scholarly journals Dissolution Testing of a Metallic Waste Form in Chloride Brine

2006 ◽  
Vol 985 ◽  
Author(s):  
Dawn E. Janney
Keyword(s):  
National Laboratory ◽  
Bulk Composition ◽  
Argonne National Laboratory ◽  
Corrosion Products ◽  
Localized Corrosion ◽  
Waste Form ◽  
Ph Values ◽  
Surface Areas ◽  
Sample Edge ◽  
Static Immersion

AbstractArgonne National Laboratory has developed an electrometallurgical process for conditioning spent sodium-bonded metallic reactor fuel from the Experimental Breeder Reactor II (EBR-II). One waste stream from this process consists of a metal waste form (MWF) whose baseline composition is stainless steel alloyed with 15 wt% Zr (SS-15Zr) and whose microstructure is a eutectic intergrowth of iron solid solutions and Fe-Zr-Cr-Ni intermetallics. This paper reports scanning electron microscope (SEM) observations of corrosion products formed during static immersion tests in which coupons of surrogate MWF containing 10 wt% U (SS-15Zr-10U) were immersed in solutions with nominal pH values of 3 and 4 and 1000 ppm added chloride for 70 days at 50 °C. Although the majority of the surface areas of the coupons appear unchanged, linear areas with localized corrosion products apparently consisting of porous materials overlying corrosion-product-filled channels formed on both coupons, cross-cutting phase boundaries in the original eutectic microstructures. Many of the linear areas intersected the sample edge at notches present before the tests or followed linear flaws visible in pre-test images. Compositions of corrosion products differed significantly from the bulk composition, and the maximum observed concentration of U exceeded that reported in actinide-bearing phases in uncorroded surrogate MWF samples with comparable concentrations of U.

TEM Analysis of Corrosion Products From a Radioactive Stainless Steel-based Alloy

2000 ◽  
Vol 6 (S2) ◽  
pp. 368-369
Author(s):  
N.L. Dietz ◽  
D.D Keiser
Keyword(s):  
Stainless Steel ◽  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Waste Form ◽  
Metallic Sodium ◽  
Geologic Repository ◽  
Tem Analysis ◽  
Waste Forms

Argonne National Laboratory has developed an electrometallurgical treatment process for metallic spent nuclear fuel from the Experimental Breeder Reactor-II. This process stabilizes metallic sodium and separates usable uranium from fission products and transuranic elements that are contained in the fuel. The fission products and other waste constituents are placed into two waste forms: a ceramic waste form that contains the transuranic elements and active fission products such as Cs, Sr, I and the rare earth elements, and a metal alloy waste form composed primarily of stainless steel (SS), from claddings hulls and reactor hardware, and ∼15 wt.% Zr (from the U-Zr and U-Pu-Zr alloy fuels). The metal waste form (MWF) also contains noble metal fission products (Tc, Nb, Ru, Rh, Te, Ag, Pd, Mo) and minor amounts of actinides. Both waste forms are intended for eventual disposal in a geologic repository.

Preliminary Studies of the Disposition of Cesium in a Glass-Bonded Sodalite Waste form

2002 ◽  
Vol 713 ◽  
Author(s):  
Marsha J. Lambregts ◽  
Steven M. Frank
Keyword(s):  
National Laboratory ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Waste Form ◽  
Resonance Spectroscopy ◽  
Ceramic Waste ◽  
Long Term Storage ◽  
Geological Repository ◽  
Term Storage

ABSTRACTArgonne National Laboratory has developed an electrometallurgical treatment for DOE spent metallic nuclear fuel. Fission products are immobilized in a durable glass bonded sodalite ceramic waste form (CWF) suitable for long term storage in a geological repository. Cesium is estimated to be in the waste form at approximately 0.1 wt.%. The exact disposition of cesium was uncertain and it was believed to be uniformly distributed throughout the waste form. A correlation of X-ray diffractometry (XRD), electron microscopy (EM), and nuclear magnetic resonance spectroscopy (NMR) performed on surrogate ceramic waste forms with high cesium loadings found a high cesium content in the glass phase and in several non-sodalite aluminosilicate phases. Cesium was not detected in the sodalite phase.

scholarly journals Development of a ceramic waste form for high-level waste disposal

1999 ◽  
Vol 556 ◽  
Author(s):  
D. W. Esh ◽  
K. M. Goff ◽  
K. T. Hirsche ◽  
T. J. Battisti ◽  
M. F. Simpson ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Waste Form ◽  
High Level Waste ◽  
Durability Test ◽  
Ceramic Waste ◽  
High Level ◽  
The Impact

AbstractA ceramic waste form is being developed by Argonne National Laboratory* (ANL) as part of the demonstration of the electrometallurgical treatment of spent nuclear fuel [1]. The halide, alkaline earth, alkali, transuranic, and rare earth fission products are stabilized in zeolite which is combined with glass and processed in a hot isostatic press (HIP) to form a ceramic composite. The mineral sodalite is formed in the HIP from the zeolite precursor. The process, from starting materials to final product, is relatively simple. An overview of the processing operations is given. The metrics that have been developed to measure the success or completion of processing operations are developed and discussed. The impact of variability in processing metrics on the durability of the final product is presented. The process is demonstrated to be robust for the type and range of operation metrics considered and the performance metric (PCT durability test) against which the operation metrics are evaluated.

scholarly journals NNWSI waste form testing at Argonne National Laboratory; Semiannual report: January-June 1987

1988 ◽  
Author(s):  
J.K. Bates ◽  
T.J. Gerding ◽  
T.A. Jr. Abrajano ◽  
W.L. Ebert ◽  
J.J. Mazer
Keyword(s):  
National Laboratory ◽  
Argonne National Laboratory ◽  
Waste Form

Consolidating Electrorefined Spent Nuclear Fuel Waste: Analysis and Experiment

2003 ◽  
Author(s):  
K. J. Bateman ◽  
D. D. Capson
Keyword(s):  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Waste Form ◽  
Processing Conditions ◽  
Production Scale ◽  
Ceramic Waste ◽  
Waste Salt ◽  
Nuclear Fuel Waste

Argonne National Laboratory has developed a process to immobilize waste salt containing fission products, uranium, and transuranic elements as chlorides in a glass-bonded ceramic waste form. This salt was generated in the electrorefining operation used in the electrometallurical treatment of spent EBR-II fuel. The ceramic waste process culminates with an elevated temperature operation. The processing conditions used by the furnace, for demonstration scale and production scale operations, are to be developed at Argonne National Laboratory. To assist in selecting the processing conditions of the furnace and to reduce the number of costly experiments, a finite difference model was developed to predict the consolidation of the ceramic waste. The model accurately predicted the heating as well as the bulk density of the ceramic waste form. The methodology used to develop the computer model and a comparison of the analysis to experimental data is presented.

scholarly journals Microstructure and Leaching Characteristics of a Technetium Containing Metal Waste Form

1999 ◽  
Vol 556 ◽  
Author(s):  
S. G. Johnson ◽  
D. D. Keiser ◽  
M. Noy ◽  
T. O'Holleran ◽  
S. M. Frank
Keyword(s):  
Stainless Steel ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Point Of View ◽  
Waste Form ◽  
Nominal Composition ◽  
Spent Fuel ◽  
Waste Forms ◽  
Metal Waste

AbstractArgonne 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, Pd, Rh, Ag. The nominal composition of this waste form is stainless steel/15 wt% zirconium/ 1–4 wt% noble metal fission products. The behavior of technetium is of particular importance from a disposal point of view for this waste form due to its long half life, 2.14E5 years, and its mobility in groundwater. To address these concerns a limited number of spiked metal waste forms were produced containing Tc. These surrogate waste forms were then studied using scanning electron microscopy (SEM) and selected leaching tests.

Production And Characterization Of A Plutonium-238 Loaded Ceramic Waste Form For The Study Of The Effects Of Alpha Decay Damage

1999 ◽  
Vol 556 ◽  
Author(s):  
M. Steven ◽  
Steven M. Frank ◽  
David W. Esh ◽  
Stephen G. Johnson ◽  
Marianne Noy ◽  
...  
Keyword(s):  
Ceramic Material ◽  
Spent Nuclear Fuel ◽  
Specific Activity ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
High Specific Activity ◽  
Fission Products ◽  
Alpha Decay ◽  
Waste Form ◽  
Ceramic Waste

AbstractArgonne National Laboratory has developed a glass-bonded sodalite ceramic waste form to immobilize fission products and plutonium that accumulate during the electrometallurgical conditioning of spent nuclear fuel. To investigate the effects of alpha decay damage on the structure and leaching characteristics of the ceramic material, 238Pu has been incorporated into the ceramic waste form. The 238pu,with its high specific activity, significantly increases the rate of alpha damage to the waste form. Long term studies have begun with periodic examination of the 238Pu loaded ceramic material. A number of characterization techniques are used to study the alpha decay damage on the structure of the waste form. In addition, PCT type leachate studies will be performed to determine the effect of alpha decay damage on the durability of the ceramic waste form. Preliminary results from this study are presented.

scholarly journals Effect of Different Glass and Zeolite-a Compositions on the Leach Resistance of Ceramic Waste Forms

1996 ◽  
Vol 465 ◽  
Author(s):  
M. A. Lewis ◽  
M. Hash ◽  
D. Glandorf
Keyword(s):  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Hot Isostatic Pressing ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Waste Form ◽  
Zeolite A ◽  
Ceramic Waste ◽  
Leach Test ◽  
Salt Phase

ABSTRACTA ceramic waste form is being developed at Argonne National Laboratory for waste generated during the electrometallurgical treatment of spent nuclear fuel. The waste is generated when fission products are removed from the electrolyte, LiCI-KCl eutectic. The ceramic waste form is a composite, fabricated by hot isostatic pressing a mixture of glass frit and zeolite occluded with fission products and salt. Past work has shown that the normalized release rate (NRR) is less than 1 g/m2d for all elements in a Material Characterization Center-Type 1 (MCC-1) leach test run for 28 days in deionized water at 90°C (363 K). This leach resistance is comparable to that of early Savannah River glasses. We are investigating how leach resistance is affected by changes in the cationic form of zeolite and in the glass composition. Composites were made with three forms of zeolite A and six glasses. We used three-day ASTM C1220–92 (formerly MCC-1) leach tests to screen samples for development purposes only. The leach test results show that the glass composites of zeolites 5A and 4A retain fission products equally well. The loss of cesium is small, varying from 0.1 to 0.5 wt%, while the loss of divalent and trivalent fission products is one or more orders of magnitude smaller. Composites of 5A retain chloride ion better in these short-term screens than 4A and 3A. The more leach resistant composites were made with durable glasses that were rich in silica and poor in alkaline earth oxides. The x-ray diffraction (XRD) results show that a salt phase was absent in the leach resistant composites of 5A and the better glasses but was present in the other composites with poorer leach performance. Thus, the data show that the absence of a salt phase in a composite's XRD pattern corresponds to improved leach resistance. The data also suggest that the interactions between the zeolite and glass depend on the composition of both.

Characterization of a Plutonium-Bearing Zirconolite-Rich Synroc

1996 ◽  
Vol 465 ◽  
Author(s):  
E. C. Buck ◽  
B. Ebbinghaus ◽  
A. J. Bakel ◽  
J. K. Bates
Keyword(s):  
Oxidation State ◽  
Nuclear Weapons ◽  
Lawrence Livermore National Laboratory ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Waste Form ◽  
Ceramic Waste

ABSTRACTA titanate-based ceramic waste form, rich in phases structurally related to zirconolite (CaZrTi2O7), is being developed as a possible method for immobilizing excess plutonium from dismantled nuclear weapons. As part of this program, Lawrence Livermore National Laboratory (LLNL) produced several ceramics that were then characterized at Argonne National Laboratory (ANL). The plutonium-loaded ceramic was found to contain a Pu-Gd zirconolite phase but also contained plutonium titanates, Gd-polymignyte, and a series of other phases. In addition, much of the Pu was remained as PuO2-x. The Pu oxidation state in the zirconolite was determined to be mainly Pu4+, although some Pu3+ was believed to be present.

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