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Title: A glass-encapsulated calcium phosphate wasteform for the immobilization of actinide-, fluoride-, and chloride-containing radioactive wastes from the pyrochemical reprocessing of plutonium metal

Abstract

The presence of halide anions in four types of wastes arising from the pyrochemical reprocessing of plutonium required an immobilization process to be developed in which not only the actinide cations but also the halide anions were immobilized in a durable waste form. At AWE, we have developed such a process using Ca3(PO4)2 as the host material. Successful trials of the process with actinide- and Cl-bearing Type I waste were carried out at PNNL where the immobilization of the waste in a form resistant to aqueous leaching was confirmed. Normalized mass losses determined at 40°C and 28 days were 12 x 10-6 g∙m-2 and 2.7 x 10-3 g∙m-2 for Pu and Cl, respectively. Accelerated radiation-induced damage effects are being determined with specimens containing 238Pu. No changes in the crystalline lattice have been detected with XRD after the 239Pu equivalent of 400 years ageing. Confirmation of the process for Type II waste (a oxyhydroxide-based waste) is currently underway at PNNL. Differences in the ionic state of Pu in the four types of waste have required different surrogates to be used. Samarium chloride was used successfully as a surrogate for both Pu(III) and Am(III) chlorides. Initial investigations into the use of HfO2more » as the surrogate for Pu(IV) oxide in Type II waste indicated no significant differences.« less

Authors:
; ; ; ; ;
Publication Date:
Research Org.:
Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
Sponsoring Org.:
USDOE
OSTI Identifier:
909995
Report Number(s):
PNNL-SA-55807
Journal ID: ISSN 0022-3115; JNUMAM; 830403000; TRN: US0704063
DOE Contract Number:
AC05-76RL01830
Resource Type:
Journal Article
Resource Relation:
Journal Name: Journal of Nuclear Materials, 361(1):78-93; Journal Volume: 361; Journal Issue: 1
Country of Publication:
United States
Language:
English
Subject:
12 MANAGEMENT OF RADIOACTIVE WASTES, AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES; 11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; ACTINIDES; ANIONS; CALCIUM PHOSPHATES; CATIONS; CHLORIDES; HALIDES; LEACHING; OXIDES; PLUTONIUM; PYROCHEMICAL REPROCESSING; RADIOACTIVE WASTES; SAMARIUM CHLORIDES; WASTE FORMS; WASTES; X-RAY DIFFRACTION; radiation damage; plutoninum immobilization; spodiosite; whitlockite; glass; chloride immobilization

Citation Formats

Donald, Ian W., Metcalfe, Brian, Fong, Shirley K., Gerrard, Lee A., Strachan, Denis M., and Scheele, Randall D. A glass-encapsulated calcium phosphate wasteform for the immobilization of actinide-, fluoride-, and chloride-containing radioactive wastes from the pyrochemical reprocessing of plutonium metal. United States: N. p., 2007. Web. doi:10.1016/j.jnucmat.2006.11.011.
Donald, Ian W., Metcalfe, Brian, Fong, Shirley K., Gerrard, Lee A., Strachan, Denis M., & Scheele, Randall D. A glass-encapsulated calcium phosphate wasteform for the immobilization of actinide-, fluoride-, and chloride-containing radioactive wastes from the pyrochemical reprocessing of plutonium metal. United States. doi:10.1016/j.jnucmat.2006.11.011.
Donald, Ian W., Metcalfe, Brian, Fong, Shirley K., Gerrard, Lee A., Strachan, Denis M., and Scheele, Randall D. Sat . "A glass-encapsulated calcium phosphate wasteform for the immobilization of actinide-, fluoride-, and chloride-containing radioactive wastes from the pyrochemical reprocessing of plutonium metal". United States. doi:10.1016/j.jnucmat.2006.11.011.
@article{osti_909995,
title = {A glass-encapsulated calcium phosphate wasteform for the immobilization of actinide-, fluoride-, and chloride-containing radioactive wastes from the pyrochemical reprocessing of plutonium metal},
author = {Donald, Ian W. and Metcalfe, Brian and Fong, Shirley K. and Gerrard, Lee A. and Strachan, Denis M. and Scheele, Randall D.},
abstractNote = {The presence of halide anions in four types of wastes arising from the pyrochemical reprocessing of plutonium required an immobilization process to be developed in which not only the actinide cations but also the halide anions were immobilized in a durable waste form. At AWE, we have developed such a process using Ca3(PO4)2 as the host material. Successful trials of the process with actinide- and Cl-bearing Type I waste were carried out at PNNL where the immobilization of the waste in a form resistant to aqueous leaching was confirmed. Normalized mass losses determined at 40°C and 28 days were 12 x 10-6 g∙m-2 and 2.7 x 10-3 g∙m-2 for Pu and Cl, respectively. Accelerated radiation-induced damage effects are being determined with specimens containing 238Pu. No changes in the crystalline lattice have been detected with XRD after the 239Pu equivalent of 400 years ageing. Confirmation of the process for Type II waste (a oxyhydroxide-based waste) is currently underway at PNNL. Differences in the ionic state of Pu in the four types of waste have required different surrogates to be used. Samarium chloride was used successfully as a surrogate for both Pu(III) and Am(III) chlorides. Initial investigations into the use of HfO2 as the surrogate for Pu(IV) oxide in Type II waste indicated no significant differences.},
doi = {10.1016/j.jnucmat.2006.11.011},
journal = {Journal of Nuclear Materials, 361(1):78-93},
number = 1,
volume = 361,
place = {United States},
year = {Sat Mar 31 00:00:00 EDT 2007},
month = {Sat Mar 31 00:00:00 EDT 2007}
}
  • The preparation and properties of a calcium phosphate ceramic wasteform based on the mineral phases apatite and spodiosite are described. This particular ceramic has been found to be an effective host for immobilizing the chloride constituents obtained from the pyrochemical reprocessing of Pu metal. We discuss the crystal phases present in the solids as determined by XRD and the chemical durability of the product in aqueous solution.
  • The choice of borosilicate glasses as a matrix for immobilizing radioactive wastes is largely justified by the properties of the final vitrified product to be stored for a long time: High chemical, thermal, and radiation resistance. The structure of borosilicate glasses makes it possible to use them as a matrix for high and medium level wastes of diverse chemical composition. However, individual components of the wastes - sulfate, chloride, and molybdate-containing components - have a limited solubility in borosilicate glasses. This results in phase separation (stable liquation) in the vitrification process and decreases the quality of the vitrified product.
  • This work consists of experimental batch extraction data for plutonium into 30 volume-percent tri-butyl phosphate at ambient temperature from such a solution matrix and a model of this data using complexation constants from the literature.
  • A a result of its former role as a producer of nuclear weapons components, the Rocky Flats Environmental Technology Site (RFETS), Golden, Colorado accumulated a variety of plutonium-contaminated materials. When the level of contamination exceeded a predetermined level (the economic discard limit), the materials were classified as residues rather than waste and were stored for later recovery of the plutonium. Although large quantities of residues were processed, others, primarily those more difficult to process, remain in storage at the site. It is planned for the residues with lower concentrations of plutonium to be disposed of as wastes at an appropriatemore » disposal facility, probably the Waste Isolation Pilot Plant (WIPP). Because the plutonium concentration is too high or because the physical or chemical form would be difficult to get into a form acceptable to WIPP, it may not be possible to dispose of a portion of the residues at WIPP. The pyrochemical salts are among the residues that are difficult to dispose of. For a large percentage of the pyrochemical salts, safeguards controls are required, but WIPP was not designed to accommodate safeguards controls. A potential solution would be to immobilize the salts. These immobilized salts would contain substantially higher plutonium concentrations than is currently permissible but would be suitable for disposal at WIPP. This document presents the results of a review of three immobilization technologies to determine if mature technologies exist that would be suitable to immobilize pyrochemical salts: cement-based stabilization, low-temperature vitrification, and polymer encapsulation. The authors recommend that flow sheets and life-cycle costs be developed for cement-based and low-temperature glass immobilization.« less
  • The presence of halide anions in four types of wastes arising from the pyrochemical reprocessing of plutonium required an immobilization process to be developed in which not only the actinide cations but also the halide anions were immobilized in a durable waste form. At AWE, we have developed such a process using Ca3(PO4)2 as the host material. Successful trials of the process with actinide- and Cl-bearing Type I waste were carried out at PNNL where the immobilization of the waste in a form resistant to aqueous leaching was confirmed. Normalized mass losses determined at 40°C and 28 days were 12more » x 10-6 g∙m-2 and 2.7 x 10-3 g∙m-2 for Pu and Cl, respectively. Accelerated radiation-induced damage effects are being determined with specimens containing 238Pu. No changes in the crystalline lattice have been detected with XRD after the 239Pu equivalent of 400 years ageing. Confirmation of the process for Type II waste (a oxyhydroxide-based waste) is currently underway at PNNL. Differences in the ionic state of Pu in the four types of waste have required different surrogates to be used. Samarium chloride was used successfully as a surrogate for both Pu(III) and Am(III) chlorides. Initial investigations into the use of HfO2 as the surrogate for Pu(IV) oxide in Type II waste indicated no significant differences.« less