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Title: Conceptual waste package interim product specifications and data requirements for disposal of borosilicate glass defense high-level waste forms in salt geologic repositories

Abstract

The conceptual waste package interim product specifications and data requirements presented are applicable specifically to the normal borosilicate glass product of the Defense Waste Processing Facility (DWPF). They provide preliminary numerical values for the defense high-level waste form parameters and properties identified in the waste form performance specification for geologic isolation in salt repositories. Subject areas treated include containment and isolation, operational period safety, criticality control, waste form/production canister identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available.

Publication Date:
Research Org.:
Battelle Memorial Inst., Columbus, OH (USA). Office of Nuclear Waste Isolation
OSTI Identifier:
6025688
Report Number(s):
ONWI-464
ON: DE83013902
DOE Contract Number:
AC06-76RL01830; AC02-83CH10140
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES; 11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; BOROSILICATE GLASS; SPECIFICATIONS; PACKAGING; RADIOACTIVE WASTE DISPOSAL; HIGH-LEVEL RADIOACTIVE WASTES; SALT DEPOSITS; CONTAINERS; PERFORMANCE; PERFORMANCE TESTING; GEOLOGIC DEPOSITS; GLASS; MANAGEMENT; MATERIALS; RADIOACTIVE MATERIALS; RADIOACTIVE WASTES; TESTING; WASTE DISPOSAL; WASTE MANAGEMENT; WASTES; 052002* - Nuclear Fuels- Waste Disposal & Storage; 054000 - Nuclear Fuels- Health & Safety

Citation Formats

Not Available. Conceptual waste package interim product specifications and data requirements for disposal of borosilicate glass defense high-level waste forms in salt geologic repositories. United States: N. p., 1983. Web. doi:10.2172/6025688.
Not Available. Conceptual waste package interim product specifications and data requirements for disposal of borosilicate glass defense high-level waste forms in salt geologic repositories. United States. doi:10.2172/6025688.
Not Available. Wed . "Conceptual waste package interim product specifications and data requirements for disposal of borosilicate glass defense high-level waste forms in salt geologic repositories". United States. doi:10.2172/6025688. https://www.osti.gov/servlets/purl/6025688.
@article{osti_6025688,
title = {Conceptual waste package interim product specifications and data requirements for disposal of borosilicate glass defense high-level waste forms in salt geologic repositories},
author = {Not Available},
abstractNote = {The conceptual waste package interim product specifications and data requirements presented are applicable specifically to the normal borosilicate glass product of the Defense Waste Processing Facility (DWPF). They provide preliminary numerical values for the defense high-level waste form parameters and properties identified in the waste form performance specification for geologic isolation in salt repositories. Subject areas treated include containment and isolation, operational period safety, criticality control, waste form/production canister identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available.},
doi = {10.2172/6025688},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Wed Jun 01 00:00:00 EDT 1983},
month = {Wed Jun 01 00:00:00 EDT 1983}
}

Technical Report:

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  • The conceptual waste package interim product specifications and data requirements presented are applicable to the reference glass composition described in PNL-3838 and carbon steel canister described in ONWI-438. They provide preliminary numerical values for the commercial high-level waste form parameters and properties identified in the waste form performance specification for geologic isolation in salt repositories. Subject areas treated include containment and isolation, operational period safety, criticality control, waste form/production canister identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data,more » analyses and regulatory requirements become available. 13 references, 1 figure.« less
  • The interim performance specifications and data requirements presented are applicable to all types of radioactive wastes prepared for disposal in salt geologic repositories. In this document, an interim numerical value is provided for only one waste-form parameter - the waste-form release rate. This is the key waste-form parameter affecting waste isolation, and is defined here for spent fuel, defense high-level waste (HLW), and commercial HLW. The waste-form performance specifications and data requirements respond to the waste package performance criteria. Subject areas treated include: containment and controlled release, operational period safety, criticality control, identification, and waste-package performance testing requirements. This documentmore » was generated for use in the development of conceptual waste-package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available.« less
  • The interim performance specifications and data requirements presented apply to conceptual waste package designs for all waste forms which will be isolated in salt geologic repositories. The waste package performance specifications and data requirements respond to the waste package performance criteria. Subject areas treated include: containment and controlled release, operational period safety, criticality control, identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available.
  • This report provides conceptual waste package designs for use by the Office of Nuclear Waste Isolation (ONWI) in preparing a repository conceptual design in salt. Included are designs for the current reference waste form configurations of Defense High Level Waste, which consists of Savannah River Laboratory wastes immobilized in borosilicate glass, Commercial High Level Waste, which is a borosilicate glass waste form that results from the immobilization of commercial spent fuel reprocessing wastes, and Spent Fuel - Form 2, which consists of consolidated spent fuel rods from PWR or BWR assemblies. Reference designs are presented which are used as amore » baseline to evaluate design alternatives resulting from variations in the waste form configuration, the design approach, and the design data base. This broad spectrum of conceptual designs for salt has been included to provide ONWI a basis for concept comparison, an indication as to which of several package and repository design approaches are more cost effective, and guidance as to which approaches should be pursued in future design efforts. Based on available data and the analyses performed, all the concepts in this report offer technically viable approaches to the containment of the waste form for at least 1000 years and for adequate isolation of radionuclides thereafter. The reference borehole-type design is one which provides containment through the use of a titanium alloy corrosion-resistant overpack, backed with a carbon steel structural reinforcing member. An alternate borehole design is one which provides containment with an all-steel overpack which is sufficiently thick to withstand expected crushing loads and to accommodate expected corrosion. The self-shielded package approach provides containment through use of a thick section of moderately corrosion resistant ferrous alloy material. Development programs are identified that will be required to support designs during licensing.« less
  • In the case of repositories in salt, adsorption/desorption reactions with salt are expected to be minimal. If solubility-controlling solids of radionuclides are present in the waste or can form in the engineered barrier system, the upper concentration limits of radionuclides that can be leached from the wastes will be solubility-limited but independent of the release scenarios, hydrologic transport characteristics, and adsorption/desorption reactions. The available thermochemical data show that most of the radioactive elements, such as actinides, that are of concern over long repository storage times form solubility-controlling solids. Therefore, it should be possible to set upper limits on the concentrationsmore » of these radioactive elements that can be leached from wastes disposed of in salt repositories. To set upper limits, data are needed for solid phases that form readily, have low solubilities, and either are present in wastes (spent fuel, waste glasses, etc.) or can form readily in the geologic environment. Factors that would lower the maximum concentrations in leachates include an increase in the crystallinity of amorphous precipitates, the presence of crystalline solids in the wastes, and the formation of solid solutions of the actinides. 11 refs.« less