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 Characterization of Composite Ceramic High Level Waste Forms

1997 ◽  
Vol 481 ◽  
Author(s):  
S. M. Frank ◽  
K. J. Bateman ◽  
T. DiSanto ◽  
S. G. Johnson ◽  
T. L. Moschetti ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Composite Ceramic ◽  
Waste Form ◽  
High Level Waste ◽  
Ceramic Waste ◽  
Chloride Form ◽  
Physical Measurements ◽  
High Level

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.

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.

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.

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 Hot Isostatic Pressing of Ceramic Waste From Spent Nuclear Fuel

2002 ◽  
Author(s):  
Kenneth J. Bateman ◽  
Richard H. Rigg ◽  
James D. Wiest
Keyword(s):  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Hot Isostatic Pressing ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Spent Fuel ◽  
Isostatic Pressing ◽  
Ceramic Waste ◽  
Pressing Operation ◽  
Waste Salt

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 electrometallurgical treatment of spent Experimental Breeder Reactor-II fuel. The ceramic waste process culminated with a hot isostatic pressing operation. This paper reviews the installation and operation of a hot isostatic press in a radioactive environment. Processing conditions for the hot isostatic press are presented for non-irradiated material and irradiated material. Sufficient testing was performed to demonstrate that a hot isostatic press could be used as the final step of the processing of ceramic waste for the electrometallurgical spent fuel treatment process.

A Methodology for the Integration of Passive System Reliability With Success Criteria in a Probabilistic Framework for Advanced Reactors

2016 ◽  
Author(s):  
Acacia J. Brunett ◽  
Dave Grabaskas ◽  
Matthew Bucknor ◽  
Stefano Passerini
Keyword(s):  
System Reliability ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Passive System ◽  
Passive Systems ◽  
Success Criteria ◽  
Mission Success ◽  
Reliability Method ◽  
High Level ◽  
The Impact

GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory are currently engaged in a joint effort to modernize and develop probabilistic risk assessment (PRA) techniques for advanced non-light water reactors. At a high level, the primary outcome of this project will be the development of next-generation PRA methodologies that will enable risk-informed prioritization of safety- and reliability-focused research and development, while also identifying gaps that may be resolved through additional research. A subset of this effort is the development of PRA methodologies that can be used for the determination of passive system reliability while integrating quantitative success criteria into the risk analysis framework. An updated passive system reliability approach has been developed for utilization in the PRISM PRA that systematically characterizes the impact of passive safety systems on key success criteria. This methodology is derived from the Reliability Method for Passive Systems (RMPS), but is refined to explicitly include consideration of overall mission success through satisfaction of success criteria, rather than only focusing on the passive system itself. This paper provides details on the integrated methodology, focusing on the interface between passive system reliability and success criteria. Specific examples for the passive systems/features of interest, RVACS and inherent reactivity feedback, are included. Additionally, aspects of the integrated passive system and success criteria methodology as they relate to the ASME/ANS Non-LWR PRA Standard are identified and discussed.

Thermal Analysis for Geologic Codisposal of SNF Packages

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.

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

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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