS 1.4 Ceramic Waste Operation with Irradiated Materials
After 100 driver assemblies are treated in the Mk-IV electrorefiner, a portion of the salt will be transferred to the HFEF where the salt and fission products will be immobilized in ceramic waste samples. This activity was originally scheduled to begin in February 1999; however, the process development testing (WBS 7.0) has encountered several problems. Currently the forecasted start of these activities is March 1999.
WBS 1.5 Facility Operations
Four driver assemblies were received from the Radioactive Scrap and Waste Facility. At the end of the reporting period, 14 driver and 26 blanket assemblies (25 irradiated, 1 unirradiated) were stored in the FCF air cell. Low enriched uranium product (120 kg) was shipped to the ZPPR storage facility. Currently a total of 422 kg of product are in interim storage with an additional 223 kg in FCF.
WBS 2.0 Equipment and Facility Modifications
This work element covers the engineering design, fabrication, assembly and testing activities that are required to implement new process equipment, equipment improvements or facility modifications that support operations or development activities.
In an effort to upgrade the performance of the Mk-V ACMs, two potential design variations of the standard Mod A anode assembly have been identified. Testing of these variations is being performed in the engineering scale glovebox facility at ANL-E. Variation 2, with double basket-to-basket and double basket-to-cathode spacing, was completed and delivered for testing. The standard Mod A baskets are being tested (WBS 3.0).
The Stripper Cathode Assemblies (SCA) for the Mk-V electrorefiner were completed and one was shipped to ANL-W. This unit will be used in place of an anode assembly on the ACM to act as a cathode for a reverse deposition operation to clean the tubular cathode surfaces in the event that the deposition becomes too thick and hard for continued normal operation. The SCA uses 12 EBR-II depleted-U blanket rods, modified in the ANL-W Machine Shop. They are mounted into collets in a horizontal flange in an inverted candelabrum configuration, to act as cathode surfaces for the deposition. Three units were built: one will be used in the Mk-V, one was provided to CMT for their Mk-V ER ACM test facility, and one was retained at ANL-E as a spare and for additional testing, if necessary.
A conceptual design was completed of a Scraper Test Anode Assembly (STAA), with easily removable scrapers, to allow testing of various scraper shapes and materials. The concept uses a Continuous Anode Basket Assembly, of which several have already been built. One basket in each ring will be replaced by a removable bar containing the test scrapers, and all the original scrapers, which were mounted in the basket electrical bus bars, will be removed. The new scraper bars will be removable cold, i.e., without remelting the salt, which is required in all the current designs, including both the original and Mod A Intermittent and the Continuous Anode Basket Assemblies. The unit is intended for testing in the CMT Mk-V ER ACM test facility.
A new bake-out oven for loading two Mk-V ER products into the cathode processor (CP) process crucible was ordered. The oven is intended to replace the existing bake-out oven that will be used to load one Mk-V ER product. The new oven requires a new handling stand for moving the oven in the argon cell. Design of the new handling stand has been started.
The design of a new coil assembly for the cathode processor is complete. The purpose of the new coil is to allow increased power input. The increased power input provides: 1) assurance that the peak temperatures for consolidation can be achieved with margin, and 2) the potential for reducing the processing time of a cathode.
WBS 3.0 Treatment Process Development
The key step in electrometallurgical treatment of spent nuclear fuel is electrorefining to separate pure uranium from the spent fuel, thus reducing the volume of high level waste. Treatment of large quantities of spent fuel, such as the EBR-II blanket, required development of the high-throughput electrorefiner (HTER).
Testing of stainless-steel clad fuel, similar to EBR-II blanket fuel, is now underway. Stainless steel clad fuel resulted in a lower cell resistance of 1.7 mW, compared with zircaloy-clad fuels that gave a cell resistance of 2.5 mW. This low resistance was probably caused by an increase in exposed uranium surface area, and it allowed operation of the ACM at a current of 200 A over the entire batch. However, this high throughput rate could not be sustained due to uranium hold-up, which caused the anode drive to stall. Operation of the ACM appears to be limited to an initial current of 200 A, followed by a decreasing current to 100 A near the end of each hour. At the end of each hour the current is reversed to strip the dense uranium deposit off the cathode. Operation of the ACM will continue in the glove box to optimize operating conditions for use in the FCF.
Electrorefining of the EBR-II blanket fuel inventory will require about 1000 kg of UCl3 to oxidize the bond sodium and the chemically reactive fission products in the fuel. No commercial source is available for this
material, so a method is being developed for large-scale production of UCl3. The production method that was selected for development includes reaction of CdCl3 with uranium metal to form UCl3. These reactions will be conducted simultaneously in a single vessel at 600 °C. About 1500 kg of uranium metal will be needed for both process development and production. A source of depleted uranium has been found, in the form of ~5-in-long by ~1-in-diameter slugs. Most of the LiCl-KCl and LiCl needed for this project are on hand, but an additional ~230 kg will be needed to complete the 100-kg production. Development of the UCl3 production process is now ready to begin.
A prototype cathode processor test was conducted in September on the full-size beryllia crucible, using a 50 kg charge of uranium. In the previous test, the crucible did crack but the loss of uranium was low, meeting goal process requirements. The large charge in the second test placed the melt line well above the existing cracks of the first test. After the test, the cracks on the inside of the crucible bottom appeared to have grown up to the new melt line. The loss is being measured but is estimated to be low. Supporting thermal and structural calculations of the BeO crucible are underway to investigate design solutions. The services of a ceramic consultant have been procured.
WBS 4.0 Process Modeling and Analysis
These activities develop and apply models to improve the understanding of various process steps; to help in design of equipment and selection of process variables; to evaluate the data on performance of the engineering-scale equipment; to provide support in planning of test campaigns; and to conduct operations.
The Nuclear Technology (NT) report on the ASIDE code for radioactive decay has been completed. The ASIDE code remedied shortcomings in the ORIGEN code and its derivatives, including mass conservation for all isotopes and rapid execution time. The ASIDE code has been implemented in the MTG. An evaluation of the necessity for performing radioactive decay of material for every processing step in the FCF has also been completed, with the conclusion that only performing the radioactive decay of the product at the completion of processing before shipping from the FCF would provide adequate accuracy. An evaluation of the effect on substitution of measured data is being done.
The report “MC&A Variance Propagation for the Processing of EBR-II Driver and Blanket Fuel Inventory” to satisfy the WBS DOE 0400580 milestone is undergoing final project review. This report will be issued in November 1998.
The NT report “Incorporation of Measurement Data from Irradiated Fuel Samples Into the Mass Tracking System: (ANL-NT-75) has been issued. This study utilized the large number of measured data from the chopped driver element fuel segment samples to estimate uncertainties in the measured and theoretical values. Biases and trends in the calculated-to-measured (C/M) values of the important data for MC&A are presented for the chopped segments through chopper batch ECBF05A. Based on these results, procedures are recommended for determining if measured and calculated values are consistent.
WBS 5.0 Metal Waste Treatment Development
The noble metal fission products and undissolved cladding hulls are immobilized into a stainless steel-zirconium alloy for geologic repository disposal. In support of waste qualification activities, small samples of the metal waste are being produced so they can be characterized to establish the performance. Also, equipment is being developed and tested in various laboratories to support design efforts on a larger casting furnace for inventory operations.
The fabrication of small scale samples has been completed.
WBS 6.0 Metal Waste Qualification Testing
The metal waste form attributes and fission product release mechanism and rates are being quantified to support repository performance modeling. Samples of the metal waste form are being leach tested by a variety of test methods, including the MCC-1 immersion test. Eighty immersion tests, 40 in deionized water and 40 in simulated Yucca Mountain J-13 well water, have been completed. The test specimens were alloy disks 10-mm diameter by 2-mm thick, polished to a 240-grit finish, sealed in Teflon vessels, and placed in a 90 °C oven for 90 days. The specimens tested in de-ionized water showed a faint gold-colored
coating. The leachate solutions are being analyzed. These results are typical of leaching behavior observed earlier with the metal waste form. Leaching conditions that would result in some attack of borosilicate glass result in very little detectable corrosion of the stainless-steel-zirconium metal waste form. The mechanism of corrosion will have to be studied using much more aggressive methods, including high temperatures and electrochemical corrosion measurement techniques. Conventional leaching methods, such as MCC-1, are being used in an attempt to relate the behavior of the metal waste form to other, less corrosion-resistant waste forms that have been studied more extensively.
Scanning electron microscopy/energy dispersive spectroscopy analysis was completed on alloy specimens from a stainless steel-15Zr-1Nb-1Pd-1Rh-1Ru ingot. This analysis was unique because it was the first examination of the behavior of rhodium, which is a significant fission product in the metal waste stream. The microstructure was essentially identical to that of the typical SS-15Zr alloy without noble metal additions. Phases rich in noble metals were not observed in the microstructure; all noble metal elements were contained within the phases of the SS-15Zr microstructure. The energy dispersive spectroscopy results show that all noble metals (including Rh and Ru) show a strong preference for the Laves Intermetallics (zirconium-bearing phases).
Electrochemical corrosion measurements on metal waste form alloys in pH = 2 solutions have been completed. Alloy corrosion rates were measured by the polarization resistance method (also known as linear polarization). The corrosion cell includes a round-bottom flask, graphite auxiliary electrodes, and a Standard Calomel Electrode, which serves as the reference electrode. The applied potential and resulting current are measured by a Verastat-II Potentialstat/Galvanostat. The equipment is computer-controlled, using SoftCorr III Corrosion Measurement software from EG&G Instruments. The corrosion rates were measured for various alloy specimens in pH = 10 and pH = 2 solutions. For most metal waste alloys, the rates in pH = 2 solutions are at least an order of magnitude larger than the rates in pH = 10 solutions, compared to type 316 stainless steel, which corrodes at about the same rate in both solutions. The corrosion rates are very low in either case, i.e., less than 4 to 5 mm/y.
For testing purposes, samples from two demonstration scale metal waste ingots have been received by the Analytical Laboratory (AL) Hot Cells and HTEF. The Toxic Characteristic Leaching Procedure (TCLP) test will be performed in the AL hot cells and initiated during the second week of November. The other samples sent to the AL will be analyzed for bulk elemental make-up and impurities (Si, C, O, N). The samples sent to HFEF will be examined using scanning electron microscopy and subjected to a leach test.
WBS 7.0 Ceramic Waste Treatment Development
The electrolyte salt is periodically removed from the electrorefiner and passed through a waste treatment system to immobilize fission products and transuranium elements for disposal. A large-scale zeolite column is being designed for testing with actual spent fuel after the Demonstration project is complete. The column size for this demonstration is 5-in diameter by 28-in high. Completion of the conceptual stage design during October represented completion of a Program milestone, “Establish design parameters for large-scale zeolite column,” scheduled for September 1998.
The ceramic waste form that results from electrometallurgical treatment of spent nuclear fuel incorporates the chloride salt, most of the fission products, and the transuranic elements into a glass-bonded sodalite composite ceramic. A series of ceramic waste form samples is being fabricated to test the effect of zeolite and glass frit particle size distributions on the waste form integrity and durability. Among the three different particle size distributions used in the present study, no differences in density or porosity were found. Differences in the amount of free chloride were observed, however. The smaller particle sizes and the narrow particle-size ranges resulted in the largest amount of free chloride. Large particle sizes with narrow particle size ranges also gave high free chloride amounts. As expected, the best performance, in terms of low free chloride, was achieved with powders having a wide range of particle sizes. Samples from these tests are being used to determine leaching behavior as a function of the zeolite and glass frit particle sizes, with the objective of determining the optimum particle size distributions for fabrication of the final waste forms.
Studies are being done with plutonium-loaded ceramic waste form samples to determine the effects of transuranics on waste form characteristics and leaching behavior. The plutonium-bearing samples were prepared by hot uniaxial pressing (HUP), rather than hot isostatic pressing (HIP), because of the risk of contamination of the HIP equipment. To prepare the samples, zeolite was loaded with simulated waste salt that contained plutonium and non-radioactive fission products at levels representative of the electrorefiner salt composition after treating 300 driver fuel assemblies. The loaded zeolite was then mixed with glass frit and consolidated by the HUP process. Completion of these samples completes the Project milestone for production of plutonium-loaded waste form samples.
During September and October, mock-up testing of the HIP was completed, and the equipment was installed in HFEF. Salt-zeolite segregation problems were encountered in the heated V-mixer. The cause was determined to be different particle sizes. Modifications were made to the mill/classifier to produce salt particles of similar size as the zeolite. These changes resolved the segregation problem and resulted in a salt-loaded zeolite product that met the process specification for free chloride. The final glovebox V-mixer tests with zeolite were completed in late October. For accelerated alpha decay studies, the first two batches of salt-loaded zeolite containing plutonium-238 were produced. Plutonium-238 ceramic waste samples were then produced using the HUP.
WBS 8.0 Ceramic Waste Qualification Testing
The ceramic waste form is being characterized so that its performance in different repository conditions and scenarios can be assessed. This work characterizes hot uniaxial pressing samples and laboratory scale and demonstration scale samples from hot isostatic pressing.
Seven demonstration scale samples produced at ANL-W are being used in a blind characterization study. The purpose of this study is to evaluate how sensitive various tests are to changes in the process variables used in fabricating seven samples. The characterization tests included visual inspection, X-ray diffraction (XRD) examination, accessible free salt measurement, density and apparent porosity measurements, and corrosion testing (7-day PCT and 3-day MCC-1 tests) in demineralized water at 90 °C. The mass and chloride losses observed in the corrosion tests were presented in the June monthly report. The cation and iodide losses have now been calculated and the results for cesium and the alkali metals and iodide are consistent with the trends shown in the mass and chloride losses. The density and apparent porosity measurements show comparatively little difference in the seven samples. To address some concerns regarding the accessible free salt data, we re-measured accessible free salt using a modified procedure in which we used 95% rather than 100% ethanol and multiple rinses instead of only one. We found more accessible salt is removed with 95% ethanol and with multiple rinses.
Corrosion testing is also being done with the components, glass and sodalite, of the ceramic waste form. Separate samples of glass frit and salt loaded zeolite were hot isostatically pressed to formed HIPed glass and sodalite products. Corrosion tests of these materials should indicate whether the behavior of the two major phases of the glass frit and the salt loaded zeolite act as a composite. The former approach is easier in some respects because more information is available on the corrosion behavior of glass and aluminosilicate minerals. Results from 7-day PCT and 3-day MCC-1 tests for glass, sodalite and the reference ceramic waste form indicate that the releases of the larger cations (Cs and Rb) and divalent cations (Ba and Sr) are higher from pure sodalite than from the reference ceramic waste form. Other elements, however, have similar behavior. Longer term corrosion tests are ongoing to determine if the higher releases of Cs, Rb, Ba, and Sr are the result of an initial spike of accessible salt. (It is possible that less salt is occluded when glass is not present.) In addition, we are planning to examine HIPed sodalite and the reference ceramic waste form with scanning electron microscopy to determine if porosities and exposed surface areas can account for the differences.
A testing program has been established to determine the durability and the corrosion properties of the reference ceramic waste form (CWF), a glass-bonded sodalite. The reference CWF differs from earlier compositions in that its glass content was reduced from 50 to 25 wt % and a larger particle size was substituted for pure zeolite powders to improve handling properties. Results have been obtained from MCC-1 and PCT leach tests of the reference CWF, with durations from 1 to 91 days. Analyses have been
done for the matrix elements, Al, Si, and B, the alkali metals and chloride. The order of elemental release was the following:
Cl > Na, Cs > Li > Al, Si > K, Rb > Sr, Ba > B.
Boron had the lowest release and its release remained very low over the entire time period, indicating that the glass has a very low corrosion rate in MCC-1 tests with low S/V. The releases of Al and Si were high initially but quickly stabilized. The releases of the alkali metals and chloride also stabilized with time. We interpret these data to mean that the initial release of salt is due to the dissolution of accessible or free salt and that further release is slow and due to dissolution of occluded salt resulting from sodalite corrosion.
PCT tests are being used to characterize saturation conditions and provide information on corrosion behavior when alteration phases are forming. The PCT results showed that the order of elemental releases for B, Si, and Al were different from the MCC-1 tests. The order of elemental release in the PCT was the following:
Cl, Cs > I >Li, Na > Rb, K > B > Sr, Ba > Si, Al.
The very low releases of Si and Al were taken as evidence of the formation of alteration phases and solubility limited reactions. An interesting aspect of the PCT was that Cs release decreased with time, indicating that Cs may be incorporated into alteration phases. Comparison of MCC-1 and PCT results for the ceramic waste form (CWF) with similar data from high-level borosilicate glasses showed that the CWF is as durable or more durable than reference high-level waste glasses under these specific test conditions.
WBS 9.0 Repository Performance Assessment Modeling
A significant element in establishing the viability of electrometallurgical treatment technology is a defensible assessment which shows that the waste to be generated from the process will perform acceptably when ultimately deposited in a geologic repository.
The document, “Status Report on Metal Waste Form Release Rate Modeling,” was issued, satisfying WBS DOE 901220 milestone.
The report “Radionuclide Release Modeling of the ANL Ceramic Waste Form” to satisfy DOE 901200 milestone is undergoing final project review. This report will be issued in November 1998.
The report “FCF Waste Form Production” to satisfy the WBS DOE 0900840 milestone is undergoing final project review. This review will be submitted to the National Spent Nuclear Fuel Program (NSNFP) for review and comment. Upon completion of the review and comment task, the report will be issued.
Scoping studies are underway to evaluate the impact on the predicted performance of the Yucca Mountain respository of disposal of the FCF ceramic and metal waste forms in the repository. The tools used in this evaluation are 1) the Yucca Mountain Project's TSPA-VA base case model and 2) a simplified version of the TSPA-VA base case model which has been developed by Golder Associates for use in repository scoping and sensitivity studies.
WBS 10.0 Environmental and Safety Support Tasks
These tasks provide the necessary safety analysis support for the electrometallurical treatment demonstration activities.
Treatment of Oxide Spent Fuels: Oxide spent fuels cannot be electrorefined directly but must first be converted to metals. Lithium metal is used as the reducing agent in molten LiCl, and the Li2O dissolves in the molten salt. The engineering-scale reduction experiment, ES-9, was completed this month. This experiment was designed to test modifications to the Mark-V fuel basket design in terms of the rate reduction of the oxide. The modifications were designed to increase the rate of access of the molten salt to the bulk fuel. About 2.3 kg of crushed UO2 in various particle sizes was loaded into a single basket. Lithium was absorbed in a porous stainless steel disk and loaded into the molten salt bath, and the salt from
the previous electrowinning experiment, ES-8, was reused. Initial analytical results indicate that a significant improvement in reduction rate was achieved with the new basket design. An unusual amount of data scatter was observed with the salt titrations to determine Li 2O content, and further study is needed to understand the reason for this scatter. X-ray diffraction analysis of the reduced uranium product will be done to confirm the titration results. Successful completion of this experiment will assist in design of fuel baskets for a full-scale oxide fuel treatment process.
The interface between the reduction and electrorefining processes is important, because residual lithium and Li2O will react with UCl3 in the electrorefiner. A series of laboratory-scale experiments has been designed to examine this potential interface problem. The current laboratory-scale electrorefiner is sized to handle batch sizes of 20 to 50 g uranium. Efforts are now focused on expanding the capacity to batch sizes of 200 to 600 g uranium. The design of the larger electrorefiner was completed, and fabrication of the cell components is in progress.
Treatment of Aluminum-Based Fuels: Demonstration of the feasibility of electrometallurgical treatment of aluminum alloy spent fuels, such as foreign and domestic research reactor fuels, has been done in laboratory-scale experiments. The key step in treatment of this fuel is electrorefining of the aluminum, which represents about 90% of the spent fuel volume, and which can, after electrorefining, be discarded as low-level waste. Preparation of the engineering-scale aluminum electrorefiner is being done by converting a small storage well in the J-118 glovebox to a furnace well for testing the engineering-scale aluminum electrorefiner. All of the components needed for this test are complete, and testing in the glovebox is ready to begin.
