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Fuel-cladding interaction layers in irradiated U-ZR and U-PU-ZR fuel elements.

Description: Argonne National Laboratory is developing an electrometallurgical treatment for spent nuclear fuels. The initial demonstration of this process is being conducted on U-Zr and U-Pu-Zr alloy fuel elements irradiated in the Experimental Breeder Reactor-II (EBR-II). The electrometallurgical treatment process extracts usable uranium from irradiated fuel elements and places residual fission products, actinides, process Zr, and cladding hulls (small segments of tubing) into two waste forms--a ceramic and a metal alloy. The metal waste form will contain the cladding hulls, Zr, and noble metal fission products, and it will be disposed of in a geologic repository. As a result, the expected composition of the waste form will need to be well understood. This report deals with the condition of the cladding, which will make up a large fraction of the metal waste form, after irradiation in EBR-II and before insertion into the electrorefiner. Specifically, it looks at layers that can be found on the inner surface of the cladding due to in-reactor interactions between the alloy fuel and the stainless steel cladding that occurs after the fuel has swelled and contacted the cladding. Many detailed examinations of fuel elements irradiated in EBR-II have been completed and are discussed in the context of interaction layer formation in irradiated cladding. The composition and thickness of the developed interaction layers are identified, along with the irradiation conditions, cladding type, and axial location on fuel elements where the thickest interaction layers can be expected to develop. It has been found that the largest interaction zones are observed at combined high power and high temperature regions of fuel elements and for fuel elements with U-Pu-Zr alloy fuel and D9 stainless steel cladding. The most prevalent, non-cladding constituent observed in the developed interaction layers are the lanthanide fission products.
Date: January 23, 2006
Creator: Keiser, D. D.
Partner: UNT Libraries Government Documents Department

Mechanical properties test data for structural materials. Semiannual progress report for period ending July 31, 1980

Description: Mechanical property investigations of Alloy 718 given either the 954/sup 0/C conventional or the INEL heat treatment are continuing. Current conventional heat-treat data include tests showing the effects of surface finish, product variability, and thermal exposure on the high-cycle fatigue properties; creep-fatigue tests at 538, 593, 649, and 704/sup 0/C with 0.1 hour hold times at peak strain; and stress-rupture tests of notched and smooth specimens showing the effect of pretest thermal exposure. A few stress-rupture tests of weld and base metals given the INEL heat treatment are also reported. High-cycle fatigue tests of Type 316 stainless steel at 593/sup 0/C are reported and compared with previous data from other sources.
Date: January 1, 1980
Creator: Keiser, D.D.
Partner: UNT Libraries Government Documents Department

Review of the physical metallurgy of Alloy 718

Description: The physical metallurgy of Alloy 718 is updated to 1976 on the basis of a survey of post-1967 literature and current experimental data. Composition, microstructures, and mechanical properties are correlated with heat treatment parameters. The current state of understanding of phase stability, strengthening mechanisms, deformation modes, recovery, and recrystallization in this material is described.
Date: February 1, 1976
Creator: Keiser, D.D. & Brown, H.L.
Partner: UNT Libraries Government Documents Department

Microstructural development in waste form alloys cast from irradiated cladding residual from the electrometallurgical treatment of EBR-II spent fuel.

Description: A metallic waste form alloy that consists primarily of stainless steel and zirconium is being developed by Argonne National Laboratory to contain metallic waste constituents that are residual from an electrometallurgical treatment process for spent nuclear fuel. Ingots have been cast in an induction furnace in a hot cell using actual, leftover, irradiated, EBR-II cladding hulls treated in an electrorefiner. The as-cast ingots have been sampled using a core-drilling and an injection-casting technique. In turn, generated samples have been characterized using chemical analysis techniques and a scanning electron microscope equipped with energy dispersive and wavelength-dispersive spectrometers. As-cast ingots contain the predicted concentration levels of the various constituents, and most of the phases that develop are analogous to those for alloys generated using non-radioactive surrogates for the various fission products.
Date: June 10, 1999
Creator: Keiser, D. D., Jr.
Partner: UNT Libraries Government Documents Department

A review of compatibility of IFR fuel and austenitic stainless steel

Description: Interdiffusion experiments have been conducted to investigate the compatibility of various austenitic stainless steels with U-Pu-Zr alloys, which are alloys to be employed as fuel for the Integral Fast Reactor being developed by Argonne National Laboratory. These tests have also studied the compatibility of austenitic stainless steels with fission products, like the minor actinides (Np and Am) and lanthanides (Ce and Nd), that are generated during the fission process in an IFR. This paper compares the results of these investigations in the context of fuel-cladding compatibility in IFR fuel elements, specifically focusing on the relative Interdiffusion behavior of the components and the types of phases that develop based on binary phase diagrams. Results of Interdiffusion tests are assessed in the light of observations derived from post-test examinations of actual irradiated fuel elements.
Date: November 1, 1996
Creator: Keiser, D.D. Jr.
Partner: UNT Libraries Government Documents Department

Yucca Mountain project canister material corrosion studies as applied to the electrometallurgical treatment metallic waste form

Description: Yucca Mountain, Nevada is currently being evaluated as a potential site for a geologic repository. As part of the repository assessment activities, candidate materials are being tested for possible use as construction materials for waste package containers. A large portion of this testing effort is focused on determining the long range corrosion properties, in a Yucca Mountain environment, for those materials being considered. Along similar lines, Argonne National Laboratory is testing a metallic alloy waste form that also is scheduled for disposal in a geologic repository, like Yucca Mountain. Due to the fact that Argonne`s waste form will require performance testing for an environment similar to what Yucca Mountain canister materials will require, this report was constructed to focus on the types of tests that have been conducted on candidate Yucca Mountain canister materials along with some of the results from these tests. Additionally, this report will discuss testing of Argonne`s metal waste form in light of the Yucca Mountain activities.
Date: November 1, 1996
Creator: Keiser, D.D.
Partner: UNT Libraries Government Documents Department

Consolidation of cladding hulls from the electrometallurgical treatment of spent fuel.

Description: To consolidate metallic waste that is residual from Argonne National Laboratory's electrometallurgical treatment of spent nuclear fuel, waste ingots are currently being cast using an induction furnace located in a hot cell. These ingots, which have been developed to serve as final waste forms destined for repository disposal, are stainless steel (SS)-Zr alloys (the Zr is very near 15 wt.%). The charge for the alloys consists of stainless steel cladding hulls, Zr from the fuel being treated, noble metal fission products, and minor amounts of actinides that are present with the cladding hulls. The actual in-dated cladding hulls have been characterized before they were melted into ingots, and the final as-cast ingots have been characterized to determine the degree of consolidation of the charge material. It has been found that ingots can be effectively cast from irradiated cladding hulls residual from the electrometallurgical treatment process by employing an induction furnace located in a hot cell.
Date: April 10, 1998
Creator: Keiser, D. D., Jr.
Partner: UNT Libraries Government Documents Department

CHARACTERIZATION OF MONOLITHIC FUEL FOIL PROPERTIES AND BOND STRENGTH

Description: Understanding fuel foil mechanical properties, and fuel / cladding bond quality and strength in monolithic plates is an important area of investigation and quantification. Specifically, what constitutes an acceptable monolithic fuel – cladding bond, how are the properties of the bond measured and determined, and what is the impact of fabrication process or change in parameters on the level of bonding? Currently, non-bond areas are quantified employing ultrasonic determinations that are challenging to interpret and understand in terms of irradiation impact. Thus, determining mechanical properties of the fuel foil and what constitutes fuel / cladding non-bonds is essential to successful qualification of monolithic fuel plates. Capabilities and tests related to determination of these properties have been implemented at the INL and are discussed, along with preliminary results.
Date: March 1, 2007
Creator: Burkes, D E; Keiser, D D; Wachs, D M; Larson, J S & Chapple, M D
Partner: UNT Libraries Government Documents Department

Microstructural Characterization of Cast Metallic Transmutation Fuels

Description: As part of the Global Nuclear Energy Partnership (GNEP) and the Advanced Fuel Cycle Initiative (AFCI), the US Department of Energy (DOE) is participating in an international collaboration to irradiate prototypic actinide-bearing transmutation fuels in the French Phenix fast reactor (FUTURIX-FTA experiment). The INL has contributed to this experiment by fabricating and characterizing two compositions of metallic fuel; a non-fertile 48Pu-12Am-40Zr fuel and a low-fertile 35U-29Pu-4Am-2Np-30Zr fuel for insertion into the reactor. This paper highlights results of the microstructural analysis of these cast fuels, which were reasonably homogeneous in nature, but had several distinct phase constituents. Spatial variations in composition appeared to be more pronounced in the low-fertile fuel when compared to the non-fertile fuel.
Date: September 1, 2007
Creator: Cole, J. I.; Keiser, D. D. & Kennedy, J. R.
Partner: UNT Libraries Government Documents Department

Characterization and Testing of Monolithic RERTR Fuel Plates

Description: Monolithic fuel plates are being developed for application in research reactors throughout the world. These fuel plates are comprised of a U-Mo alloy foil encased in aluminum alloy cladding. Three different fabrication techniques have been looked at for producing monolithic fuel plates: hot isostatic pressing (HIP), transient liquid phase bonding (TLPB), and friction stir welding (FSW). Of these three techniques, HIP and FSW are currently being emphasized. As part of the development of these fabrication techniques, fuel plates are characterized and tested to determine properties like hardness and the bond strength at the interface between the fuel and cladding. Testing of HIPed samples indicates that the foil/cladding interaction behavior depends on the Mo content in the U-Mo foil, the measured hardness values are quite different for the fuel, cladding, and interaction zone phase and Ti, Zr and Nb are the most effective diffusion barriers. For FSW samples, there is a dependence of the bond strength at the foil/cladding interface on the type of tool that is employed for performing the actual FSW process.
Date: March 1, 2007
Creator: Keiser, D. D.; Jue, J. F. & Burkes, D. E.
Partner: UNT Libraries Government Documents Department

U-Mo Foil/Cladding Interactions in Friction Stir Welded Monolithic RERTR Fuel Plates

Description: Interaction between U-Mo fuel and Al has proven to dramatically impact the overall irradiation performance of RERTR dispersion fuels. It is of interest to better understand how similar interactions may affect the performance of monolithic fuel plates, where a uranium alloy fuel is sandwiched between aluminum alloy cladding. The monolithic fuel plate removes the fuel matrix entirely, which reduces the total surface area of the fuel that is available to react with the aluminum and moves the interface between the fuel and cladding to a colder region of the fuel plate. One of the major fabrication techniques for producing monolithic fuel plates is friction stir welding. This paper will discuss the interactions that can occur between the U-Mo foil and 6061 Al cladding when applying this fabrication technique. It has been determined that the time at high temperatures should be limited as much as is possible during fabrication or any post-fabrication treatment to reduce as much as possible the interactions between the foil and cladding. Without careful control of the fabrication process, significant interaction between the U-Mo foil and Al alloy cladding can result. The reaction layers produced from such interactions can exhibit notably different morphologies vis-à-vis those typically observed for dispersion fuels.
Date: October 1, 2006
Creator: Keiser, D.D.; Jue, J.F. & Clark, C.R.
Partner: UNT Libraries Government Documents Department

AN EVALUATION OF POTENTIAL LINER MATERIALS FOR ELIMINATING FCCI IN IRRADIATED METALLIC NUCLEAR FUEL ELEMENTS

Description: Metallic nuclear fuels are being looked at as part of the Global Nuclear Energy Program for transmuting longlive transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products. In order to optimize the performance of these fuels, the concept of using liners to eliminate the fuel/cladding chemical interactions that can occur during irradiation of a fuel element has been investigated. The potential liner materials Zr and V have been tested using solid-solid diffusion couples, consisting of liner materials butted against fuel alloys and against cladding materials. The couples were annealed at the relatively high temperature of 700°C. This temperature would be the absolute maximum temperature present at the fuel/cladding interface for a fuel element in-reactor. Analysis was performed using a scanning electron microscope equipped with energy-dispersive and wavelengthdispersive spectrometers (SEM/EDS/WDS) to evaluate any developed diffusion structures. At 700°C, minimal interaction was observed between the metallic fuels and either Zr or V. Similarly, limited interaction was observed between the Zr and V and the cladding materials. The best performing liner material appeared to be the V, based on amounts of interaction.
Date: September 1, 2007
Creator: Keiser, D. D. & Cole, J. I.
Partner: UNT Libraries Government Documents Department

A powder metallurgy approach for production of innovative radioactive waste forms

Description: The feasibility of producing a single metal-matrix composite form rather than two separate forms consisting of a cast metal alloy ingot (such as Type 316SS + Zr) and a ceramic glass-bonded zeolite Na{sub 12}(AlO{sub 2}){sub 12}(SiO{sub 2}){sub 12} has been demonstrated. This powder metallurgy approach consists of mixing the powder of the two separate waste forms together followed by compaction by hot isostatic pressing. Such a radioactive waste form would have the potential advantages of reducing the total waste volume, good thermal conductivity, stability, and surfaces with limited oxide layer formation. 5 refs., 8 figs., 2 tabs.
Date: July 1, 1997
Creator: Keiser, D.D. Jr.; Crawford, D.C. & Bhaduri, S.
Partner: UNT Libraries Government Documents Department

Actinide-containing metal disposition alloys

Description: Argonne National Laboratory is currently developing an electro-metallurgical process for treating a wide array of spent nuclear fuels. As part of this process, two waste streams will be consolidated into waste forms; one will be a mineral and the other a metal alloy. The metal waste form is an alloy that contains cladding hulls, ``noble`` metal fission products, and Zr from alloy fuels. The nominal composition of the metal waste form alloys are stainless steel-15 wt.% Zr (SS-15Zr) for stainless steel clad fuel and Zircaloy-8 wt.% stainless steel (Zr-8SS) for Zircaloy clad fuel, with both alloys also containing up to 4 wt.% noble metal fission products. This paper investigates using the two nominal metal alloy compositions described above as a possible Pu and TRU disposition form.
Date: May 1, 1996
Creator: Keiser, D.D. Jr. & McDeavitt, S.M.
Partner: UNT Libraries Government Documents Department

Monitoring the consistency of the metallic waste form derived from electrometallurgical processing.

Description: A metallic waste form alloy that consists primarily of stainless steel and zirconium is being developed by Argonne National Laboratory to contain metallic waste constituents that are residual from an electrometallurgical treatment process for spent nuclear fuel. An approach for monitoring the consistency of metallic waste forms (MWFs) is developed based on consideration of the intent of regulatory requirements, production method, measured physical and chemical properties of the MWF, and analytical capabilities. It is recommended that the Zr content of the MWF be measured and tracked to monitor consistency because the Zr content: (1) provides a measure of the amount of the Zr(Fe,Ni,Cr){sub 2+x} intermetallic phase, which sequesters the majority of radionuclides in the MWF and affects its physical robustness and (2) indicates that the desired Fe-Zr eutectic was obtained, which provides a direct indicator that the appropriate process conditions (time, temperature) were employed. It is recommended that the Zr content be measured by chemical analyses of drill shavings taken from the MWF products.
Date: August 26, 2002
Creator: Keiser, D. D., Jr.; Johnson, S. G. & Ebert, W. L.
Partner: UNT Libraries Government Documents Department

TEM CHARACTERIZATION OF IRRADIATED U3SI2/AL DISPERSION FUEL

Description: The silicide dispersion fuel of U3Si2/Al has been recognized as a reasonably good performance fuel for nuclear research and test reactors except that it requires the use of high enrichment uranium. An irradiated U3Si2/Al dispersion fuel (~75% enrichment) from the high flux side of a RERTR-8 (U0R040) plate was characterized using transmission electron microscopy (TEM). The fuel plate was irradiated in the advanced test reactor (ATR) for 105 days. The average irradiation temperature and fission density of the fuel particles for the TEM sample are estimated to be approximately ~110 degrees C and 5.4 x 10-21 f/cm3. The characterization was performed using a 200KV TEM with a LaB6 filament. Detailed microstructural information along with composition analysis is obtained. The results and their implication on the performance of this silicide fuel are discussed.
Date: October 1, 2010
Creator: Gan, J.; Miller, B.; Keiser, D.; Robinson, A.; Medvedev, P. & Wachs, D.
Partner: UNT Libraries Government Documents Department

US RERTR FUEL DEVELOPMENT POST IRRADIATION EXAMINATION RESULTS

Description: Post irradiation examinations of irradiated RERTR plate type fuel at the Idaho National Laboratory have led to in depth characterization of fuel behavior and performance. Both destructive and non-destructive examination capabilities at the Hot Fuels Examination Facility (HFEF) as well as recent results obtained are discussed herein. New equipment as well as more advanced techniques are also being developed to further advance the investigation into the performance of the high density U-Mo fuel.
Date: October 1, 2008
Creator: Robinson, A. B.; Wachs, D. M.; Burkes, D. E. & Keiser, D. D.
Partner: UNT Libraries Government Documents Department

Stainless steel-zirconium alloy waste forms

Description: An electrometallurgical treatment process has been developed by Argonne National Laboratory to convert various types of spent nuclear fuels into stable storage forms and waste forms for repository disposal. The first application of this process will be to treat spent fuel alloys from the Experimental Breeder Reactor-II. Three distinct product streams emanate from the electrorefining process: (1) refined uranium; (2) fission products and actinides extracted from the electrolyte salt that are processed into a mineral waste form; and (3) metallic wastes left behind at the completion of the electrorefining step. The third product stream (i.e., the metal waste stream) is the subject of this paper. The metal waste stream contains components of the chopped spent fuel that are unaffected by the electrorefining process because of their electrochemically ``noble`` nature; this includes the cladding hulls, noble metal fission products (NMFP), and, in specific cases, zirconium from metal fuel alloys. The selected method for the consolidation and stabilization of the metal waste stream is melting and casting into a uniform, corrosion-resistant alloy. The waste form casting process will be carried out in a controlled-atmosphere furnace at high temperatures with a molten salt flux. Spent fuels with both stainless steel and Zircaloy cladding are being evaluated for treatment; thus, stainless steel-rich and Zircaloy-rich waste forms are being developed. Although the primary disposition option for the actinides is the mineral waste form, the concept of incorporating the TRU-bearing product into the metal waste form has enough potential to warrant investigation.
Date: July 1996
Creator: McDeavitt, S. M.; Abraham, D. P.; Keiser, D. D., Jr. & Park, J. Y.
Partner: UNT Libraries Government Documents Department

Alloy waste forms for metal fission products and actinides isolated by spent nuclear fuel treatment

Description: Waste form alloys are being developed at Argonne National Laboratory for the disposal of remnant metallic wastes from an electrometallurgical process developed to treat spent nuclear fuel. This metal waste form consists of the fuel cladding (stainless steel or Zircaloy), noble metal fission products (e.g., Ru, Pd, Mo and Tc), and other metallic wastes. The main constituents of the metal waste stream are the cladding hulls (85 to 90 wt%); using the hulls as the dominant alloying component minimizes the overall waste volume as compared to vitrification or metal encapsulation. Two nominal compositions for the waste form are being developed: (1) stainless steel-15 wt% zirconium for stainless steel-clad fuels and (2) zirconium-8 wt% stainless steel for Zircaloy-clad fuels. The noble metal fission products are the primary source of radiation in the metal waste form. However, inclusion of actinides in the metal waste form is being investigated as an option for interim or ultimate storage. Simulated waste form alloys were prepared and analyzed to determine the baseline alloy microstructures and the microstructural distribution of noble metals and actinides. Corrosion tests of the metal waste form alloys indicate that they are highly resistant to corrosion.
Date: October 1, 1996
Creator: McDeavitt, S.M.; Abraham, D.P.; Keiser, D.D. Jr. & Park, J.Y.
Partner: UNT Libraries Government Documents Department

Initial results of metal waste form development activities at ANL-West

Description: Argonne National Laboratory is developing a metal alloy to contain metallic waste constituents from the electrometallurgical treatment of spent nuclear fuel. This alloy will contain stainless steel (from stainless steel-clad fuel elements), {approximately}15 wt.% zirconium (from alloy fuel), fission products noble to the process (e.g., Ru, Pd, Tc, etc.), and minor amounts of actinides. The alloy will serve as a final waste form for these components and will be disposed of in a geologic repository. The alloy ingot is produced in an induction furnace situated in a hot cell using Ar cover gas. This paper discusses results from the melting campaigns that have been initiated at ANL-West to generate the metal waste form using actual process materials. In addition, metal waste form samples have been doped with Tc and selected actinides and are described in the context of how elements of interest partition between various phases in the alloy and how this distribution of elements in the alloy may affect the leaching behavior of the components in an aqueous environment. 3 refs.
Date: October 1, 1997
Creator: Keiser, D.D. Jr.; Westphal, B.R.; Hersbt, R.S. & Johnson, S.G.
Partner: UNT Libraries Government Documents Department

The effect of actinides on the microstructural development in a metallic high-level nuclear waste form

Description: Waste forms to contain material residual from an electrometallurgical treatment of spent nuclear fuel have been developed by Argonne National Laboratory. One of these waste forms contains waste stainless steel (SS), fission products that are noble to the process (e.g., Tc, Ru, Pd, Rh), Zr, and actinides. The baseline composition of this metallic waste form is SS-15wt.% Zr. The metallurgy of this baseline alloy has been well characterized. On the other hand, the effects of actinides on the alloy microstructure are not well understood. As a result, SS-Zr alloys with added U, Pu, and/or Np have been cast and then characterized, using scanning electron microscopy, transmission electron microscopy, and neutron diffraction, to investigate the microstructural development in SS-Zr alloys that contain actinides. Actinides were found to congregate non-uniformally in a Zr(Fe,Cr,Ni){sub 2+x} phase. Apparently, the actinides were contained in varying amounts in the different polytypes (C14, C15, and C36) of the Zr(Fe,Cr,Ni){sub 2+x} phase. Heat treatment of an actinide-containing SS-15 wt.% Zr alloy showed the observed microstructure to be stable.
Date: October 25, 1999
Creator: Keiser, D. D., Jr.; Sinkler, W.; Abraham, D. P.; Richardson, J. W., Jr. & McDeavitt, S. M.
Partner: UNT Libraries Government Documents Department

Leaching characteristics of the metal waste form from the electrometallurgical treatment process: Product consistency testing

Description: Argonne National Laboratory is developing an electrometallurgical treatment for spent fuel from the experimental breeder reactor II. A product of this treatment process is a metal waste form that incorporates the stainless steel cladding hulls, zirconium from the fuel and the fission products that are noble to the process, i.e., Tc, Ru, Nb, Pd, Rh, Ag. The nominal composition of this waste form is stainless steel/15 wt% zirconium/1--4 wt% noble metal fission products/1--2 wt % U. Leaching results are presented from several tests and sample types: (1) 2 week monolithic immersion tests on actual metal waste forms produced from irradiated cladding hulls, (2) long term (>2 years) pulsed flow tests on samples containing technetium and uranium and (3) crushed sample immersion tests on cold simulated metal waste form samples. The test results will be compared and their relevance for waste form product consistency testing discussed.
Date: November 11, 1999
Creator: Johnson, S. G.; Keiser, D. D.; Frank, S. M.; DiSanto, T. & Noy, M.
Partner: UNT Libraries Government Documents Department

Electrochemical corrosion testing of metal waste forms

Description: Electrochemical corrosion tests have been conducted on simulated stainless steel-zirconium (SS-Zr) metal waste form (MWF) samples. The uniform aqueous corrosion behavior of the samples in various test solutions was measured by the polarization resistance technique. The data show that the MWF corrosion rates are very low in groundwaters representative of the proposed Yucca Mountain repository. Galvanic corrosion measurements were also conducted on MWF samples that were coupled to an alloy that has been proposed for the inner lining of the high-level nuclear waste container. The experiments show that the steady-state galvanic corrosion currents are small. Galvanic corrosion will, hence, not be an important mechanism of radionuclide release from the MWF alloys.
Date: December 14, 1999
Creator: Abraham, D. P.; Peterson, J. J.; Katyal, H. K.; Keiser, D. D. & Hilton, B. A.
Partner: UNT Libraries Government Documents Department