Wipp Waste Package Testing on Simulated DHLW: Emplacement

1984 ◽  
Vol 44 ◽  
Author(s):  
Martin A. Molecke
Keyword(s):  
Thermal Loading ◽  
Accelerated Aging ◽  
Waste Form ◽  
Future Application ◽  
High Level Waste ◽  
Glass Waste ◽  
Waste Package ◽  
High Level ◽  
Artificial Introduction ◽  
Materials Behavior

AbstractSeveral series of simulated (nonradioactive) defense high-level waste (DHLW) package tests have recently been emplaced in the WIPP, a research and development facility authorized to demonstrate the safe disposal of defense-related wastes. The primary purpose of these 3-to-7 year duration tests is to evaluate the in situ materials performance of waste package barriers (canisters, overpacks, backfills, and nonradioactive DHLW glass waste form) for possible future application to a licensed waste repository in salt. This paper describes all test materials, instrumentation, and emplacement and testing techniques, and discusses progress of the various tests.These tests are intended to provide information on materials behavior (i.e., corrosion, metallurgical and geochemical alterations, waste form durability, surface interactions, etc.), as well as comparison between several waste package designs, fabrications details, and actual costs.These experiments involve 18 full-size simulated DHLW packages (approximately 3.0 m x 0.6 m diameter) emplaced in vertical boreholes in the salt drift floor. Six of the test packages contain internal electrical heaters (470 W/canister), and were emplaced under approximately reference DHLW repository conditions. Twelve other simulated DHLW packages were emplaced tinder accelerated-aging or overtest conditions, including the artificial introduction of brine, and a thermal loading approximately three to four times higher than reference. Eight of these 12 test packages contain 1500 W/canister electrical heaters; the other four are filled with DHLW glass.

Break-up testing of waste-form materials

2012 ◽  
Vol 76 (8) ◽  
pp. 2975-2983 ◽  
Author(s):  
M. P. Metcalfe ◽  
W. Koch ◽  
G. Turner
Keyword(s):  
Waste Form ◽  
High Level Waste ◽  
Experimental Medicine ◽  
Nuclear Facilities ◽  
Waste Package ◽  
Impact Performance ◽  
Nuclear Decommissioning ◽  
Safety Case ◽  
Break Up ◽  
High Level

AbstractThe Nuclear Decommissioning Authority (NDA) is developing a safety case for the long-term management of higher activity wastes. This includes safety assessments of transport to and operations at the repository. One of the main faults and hazards to be considered is waste package response to impact accidents.The criteria of impact performance for waste packages are based upon activity release of particulates generated from the break up of the waste form during impact. The NDA approach to impact performance is based upon waste package response from finite element modelling in combination with break-up tests.Previous break up research commissioned by the NDA has concentrated on commercial graphite and glass samples. These extended studies, undertaken by the National Nuclear Laboratory in collaboration with the Department of Aerosol Technology of the Fraunhofer Institute of Toxicology and Experimental Medicine, provide break-up data specific to nuclear facilities and waste materials. These include archived unirrradiated graphite used to construct Magnox reactor cores and reflectors, simulant high level waste glass, selected grout formulations and selected metal-in-grout formulations.

scholarly journals High Level Waste (HLW) Vitrification Experience in the US: Application of Glass Product/Process Control to Other HLW and Hazardous Wastes

2008 ◽  
Vol 1107 ◽  
Author(s):  
Carol M. Jantzen ◽  
James C. Marra
Keyword(s):  
Process Control ◽  
High Level Waste ◽  
Glass Product ◽  
Mechanistic Models ◽  
Glass Waste ◽  
Statistical Process ◽  
Feed Composition ◽  
Feed Forward ◽  
Bonding Structure ◽  
High Level

AbstractVitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. At the Savannah River Site (SRS) actual HLW tank waste has successfully been processed to stringent product and process constraints without any rework into a stable borosilicate glass waste since 1996. A unique “feed forward” statistical process control (SPC) has been used rather than statistical quality control (SQC). In SPC, the feed composition to the melter is controlled prior to vitrification. In SQC, the glass product is sampled after it is vitrified. Individual glass property models form the basis for the “feed forward” SPC. The property models transform constraints on the melt and glass properties into constraints on the feed composition. The property models are mechanistic and depend on glass bonding/structure, thermodynamics, quasicrystalline melt species, and/or electron transfers. The mechanistic models have been validated over composition regions well outside of the regions for which they were developed because they are mechanistic. Mechanistic models allow accurate extension to radioactive and hazardous waste melts well outside the composition boundaries for which they were developed.

scholarly journals Spent Fuel as a Waste Form - Data Needs to Allow Long Term Performance Assessment Under Repository Disposal Conditions.

1986 ◽  
Vol 84 ◽  
Author(s):  
V. M. Oversby
Keyword(s):  
Performance Assessment ◽  
Waste Form ◽  
High Level Waste ◽  
Detailed Knowledge ◽  
Spent Fuel ◽  
Expected Performance ◽  
High Level ◽  
Term Performance ◽  
Physical And Chemical ◽  
Waste Repositories

AbstractPerformance assessment calculations are required for high level waste repositories for a period of 10,000 years under NRC and EPA regulations. In addition, the Siting Guidelines (IOCFR960) require a comparison of sites following site characterization and prior to final site selection to be made over a 100,000 year period. In order to perform the required calculations, a detailed knowledge of the physical and chemical processes that affect waste form performance will be needed for each site. While bounding calculations might be sufficient to show compliance with the requirements of IOCFR60 and 40CFRI91, the site comparison for 100,000 years will need to be based on expected performance under site specific conditions. The only case where detailed knowledge of waste form characteristics in the repository would not be needed would be where radionuclide travel times to the accessible environment can be shown to exceed 100,000 years. This paper will review the factors that affect the release of radionuclides from spemt fuel under repository conditions, summarize our present state of knowledge, and suggest areas where more work is needed in order to support the performance assessment calculations.

Materials Interactions Relating to Long-Term Geologic Disposal of Nuclear Waste Glass

1986 ◽  
Vol 84 ◽  
Author(s):  
Ned E. Bibler ◽  
Carol M. Jantzen
Keyword(s):  
Nuclear Waste ◽  
Review Paper ◽  
Waste Glass ◽  
Waste Form ◽  
Glass Waste ◽  
Waste Package ◽  
Geologic Disposal ◽  
Nuclear Waste Glass ◽  
Backfill Materials

AbstractIn the geologic disposal of nuclear waste glass, the glass will eventually interact with groundwater in the repository system. Interactions can also occur between the glass and other waste package materials that are present. These include the steel canister that holds the glass, the metal overpack over the canister, backfill materials that may be used, and the repository host rock. This review paper systematizes the additional interactions that materials in the waste package will impose on the borosilicate glass waste form-groundwater interactions. The repository geologies reviewed are tuff, salt, basalt, and granite. The interactions emphasized are those appropriate to conditions expected after repository closure, e.g. oxic vs. anoxic conditions. Whenever possible, the effect of radiation from the waste form on the interactions is examined. The interactions are evaluated based on their effect on the release and speciation of various elements including radionuclides from the glass. It is noted when further tests of repository interactions are needed before long-term predictions can be made.

Lessons Learned Siting and Successfully Processing U.S. DOE Radioactive Wastes Using a High Throughput Vitrification Process

2003 ◽  
Author(s):  
Robert E. Prince ◽  
Bradley W. Bowan
Keyword(s):  
The United States ◽  
Lessons Learned ◽  
Volume Reduction ◽  
Waste Form ◽  
High Level Waste ◽  
Hanford Site ◽  
High Level ◽  
Low Activity ◽  
The U.S ◽  
Ceramic Melter

This paper describes actual experience applying a technology to achieve volume reduction while producing a stable waste form for low and intermediate level liquid (L/ILW) wastes, and the L/ILW fraction produced from pre-processing of high level wastes. The chief process addressed will be vitrification. The joule-heated ceramic melter vitrification process has been used successfully on a number of waste streams produced by the U.S. Department of Energy (DOE). This paper will address lessons learned in achieving dramatic improvements in process throughput, based on actual pilot and full-scale waste processing experience. Since 1991, Duratek, Inc., and its long-term research partner, the Vitreous State Laboratory of The Catholic University of America, have worked to continuously improve joule heated ceramic melter vitrification technology in support of waste stabilization and disposition in the United States. From 1993 to 1998, under contact to the DOE, the team designed, built, and operated a joule-heated melter (the DuraMelterTM) to process liquid mixed (hazardous/low activity) waste material at the Savannah River Site (SRS) in South Carolina. This melter produced 1,000,000 kilograms of vitrified waste, achieving a volume reduction of approximately 70 percent and ultimately producing a waste form that the U.S. Environmental Protection Agency (EPA) delisted for its hazardous classification. The team built upon its SRS M Area experience to produce state-of-the-art melter technology that will be used at the DOE’s Hanford site in Richland, Washington. Since 1998, the DuraMelterTM has been the reference vitrification technology for processing both the high level waste (HLW) and low activity waste (LAW) fractions of liquid HLW waste from the U.S. DOE’s Hanford site. Process innovations have doubled the throughput and enhanced the ability to handle problem constituents in LAW. This paper provides lessons learned from the operation and testing of two facilities that provide the technology for a vitrification system that will be used in the stabilization of the low level fraction of Hanford’s high level tank wastes.

A Modest Proposal: A Robust, Cost-Effective Design for High-Level Waste Packages

1997 ◽  
Vol 506 ◽  
Author(s):  
M.J. Apted
Keyword(s):  
Mild Steel ◽  
Cost Effective ◽  
High Level Waste ◽  
Package Design ◽  
Waste Package ◽  
Modest Proposal ◽  
Proposal A ◽  
Cost Efficient ◽  
High Level ◽  
Effective Design

ABSTRACTAn alternative waste-package design for the geological disposal of high-level waste (HLW) glass is presented. In conventional designs, a massive buffer of compacted bentonite is placed around a thick-walled, mild-steel overpack; in the revised design, a much thinner buffer is placed within a thin-walled, mild-steel overpack. This simple expedient eliminates certain performance concerns in existing waste-package designs, while not necessitating the study of any new materials. This integrated waste package (IWP) design has comparable release-rate performance as current package designs for HLW. In addition, the 1WP design requires far-less rock excavation, permits significantly higher temperatures for longer periods, leads to a 20-50% reduction in repository area, and is more cost efficient than previous designs.

Modelling Radionuclide Speciation and Solubility Limits in The Near-Field of a Deep Repository

1985 ◽  
Vol 50 ◽  
Author(s):  
Hans Wanner
Keyword(s):  
Near Field ◽  
Cast Steel ◽  
Significant Rise ◽  
High Level Waste ◽  
Glass Waste ◽  
Carbonate Concentration ◽  
Realistic Description ◽  
Carbonate Complexation ◽  
Main Components ◽  
High Level

AbstractIn the safety analysis recently reported for a potential Swiss high-level waste repository, radionuclide speciation and solubility limits are calculated for expected granitic groundwater conditions. With the objective of deriving a more realistic description of radionuclide release from the near-field, an investigation has been initiated to quantitatively specify the chemistry of the near-field. In the Swiss case, the main components of the near-field are the glass waste-matrix, a thick cast steel canister horizontally stored in a drift, and a backfill of highly compacted bentonite.Based on available experimental data, an ion-exchange model for sodium, potassium, magnesium, and calcium has been developed, in order to simulate the reaction of sodium bentonite backfill with groundwater. The model assumes equilibrium with calcite as long as sufficient carbonates remain in the bentonite, as well as quartz saturation. The application of this model to the reference groundwater used in ‘Project Gewaehr 85’ results in a significant rise in pH (by up to 3 units) as well as a marked increase in the carbonate concentration.Neptunium and plutonium speciation and solubility limits are calculated for the reference groundwater chemistry gradually altered to that of saturated bentonite water and back again by a water exchange cycle model. The solubility limits estimated in this way generally turn out to be higher for the bentonite water than for the reference groundwater, mainly due to carbonate complexation of the actinide components AnO2+ and AnO22+. Uncertainties are particularly large for neptunium solubility due to its strong Eh dependence in bentonite water.

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