scholarly journals Preliminary performance assessment for the Waste Isolation Pilot Plant, December 1992. Volume 4: Uncertainty and sensitivity analyses for 40 CFR 191, Subpart B

1993 ◽  
Author(s):  
Keyword(s):  
Performance Assessment ◽  
Pilot Plant ◽  
Sensitivity Analyses ◽  
Waste Isolation Pilot Plant ◽  
40 Cfr 191

Influence of Carbonate on Uranium Solubility in the WIPP

2012 ◽  
Vol 1444 ◽  
Author(s):  
Jean-Francois Lucchini ◽  
Sally Ballard ◽  
Hnin Khaing
Keyword(s):  
Performance Assessment ◽  
Pilot Plant ◽  
Solubility Limit ◽  
Experimental Results ◽  
Waste Isolation Pilot Plant ◽  
Long Term Experiments ◽  
Order Of Magnitude ◽  
Carbonate Complexation ◽  
Borate Complexation

ABSTRACTIn the performance assessment (PA) for the Waste Isolation Pilot Plant (WIPP), the solubility of uranium (VI) was conservatively set at 10-3 M for all expected WIPP conditions, including the potential and likely effects of carbonate complexation [1]. Under WIPP-relevant conditions, long-term experiments were performed to establish the uranium (VI) solubility limits in WIPP-simulated brine over a broad range of pCH+ values [7.5-12.5] and to evaluate the contribution of carbonate complexation and hydrolysis to uranium (VI) speciation. Data obtained in carbonate-free ERDA-6 brine, a simulated WIPP brine, were reported earlier [2]. In the absence of carbonate, uranium solubility approached 10-7 M at the expected pCH+ in the WIPP (~ 9.5). In the presence of a significant amount of carbonate (millimole levels), recent experimental results showed that uranium (VI) concentrations will not exceed 10-4M. This measured solubility limit is an order of magnitude lower than the uranium solubility value currently used in the WIPP PA [3]. A small effect of borate complexation was found in the pCH+ range [7.5-10]. At pCH+ ≥ 10, hydrolysis overwhelmed carbonate effects, and no amphoteric effect was observed.

scholarly journals The Advantages of a Salt/Bentonite Backfill for Waste Isolation Pilot Plant Disposal Rooms

1993 ◽  
Vol 333 ◽  
Author(s):  
B.M. Butcher
Keyword(s):  
Radioactive Waste ◽  
Pilot Plant ◽  
Selection Criteria ◽  
Final State ◽  
Government Regulations ◽  
Waste Isolation Pilot Plant ◽  
40 Cfr 191 ◽  
Backfill Materials ◽  
The Impact ◽  
Sorption Potential

ABSTRACTThis paper concludes that a 70 wt% salt/30 wt% bentonite mixture is preferable to pure crushed salt as backfill for disposal rooms in the Waste Isolation Pilot Plant. The performance of two backfill materials is examined with regard to various selection criteria related to compliance with the transuranic radioactive waste standard 40 CFR 191, Subpart B, such as the need for low liquid permeability after closure, chemical stability, strength, ease of emplacement, and sorption potential for brine and radionuclides. Both salt and salt/bentonite are expected to consolidate to a final state of permeability ≤ 10-18 m2, which is adequate for satisfying government regulations for nuclear repositories. The real advantage of the salt/bentonite backfill depends, therefore, on bentonite’s potential for sorbing brine and radionuclides. Estimates of the impact of these properties on backfill performance are presented.

Sign in / Sign up

Export Citation Format

Share Document