Characterization of Composite Ceramic High Level Waste Forms

ABSTRACTArgonne National Laboratory has developed a composite ceramic waste form for the disposition of high level radioactive waste produced during electrometallurgical conditioning of spent nuclear fuel. The electrorefiner LiCl/KCl eutectic salt, containing fission products and transuranics in the chloride form, is contacted with a zeolite material which removes the fission products from the salt. After salt contact, the zeolite is mixed with a glass binder. The zeolite/glass mixture is then hot isostatic pressed (HIPed) to produce the composite ceramic waste form. The ceramic waste form provides a durable medium that is well suited to incorporate fission products and transuranics in the chloride form. Presented are preliminary results of the process qualification and characterization studies, which include chemical and physical measurements and product durability testing, of the ceramic waste form.

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

MRS Proceedings ◽

10.1557/proc-556-107 ◽

1999 ◽

Vol 556 ◽

Cited By ~ 5

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.

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Consolidating Electrorefined Spent Nuclear Fuel Waste: Analysis and Experiment

Heat Transfer, Volume 1 ◽

10.1115/imece2003-41806 ◽

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.

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

MRS Proceedings ◽

10.1557/proc-556-1193 ◽

1999 ◽

Vol 556 ◽

Cited By ~ 1

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.

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

MRS Proceedings ◽

10.1557/proc-465-433 ◽

1996 ◽

Vol 465 ◽

Cited By ~ 5

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.

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

Microscopy and Microanalysis ◽

10.1017/s1431927600034334 ◽

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.

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

MRS Proceedings ◽

10.1557/proc-713-jj11.21 ◽

2002 ◽

Vol 713 ◽

Cited By ~ 1

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.

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Thermal Analysis for Geologic Codisposal of SNF Packages

10th International Conference on Nuclear Engineering, Volume 4 ◽

10.1115/icone10-22390 ◽

2002 ◽

Author(s):

Si Y. Lee

Keyword(s):

Thermal Analysis ◽

Nuclear Fuel ◽

Spent Nuclear Fuel ◽

Peripheral Region ◽

Waste Form ◽

Temperature History ◽

High Level Waste ◽

Geologic Repository ◽

Geologic Disposal ◽

High Level

The engineering viability of disposal of aluminum-clad, aluminum-based spent nuclear fuel (Al-SNF) in a geologic repository requires a thermal analysis to provide the temperature history of the waste form. Calculated temperatures are used to demonstrate compliance with criteria for waste acceptance into the geologic disposal system and as input to assess the chemical and physical behavior of the waste form within the Waste Package (WP). The leading codisposal WP design proposes that a central DOE Al-SNF canister be surrounded by five Defense Waste Process Facility (DWPF) glass log canisters, that is, High-level Waste Glass Logs (HWGL’s), and placed into a WP in a geologic disposal system. A DOE SNF canister having about 0.4318m diameter is placed along the central horizontal axis of the WP. The five HWGL’s will be located around the peripheral region of the DOE SNF canister within the cylindrical WP container. The codisposal WP will be laid down horizontally in a drift repository. In this situation, two waste form options for Al-SNF disposition are considered using the codisposal WP design configurations. They are the direct Al-SNF form and the melt-dilute ingot. In the present work, the reference geologic and design conditions are assumed for the analysis even though the detailed package design is continuously evolved. This paper primarily dealt with the thermal performance internal to the codisposal WP for the qualification study of the WP containing Al-SNF. Thermal analysis methodology and decay heat source terms have been developed to calculate peak temperatures and temperature profiles of Al-SNF package in the DOE spent nuclear fuel canister within the geologic codisposal WP.

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An Overview of Risk Management Planning for Hot-Isostatic Pressure Treatment of High-Level Waste Calcine for the Idaho Cleanup Project

Volume 5: High-Pressure Technology; ASME NDE Division ◽

10.1115/pvp2012-78150 ◽

2012 ◽

Author(s):

Brian D. Preussner ◽

Joseph A. Nenni ◽

Vondell J. Balls

Keyword(s):

Risk Management ◽

Systems Engineering ◽

Large Scale ◽

Spent Nuclear Fuel ◽

Waste Form ◽

High Level Waste ◽

Simultaneous Application ◽

Time And Motion ◽

High Level ◽

Hip Process

The Calcine Disposition Project (CDP) of the Idaho Cleanup Project (ICP) has the responsibility to retrieve, treat, and dispose of the calcine stored at the Idaho Nuclear Technology and Engineering Center (INTEC) located at the Idaho National Laboratory. Calcine is the granular product of thermally treating, or calcining liquid high-level waste (HLW) that was produced at INTEC during the reprocessing of spent nuclear fuel (SNF) to recover uranium. The CDP is currently designing the Hot Isostatic Pressure (HIP) treatment for the calcine to provide monolithic, glass-ceramic waste form suitable for transport and disposition outside of Idaho by 2035 in compliance with the Idaho Settlement Agreement. The HIP process has been used by industry since its invention, by Battelle Institute, in 1955. Hot isostatic pressing can be used for upgrading castings, densifying pre-sintered components, and consolidate powders. It involves the simultaneous application of a high pressure and temperature in a specially constructed vessel. The pressure is applied on all sides with a gas (usually inert) and, so, is isostatic. The CDP will use this treatment process (10,000 psi at 1,150 C) to combine physically and chemically a mixture of calcine and granular additives into a non leachable waste-form. The HIP process for calcine involves filling a metal can with calcine and additives, heating and evacuating the can to remove volatiles, sealing the can under vacuum, and placing the can within the HIP machine for treatment. Although the HIP process has been in use for over 50 years it has not been applied in large scale radioactive service. Challenges with retrofitting such a system for Calcine treatment include 1) filling and sealing the HIP can cleanly and remotely, 2) remotely loading and unloading the HIP machine, and 3) performing maintenance and repair on a 300 ton, hydraulically actuated machine in a highly radioactive hot cell environment. In this article, a systems engineering approach, including use of industry-proven design-for-quality tools and quantitative assessment techniques is summarized. Discussions on how these techniques were used to improve high-consequence risk management and more effectively apply failure mode, RAMI, and time and motion analyses at the earliest possible stages of design are provided.

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Incorporation of radionuclides from the electrometallurgical treatment of spent fuel into a ceramic waste form

MRS Proceedings ◽

10.1557/proc-556-115 ◽

1999 ◽

Vol 556 ◽

Cited By ~ 11

Author(s):

C. Pereira ◽

M. C. Hash ◽

M. A. Lewis ◽

M. K. Richmann ◽

J. Basco

Keyword(s):

Ion Exchange ◽

Molten Salt ◽

Nuclear Fuel ◽

Argonne National Laboratory ◽

Fission Products ◽

Experimental Results ◽

Waste Form ◽

Chloride Salt ◽

Zeolite A ◽

Ceramic Waste

AbstractAn electrometallurgical process is being developed at Argonne National Laboratory to treat spent metallic nuclear fuel. In this process, the spent nuclear fuel is electrorefined in a molten salt to separate uranium from the other constituents of the fuel. The treatment process generates a contaminated chloride salt that is incorporated into a ceramic waste form. The ceramic waste form, a composite of sodalite and glass, contains the fission products (rare earths, alkalis, alkaline earth metals, and halides) and transuranic radionuclides that accumulated in the electrorefiner salt. These radionuclides are incorporated into zeolite A, which can fully accommodate the salt in its crystal structure. The radionuclides are incorporated into the zeolite by hightemperature blending or by ion exchange. In the blending process the salt and zeolite are simply tumbled together at >450°C (723 K), but in the ion exchange process, which yields a product more highly concentrated in fission products, the molten salt is passed through a bed of the zeolite. In either case, the salt-loaded zeolite A is mixed with glass frit and hot isostatically pressed to produce a monolithic leach resistant waste form.Zeolite is converted to sodalite during hot pressing. This paper presents experimental results on the experimental results on the fission product uptake of the zeolite as a function of time and salt composition.

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Natural Clay Minerals as a Starting Material for Matrices for the Immobilization of Radioactive Waste from Pyrochemical Processing of SNF

Sustainability ◽

10.3390/su131910780 ◽

2021 ◽

Vol 13 (19) ◽

pp. 10780

Author(s):

Anna V. Matveenko ◽

Andrey P. Varlakov ◽

Alexander A. Zherebtsov ◽

Alexander V. Germanov ◽

Ivan V. Mikheev ◽

...

Keyword(s):

Radioactive Waste ◽

Nuclear Fuel ◽

Storage Time ◽

Spent Nuclear Fuel ◽

Fission Products ◽

Bentonite Clay ◽

Ceramic Materials ◽

High Level Waste ◽

Natural Clay ◽

High Level

Pyrochemistry is a promising technology that can provide benefits for the safe reprocessing of relatively fresh spent nuclear fuel with a short storage time (3–5 years). The radioactive waste emanating from this process is an electrolyte (LiCl–KCl) mixture with fission products included. Such wastes are rarely immobilized through common matrices such as cement and glass. In this study, samples of ceramic materials, based on natural bentonite clay, were studied as matrices for radioactive waste in the form of LiCl–KCl eutectic. The phase composition of the samples, and their mechanical, hydrolytic, and radiation resistance were characterized. The possibility of using bentonite clay as a material for immobilizing high-level waste arising from pyrochemical processing of spent nuclear fuel is further discussed in this paper.

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