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STAINLESS STEEL-ZIRCONIUM ALLOY WASTE FORMS FOR METALLIC FISSION PRODUCTS AND ACTINIDES ISOLATED DURING TREATMENT OF SPENT NUCLEAR FUEL*

Sean M. McDeavitt Daniel P. Abraham Dennis D. Keiser, Jr. Jag-Yul Park Argonne National Laboratory Argonne National Laboratory Argonne National Laboratory Argonne National Laboratory 9700 Cass Avenue 9700 Cass Avenue P.O. 2528 9700 Cass Avenue Argonne,South Illinois 60439 Argonne,South Illinois 60439 IdahoBox Falls, Idaho 83403 Argonne,South Illinois 60439 (708) 252-4308 (708) 252-5486 (208) 533-7298 (708) 252-5030

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To Be Presented At: SPECTRUM ‘96 Seattle, Washington, USA August 18-23, 1996

*Work supported by the U.S. Department of Energy, Nuclear Energy Research and Development Program, under Contract No. W-3 1-109-Eng. 38.

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report was prepared as an account of work sponsored by an agency of the ThisUnited States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or use- fulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufac- turer, or otherwise does not necessarily constitute or imply its endorsement, wm- mendhtion, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. STAINLESS STEEL-ZIRCONIUMALLOY WASTE FORMS FOR METALLIC FISSION PRODUCTS AND

ACTINIDES ISOLATED DURTNG TREATMENT OF SPENT NUCLEAR FUEL

Sean M. McDeavitt Daniel P. Abraham Dennis D. Keiser, Jr. Jang-Yul Park Argonne National Laboratory Argonne National Laboratory Argonne National Laboratory Argonue National Laboratory 9700 South Cass Avenue 9700 South Cass Avenue P.O. Box 2528 9700 South Cass Avenue Argonne, Illinois 60439 Argonne, Illinois 60439 Idaho Falls, Idaho 83403 Argonne, Illinois 60439 (708) 252-4308 (708) 252-5486 (208) 533-7298 (708) 252-5030

ABSTRACT The metal waste form comprises the spent fie1 cladding, the noble metal fission products (NMFP) Stainless steel-zirconium waste form alloys are being (e.g., Ru, Rh, Pd, Zr, and Tc) that do not dissolve in the developed for the disposal of metallic wastes recovered electrolyte salt, and, in some cases, zirconium metal fiom from spent nuclear fuel using an electrometallurgical alloy nuclear fuels. These metallic wastes are not process developed by Argonne National Laboratory. The generated in the electrorefineq they are present with the metal waste form comprises the fuel cladding, noble metal spent fuel before treatment. The spent fuel cladding hulls fssion products, and other metallic constituents. Two and NMFPs are inert in the electrorefiner environment nominal waste form compositions are being developed and, therefore, remain in the charge basket as the uranium, (1) stainless steel-15 wt% zirconium for stainless other actinides, and active fission products are steel-clad fuels and (2) zirconium-8 wt% stainless steel for electrochemically dissolved or transported. Zircaloy-clad fuels. The noble metal fission products are the primary source of radiation and their contribution to the waste form radioactivity has been calculated. The The term “waste form” refers to radioactive materials disposition of actinide metals in the waste alloys is also and any encapsulating or stabilizing matrix that will be being explored. Simulated waste form alloys were placed into a “waste package” for disposal. The prepared to study the baseline alloy microstructures and radioactive materials may be stabilized for disposal by the microstructural distribution of noble metals and encapsulation, chemical transformation, and/or inclusion actinides, and to evaluate corrosion performance. in a stable matrix material. Vitrification in borosilicate glass has been widely selected as a primary stabilization method for high level wastes.’ However, at least two I. INTRODUCTION issues make vitrification an undesirable option for the metallic waste stream considered here. First, NMFPs are Alloy waste forms are being developed at Argonne strong crystal formers in vitrified waste forms and crystal National Laboratory for the immobilization of metallic formation decreases the mechanical integrity of glass? materials left behind followin& the electrometallurgical Second, vitrification of the metallic wastes would result in treatment of spent nuclear fuel. This treatment process very significantmass and volume increases. Minimizing is being implemented as a full-scale demonstration in the the waste form volume is critical because the unit volume Fuel Conditioning Facility (FCF)’ in conjunction with cost for repository disposal is expected to be quite high. the shutdown and dismantling of the Experimental Breeder Reactor-II (EBR-II). The principal step in this process is the electrorefining of uranium metal in a molten Two types of fuel were used in EBR-11: driver ibel salt e~ectro~yte;-’Three distinct material streams emanate and blanket fuel.” The driver fuel is a U-10 wt% Zr from the electrorefiner: (1) refined metallic uranium; alloya with Type 316 and D9 stainless steel cladding, and (2) fission products and actinides extracted hm the the blanket fuel is uranium metal with Type 304 stainless electrolyte salt that are processed into a mineral waste steel cladding. The cladding materials plus the NMFP form; and (3) metallic wastes that are consolidated into and Zr fiom the driver fuel will be melted together into a the metal waste form. uniform, corrosion-resistant alloy waste form. Approximately 0.5 to 4 wt% NMFP will be present in

a All compositions are in wt% unless indicated. the metal waste form, depending on the fuel burnup. The of -133OOC. The Zr-8SS composition is not near a alloying process will be carried out in a high-temperahm, eutectic, but its liquidus temperature is -1500OC. controlled-atmosphere melting furnace. A molten salt flux Therefore, both alloys may be produced at temperatures may be used as a solvent to dissolve residual electrolyte near 1600°C. A large-scale tilt-pour casting furnace (3 kg salt and maintain metal purity. capacity) was designed, built, and connected to an inert atmosphere glovebox for use in developing the waste form alloying procedures. This furnace is being used to Although EBR-I1 fuel will be treated first, other spent generate large ingots and to test processing parameters. fuels with stainless steel and Zircaloy claddings are being evaluated for fuhm treatment. In all cases, the cladding hulls represent over 85% of the metal waste stream. By Small-scale (-20 g) stainless steel-zirconium alloys using the hulls as the major alloying component, the total were generated that bad a range of zirconium compositions waste form volume is minimized. This approach gave to simulate the SS-15Zr and Zr-8SS waste forms and rise to the parallel development of two metal waste form other SS-Zr alloy compositions.2 The alloys were compositions: (1) stainless steel-15 wt % zirconium prepared from stainless steel (Types HT9, 304, or 316), (SS-15Zr) for stainless steel-clad fuel and zirconium, and selected noble metals &e., Ru, Pd, Mo, (2) Zircaloy-8 wt% stainless steel (Zr-SSS) for and Ag) that were melted at 1600°C in yttria (Y2G) Zircaloyclad fuel. The metal waste form development crucibles under an argon atmosphere for 1-2 h and effort includes the ongoing evaluation of the physical solidified by cooling slowly. The microshuctuxal metallurgy, corrosion performance, thermophysical evolution and phase development of SS-Zr alloys with properties, and process variables important to waste form various Zr contents was studied by using a scanning generation and qualificftion. Metal waste forms electron microscope (SEM), energy dispersive X-my containing actinide metals are also of interest as a backup analysis (EDX), and X-ray dBhction2 More recently, option to actinide-bearing mineral waste form. neutron diffiaction has been used to characterize phases the that have only a minor presence in the waste form alloys.

II. METAL WASTE FORM ALLOYS Stainless steel-rich alloys containing 5 to 40 wt% Zr The nominal alloy compositions SS-15Zr and exhibit varying proportions of an iron solid solution Zr-8SS were selected on the basis of initial phase and a Laves-- intermetallic (AB2 crystal characterization and corrosion data. In addition, such structure) that we have designated Zr(Fe,Cr,Ni)2w.2 The parameters as alloying temperature and minimal nonwaste microstructure of the nominal waste form for stainless additions were also considered. The SS-15Zr steelclad fuel, SS-15Zr, is presented in Figure 1. As a composition is near a eutectic having a melt temperature first approximation, the SS-15Zr alloy contains a

Ii p' P

Figure 1. Backscattered Electron Image (200s) of the SS-1SZr Eutectic St~~cture.The dark phase sis an iron solid solution, and the bright phase is the Zr(Fe,Cr,Ni)z, intermetallic. Table 1. Composition of Observed Phases in SS-15Zr and Zr-8SS Alloys.’

two-phase structure, where the dark phase is a ferritic Fe relative proportion of Zr(Fe,Cr,Ni)2+x increases with (a-Fe) solid solution and the bright contrast phase is the increasing zirconium concentration until -40 wt% Zr, intermetallic compound Zr(Fe,Cr,Ni)2+,. The when the alloy is -100% intermetallic and brittle. In compositions of the various SS-15Zr alloy phases are addition, a minor quantity of a second intermetallic phase given in Table 1. was identified by neutron dif€raction and then observed using the SEM. phase was identified as ZrsFeu, a stable Fe-Zr intermetallicThis phase.” Below -15 wt% Zr, the Fe solid solution phase is found to be a mixture of a-Fe and austenitic Fe (y-Fe), both of which contain Cr and Ni levels corresponding to Zirconium-rich alloys (>40 wt% Zr) contain those of femtic and austenitic stainless steels. A minor multi-phase mixtures of various brittle intermetallic volume fraction of the y-Fe phase is observed in SS-15Zr phases up to -84 wt% Zr (16wt% SS).* As the Zr alloys genemted with Type 3 16 stainless steel, whereas content increases from 84 to 100 wt%, a zirconium solid only the a-Fe phase is present when Type 304 stainless solution phase (a-Zr) is observed in increasing quantity, steel is used. This minor Wereme in structure is the along with decreasing quantities of the intermetallic result of a higher nickel content in Type 316 stainless phases. Figure 2 shows the microstructure of Zr-8SS, the steel vs. Type 304 stainless steel. The Zr(Fe,Cr,Ni)2+x nominal waste form alloy for Zmaloyclad fuel. The intermetallic is a strong sink for Ni, an austenite Zr-8SS alloy possesses a multi-phase microstiucture stabilizer, but it saturates at low Zr concentrations, dominated by the primary a-Zr solid solution surrounded leaving excess Ni to stabilize the y-Fe phase. The by a complex eutectic structure containing the a-Zr phase

253 2

Figure 2. Backscattered Electron Image (200x) of the Zr-8SS Structure. The major phase is a zirconium solid solution, and the darker phases are intermetallics. Table 2. Chemical Inventory and Radioactive Source Strength for EBR-11 Metal Waste Stream

Calculated 1996 Levels Calculated Levels After 1000 y Decay fig) ictivity (Ci) Ci ms wt% % ms(kP) wto/o Activity (Ci) Ci % 64 1.5 1.0~10~ 4.1 91 2.0 5.1 6.6 Added 615 13.5 . . 615 13.5 e . Fe 2300 50.8 1.6~10~ 6.5 2300 50.4 . Cr 690 15.3 . . 690 15.2 e . Cladding Ni’ 730 16.1 e . 730 16.0 6.7 8.8 Components Mo. 25 0.6 . . 25 0.6 4.3 5.6 Mn* 37 0.8 1.0~10~ 4.2 37 0.8 b . Si 19 0.4 0 0 19 0.4 0 CO* 9 0.2 2.3~10~ 0.9 9 0.2 e . Ag 0.3 0.01 . e 0.3 0.01 . . Nb 6.1 0.14 2.3~10~ 9.3 6.1 0.13 12.9 16.9 Pd 3.3 0.07 . . 3.3 0.07 0.3 0.4 NMFP Rh 2.1 0.05 8.8~lo4 35.7 2.1 0.05 . . Elements Ru 6.6 0.15 8.8~10~ 35.7 6.6 0.14 . e Sb 0.06 0.00 8.4~10~ 3.4 0.05 0.00 3.0 4.0 Sn 0.2 0.01 4.5~10’ 0.2 0.2 0.01 2.7 3.5 Ta 0.9 0.02 . e 0.9 0.02 . . Tc 2.5 0.05 . e 2.4 0.05 41.4 54.2 NMFP Totals’ 22 0.49 2.2x10’ 88.4 22 0.49 65.4 85.6 Column Totals 4511 99.7 2.5~10’ 100 4538 99.6 76.4 100 ivation product,- andor cay product isotopes. 7 NMFP activity totals include numbers from fuel Zr. Activity either zero or orders of magnitude below dominant radioactive species. and intermetallic compounds; these intermetallics have (in Curies) of the EBR-11 metal waste stream. The values been qualitatively identified as ZrZ(Fe,Ni) and Zr(Fe,Cr)z. listed are combined totals for Mark-III driver fuel and The compositions of the Zr-8SS alloy phases are blanket fuel. The NMFP content in the driver fuel is presented in Table 1. expected to be as high as 2 to 4 wt%, whereas the NMFP content in the blanket is expected to be below 0.5 wt%. The zirconium data was separated to distinguish between III. NOBLE METAL FISSION PRODUCTS Zr from the spent fuel and non-reactor Zr that will be used to genelate the SS-15Zr alloy. Several minor cladding The fission product composition in spent nuclear fuel components and impurities (e.g., C, Ti, and Cu) and is dependent upon the fuel’s accumulated bumup. That NMFPs (e.g., Cd, In, V, and Zn) are not included in is, as nuclear fission occurs, the fuel isotopes Table 2. (The missing cladding materials account for the (e.g., U-235) split into a wide variety of lighter isotopes. 0.3 wt% deficit in the second column, whereas the The entire fission product inventory includes gases and missing NMFPs repment <0.01 wt% of the entire chemically active isotopes that are not included in the inventory.) metal waste stream. NMFPs represent -35% of the total fission product inventory. The actual quantity and composition of NMFl?s present in a given metal waste Each element listed in Table 2 is present as several stream are dependent on starting fuel composition, different isotopes, and the intense initial radioactivity neutron energy spectrum, and the duration of inadiation comes from NMFP isotopes present in very minor concentrations. The metal waste form composition is, therefore, not significantly altered as the high-activity, A nuclear fuel modeling code, ORIGEN, was used to short-lifetime isotopes decay away. For example, calculate the fission product inventory of EBR-11 fuelb in ORlGEN predicts -6.6 kg of Ru in EBR-11 fuel, but the its current state and at future times by simulating changes -88,000 Ci initial activity comes exclusively from an due to radioactive decay. The data listed in Table 2 estimated 0.3 g of Ru-106 and Ru-103, which have represent the chemical inventory (in kg) and ladioactivity half-lives of 372.6 d and 39.2 d, respectively.

The intense initial radioactivity decays rapidly b Calculation by R. N. Hill, Reactor Analysis Division, as Argonne National Laboratory. transmutation occurs. Figure 3 shows the time-dependent 10-1 0 Comnarison Data Typical LWR Spent Fuel -1 ci/g - - 1W2 1996 Metal Waste Form 5.4~10~~Ci/g 2996 Metal Waste Form 1.7~10”Ci/g Natural Uranium Oxide 3.1~10-~Ci/g 10” - -

lo4 - -

0 1o-~ 1.1,1.1.1.1.1.1.1.1.1

Decay Time (years afier 1996)

Figure 3. Calculated Specific Activity of EBR-11 Metal Waste Form vs. Time. change in specific activity for an “average” EBR-11 metal solution phase. The noble metal distribution has not yet waste form (i.e., Table 2 composition). The radiation been quantified for the Zr-8SS alloy, but a similar absence level drops precipitously over the fust 100 years and of noble metal precipitation was seen. These observations settles into a long-term rate of decay dominated by indicate that the stainless steel-zirconium waste form long-lived NMFP isotopes, primarily Tc-99. The metal alloys are indeed viable as NMFP disposition alloys. waste form activity is relatively benign when compared to spent light water reactor (LWR) fuel and other radioactive waste forms that may be permanently distosed of in a IV. ACTINIDE-BEARING WASTE FORMS geologic repository. ORIGEN calculations indicate that the long-term specific activity of typical LWR spent fuel Small-scale samples of simulated waste form alloys is on the order of 1 Ci/g, even after lo6 years. The EBR- (-30 g) containing U, Pu, and Np were generated by 11 metal waste form activity will be only -0.05 Ci/g at its melting in yttria crucibles at 160OOC for 2 h under a peak, and it decays rapidly to -3x10 Ci/g in the first flowing argon atmosphere and cooling slowly to room 100 y. This value is only two orders of magnitude higher temperature. The samples included SS-15Zr alloys with than the specific activity of natural uranium oxide. actinide compositions of 0.5U-OSPu, 2U-2Pu, 6Pu, lOPu, 6Pu-2Np, and 2Np and Zr-8SS alloys with 4, 7, and 10 wto/o Pu. The SS-15Zr alloys were generated The noble metals Ru, Re, Pd, Mo, and Ag were using Type 316 stainless steel, and the Zr-8 stainless added experimentally to SS-15Zr and Zr-8SS alloy steel alloys were generated using Type 304 stainless steel. samples to simulate the presence of NMFPs, and the total The high actinide concentrations used in this study were nobl metal content ranged from 1 to 5 wt% (a typical selected to provide insight into the actinide interactions addition to SS-15Zr was 2 Ru-1.5 Pd-0.5 Ag). Alloy with the existing phases. If actinides are placed into the microstructures were examined, and discrete noble metal metal waste form in actual practice, the concentration will phases were not observed. No differences were seen be between 1 and 10 wt%. between noble metal-containing microstructures and the baseline microstructures(e.g., Figures 1 and 2). For SS- 15Zr alloys, the noble metals were dissolved and Figure 4 presents a representative microstructure far distributed between the intermetallic and the iron solid the actinide-bearing SS-15Zr alloys that is similar to the solution phase. Some elements (Le., Ru, Pd, and Ag) baseline SS-15Zr microstructure in Figure 1. There is showed a -2: 1 preference for the intermetallic phase, while 2 atom % Pu in the Zr(Fe,Cr,Ni)Z+, intermetallic, but no others (Le., Re) showed a similar preference for the iron actinides are detectable in the iron solid solution phase. 10

Figure 4. Backscattered Electron Image of SS-15Zr Alloy Containing 2 wt% U and 2 wt% Pu. The bright contrast phases are rich in U and Pu.

The high-contrast phase is an intermetallic that is The Zr content was as low as 13 atom % in the sample apparently miscible with the Zr(Fe,Cr,Ni)2, phase; its with 2 atom % Np. For actinide-bearing alloys without composition is 33 atom % Fe-33 atom '% Ni-20 atom % Np, the Zr content in the Zr(Fe,CrYNi)2+,phase was -22 actinide (U, Pu, and/or Np) plus small amounts of Zr and atom %y similar to the 24atom % reported for the Cr. This phase was observed in all of the actinide-bearing baseline alloy in Table 1. The Fe solid solution phase SS-15Zr alloys, irrespective of the actinide or group cf did not exhiiit aqconcentration differences. actinides included in the sample; increasing the actinide content resulted in a higher volume fraction of this phase. Additionally, changes in the zirconium content of the Figure 5 presents a representative microstructure for Zr(Fe,Cr,Ni)Z+, phase were observed in Np-bearing alloys. the Zr-8SS-xPu alloys. Again, the microstructure

Figure 5. Backscattered Electron Image of Zr-8SS Alloy Containing 10 wt!?? Pu (300x). The bright-contrast features are a-Zr phase boundaries with -12 atom % Pu resembles the corresponding baseline alloy (Figure 2). The Pu was observed in solution in the a-Zr metal General immersion corrosion tests were carried out matrix, but it was also found in higher concentrations at using a test procedure based on the MCC-1 Static Leach the a-phase boundaries (i.e., the bright spike-like fm Test developed for glass-based waste forms. Disk in Figure 5). The formation of these Pu-rich features may specimens, 15.9 mm diameter and 3 mm thick, were be explained as follows: (1) the high-temperature polished to a 600 grit finish, immersed in the J-13 zircoNum metal phase, p-Zr, is completely miscible with solution in a sealed Teflon vessel, and placed in an oven the high-temperature plutonium metal phase, E-Pu, but at 90°C. Test specimens made from SS-15Zr and Zr-8SS the low-temperature zirconium metal phase, a-Zr, has alloys and other off-nominal Zr compositions were tested. limited solubility for plutonium (43 atom %); (2) the However, the test solution was benign to the stainless p-Zr metal forms first upon cooling, with up to 100% cf steel-zirconium alloys, and corrosion rates were not the plutonium in solution; (3) p-Zr transforms to a-Zr at measurable. In fact, most of the surfaces of the immersion -863OC; and (4) the excess Pu that exceeds a-Zr specimens remained shiny after exposure durations up to solubility becomes concenmted at newly formed a-Zr 10,000 h (381 d). This is a positive result, but it does phase boundaries. Increasing the amount of Pu added to not provide a quantitative means to evaluate the waste the alloy from 4 to 10 wto/o Pu resulted in higher amounts form pesormance. of Pu both in the a-Zr matrix phase (from 1.5 to 5 atom% Pu) and at the a-phase boundaries (up to 12 atom % Another test method employed is the electrochemical Pu). linear polarization measurement of corrosion rates. This method can measure very low corrosion rates in a short Neither actinide-bearing alloy exhibits pure actinide duration test. The electrochemical cell current is phases. Since the actinides are entrained in complex, but measured and mathematically converted into a uniform stable, matriv phases, the stainless steel-zirconium alloy corrosion rate; localized corrosion may afkt this waste forms have potential application as actinide disposal measurement, but we have not observed any evidence d alloys, Performance testing must be carried out in the this phenomena. Measurements were made at pH = 2, 4, 7, and 10 to cover a mnge of potential repository future to verify this viability, but since Pu has only a minor effect on the alloy microstructure, the performance conditions; pH = 2 represents an extreme acidic condition of the baseline alloys (discussed in Section V) may be that may not occur in the repositoly environment, but it used to infer the performance of the actinidehxuing provides an aggressive test to compare the relative alloys. performance of these lowcorrosion-rate metals.

V. WASTE FORM EVALUATION The electrochemical corrosion rates are shown in Table 3 for the waste form alloys, commercial zirconium Small-scale samples with wrious zirconium contents and stainless steel, and selected candidate canister were machined and polished into disks and tested using materials. The waste form corrosion rates at pH = 7 are general immersion and electrochemical corrosion comparable or slightly lower than the measured rates for methods. Large-scale ingots made in the 3 kgcapacity Types 316 and 304 stainless steel and zirconium metal. tilt-pour hcewere used to generate specimens for Noble metal additions do not significantly &ct the testing corrosion, mechanical, and thermophysical corrosion rates of the waste form alloys. Also, the properties. Preliminary data indicate that the SS-15Zr measured waste form corrosion rates are similar to the rate alloy is a very strong metal with minimal tensile for Incoloy 825, lower than the rate for pure copper, and ductility. Uniaxial tension tests on specimens ~vealed two orders of magnitude lower than the rate for mild steel. elongation of <1% and a fracture strength on the order CE 40 hi. VI. SUMMARY Corrosion resistance is the primary pelformance Stainless steel-zirconium alloy waste forms are being indicator for the metal waste form. Corrosion testing was developed for the remnant metallic wastes from the typically carried out using simulated J-13 well water, electrometallurgical treatment of spent nuclear fuel. The which is representative of the groundwater at the Yucca alloys SS-15Zr and Zr-8SS are the baseline waste form Mountain site in Nevada that has been proposed for a alloys for stainless steelclad and Zircaloy-clad fuels, high-level nuclear waste repository. The ionic respectively. The microstructures of these alloys were concentration of 5-13 well water is (in ma): 11.5 Ca, characterized to provide data for the evaluation of NMFP and actinide inclusion. The NMFPs from the spent 1.76 Mg, 43.0 Na, 5.3 K, 0.06 Li, 0.04 Fe 0.001 Mn, fuel 0.03 AI, 30.0 Si, 2.1 F, 6.4 Cr, 18.1 SO?: 10.1 NO;, are present in small quantities and are distriiuted in solid HCO;, and 5.7 dissolved ovgen. solution. Corrosion tests have shown that the metal 143.0 waste form alloys are highly corrosion resistant, both with Corrosion Rates WY*> Alloy (in wt%) pH=2 pH=4 pH=7 pH=lO SS-15Zr 0.1-0.4 0.08-0.2 0.02-0.08 0.01-0.02 SS-l5Zr-2Ru-1 SPd-0.5Ag 0.4-0.5 0.2 0.04-0.1 0.06-0.09 Zr-8SS 0.02-0.08 0.05-0.06 0.02-0.03 0.01 Zr-8SS-lRu-lMo-O.SPd 0.08 0.04 0.03 0.01 zirconium 0.02 0.1 0.06-0.09 0.07 Type 304 SS 0.1 0.05 0.05 0.03 Type 316SS 0.3 0.05 0.04 0.03 Incoloy 825 0.03 0.05 0.07 0.03 CU-~AI(CDA614) 2.3 6.9 3.6 2.0 A106 Grade B low alloy steel 50 23 12 12 and without the presence of NMFps. Actinides slightly 5. J.E. Battles, J. J. Laidler, C. C. McPheeters, and modify the phase structure of the alloys, but data is not W. E. Miller, “Pyrometallurgical Process for Recovery d yet available regarding the effects on corrosion Actinide Elements,” Actinide Processing: Methods and performance. Materials, eds. B. Mishra and W.A. Averill, p. 135-151, The Minerals, Metals, & Materials Society, Warrendale, Pennsylvania (1994). ACKNOWLEDGMENTS 6. E. C. Gay and W.E. Miller, “Electrorefining The authors would like to acknowledge J. K. Basco NReactor Fuel,” Proc. DOE Spent Nuclear Fuel and P. A. Hansen for their supporting efforts in this work. Challenges and Initiatives, Salt Lake City, Utah, Dec. This work was supported by U. S. Department of Energy 13-16, 1994, p. 267, American Nuclear Society, under contract W-3 1-109-Eng-38. LaGrange Park, Illinois (1994). 7. J. P. Ackerman, “Chemical Basis for Pyrochemical REFERENCES Reprocessing of Nuclear Fuel,” Ind. Eng. Chem. Res., 30, 29 (1991). 1. S.M. McDeavitt, J.Y. Park, and J.P. Ackerman, “Defining a Metal-Based Waste Form for IFR 8. A. J. G. Ellison, J. J. Mazer, and W. L. Ebert, Effect Pyroprocessing Wastes,” Actinide Processing: Methods of Glass Composition on Waste Form Durability: A and Materials, eds. B. Mishra and W.A. Averill, Critical Review, Argonne National Laboratory Report p. 305-319, The Minerals, Metals, & Materials Society, ANL-94/28 (1994). Warrendale, Pennsylvania (1994). 9. High-Level Waste Borosilicate Glass, A 2. D.P. Abraham, S.M. McDeavitt, and J.Y. Park, Compendium of Corrosion Characteristics, ea. “Microstructure and Phase Identification in Type 304 J. J. Cunnane, Vol. 1, U.S. Department of Energy Report Stainless Steel-Zirconium Alloys,” Accepted fbr DOE-EM-0177 (1994). publication in Metall. and Mater. Trans. (1996). 10. L. C. Walters, B. R. Seidel, and J.H. Kittel, “Metallic Fuels and Blankets in LMFBRs,” Nuclear 3. D.P. Abraham, S.M. McDeavitt, and J.Y. Park, “Metal Waste Forms From The Electrometallurgical Technology, 65, 179 (1984). Treatment of Spent Nuclear Fuel,” Proc. DOE Spent Nuclear Fuel & Fissile Material Management, Reno, 11. Y. Lis S. M. Allen, and J. D. Livingston, “An Nevada, June 16-20, 1996, American Nuclear Society, Investigation of FeZr Phase,” Scripta Metallurgica et LaGrange Park, Illinois (1996). Materialia, 32, 1129 (1995).

4. D.D. Keiser, Jr. and S.M. McDeavitt, “Actinide-Containing Metal Disposition Alloys,” Proc. DOE Spent Nuclear Fuel & Fissile Material Management, Reno, Nevada, June 16-20,1996, American Nuclear Society, LaGrange Park, Illinois (1996).