MacroscoMacroscopic Cracking Determination in LaBS Glasspic Cracking Determination in LaBS Glass
The DOE/EM plans to conduct the Plutonium Vitrification Project at the Savannah River Site (SRS). An important part of this project is to reduce the attractiveness of the plutonium by fabricating a plutonium glass form and immobilizing the Pu form within the high level waste (HLW) glass prepared in the Defense Waste Processing Facility (DWPF). This requires that a project schedule that is consistent with EM plans for DWPF and cleanup of the SRS be developed. Critical inputs to key decisions in the vitrification project schedule are near-term data that will increase confidence that the lanthanide borosilicate (LaBS) glass product is suitable for disposal in the Yucca Mountain Repository. A workshop was held on April 28, 2005 at Bechtel SAIC Company (BSC) facility in Las Vegas, NV to define the near term data needs. Dissolution rate data and the fate of plutonium oxide and the neutron absorbers during the dissolution process were defined as key data needs. A suite of short-term tests were defined at the workshop to obtain the needed data. The objectives of these short-term tests are to obtain data that can be used to show that the dissolution rate of a LaBS glass is acceptable and to show that the extent of Pu separation from neutron absorbers, as the glass degrades and dissolves, is not likely to lead to criticality concerns. An additional data need was identified regarding the degree of macroscopic cracking and/or voiding that occurs during processing of the Pu glass waste form and subsequent pouring of HLW glass in the DWPF. A final need to evaluate new frit formulations that may increase the durability of the plutonium glass and/or decrease the degree to which neutron absorbers separate from the plutonium during dissolution was identified. This task plan covers the need to evaluate the degree of macroscopic cracking and/or voiding that occurs during processing of the Vitrified Plutonium Waste Form (i.e. the can-in-canister configuration containing the vitrified Pu product). Separate task plans were developed for Pu glass performance testing of the current baseline LaBS glass composition and development of alternative frit formulations. Recent results from Pressurized Unsaturated Flow (PUF) testing showed the potential separation of Pu from Gd during the glass dissolution process [3]. Post-test analysis of the LaBS glass from a 6-year PUF test showed a region where Pu had apparently accumulated in a Pu-bearing disk-like phase that had become separated from neutron absorber (Gd). It should be noted that this testing was conducted on the early LaBS Frit A glass composition that was devoid of HfO{sub 2} as a neutron absorber. PUF testing is currently being initiated using the LaBS Frit B composition that contains HfO{sub 2}. The potential for fissile material and neutron absorber separation is a criticality risk for the repository. The surface area that is available for leaching (i.e. due to the degree of cracking or voiding within the Pu glass cylinder) is a factor in modeling the amount of fissile material and neutron absorber released during the dissolution process. A mathematical expression for surface area is used in the Total Systems Performance Assessment (TSPA) performed by BSC personnel. Specifically, the surface area available for leaching is being used in current external criticality assessments. The planned processing steps for producing a VPWF assembly involves processing Pu feed and LaBS frit to produce a can of Pu LaBS glass, packaging this can into a second can (i.e. bagless transfer) for removal from the glovebox processing environment, placing a series of bagless transfer cans into a DWPF canister, and pouring HLW glass into the DWPF canister to encapsulate bagless transfer cans. The objective of this task is to quantify the degree of cracking and/or voiding that will occur during the processing of the VPWF.
- Research Organization:
- Savannah River Site (SRS), Aiken, SC
- Sponsoring Organization:
- USDOE
- DOE Contract Number:
- DE-AC09-96SR18500
- OSTI ID:
- 881503
- Report Number(s):
- WSRC-RP-2005-01702; TRN: US0603152
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
BOROSILICATE GLASS
CONTAINERS
FISSILE MATERIALS
NEUTRON ABSORBERS
PERFORMANCE TESTING
PLUTONIUM OXIDES
RARE EARTHS
SAVANNAH RIVER PLANT
SURFACE AREA
WASTE FORMS
WASTE PROCESSING
YUCCA MOUNTAIN