Estimating the Alteration Kinetics of the French Vitrified High-Level Waste Package in a Geologic Repository

1998 ◽  
Vol 123 (1) ◽  
pp. 67-81 ◽  
Author(s):  
Patrick Jollivet ◽  
Michèle Nicolas ◽  
Etienne Vernaz
Keyword(s):  
High Level Waste ◽  
Geologic Repository ◽  
Waste Package ◽  
High Level ◽  
Kinetics Of

High-Level Waste Package Licensing Considerations for Extrapolating test Data

1986 ◽  
Vol 84 ◽  
Author(s):  
T.C. Johnson ◽  
K.C. Chang ◽  
T.L. Jungling ◽  
L.S. Person ◽  
C.H. Peterson ◽  
...  
Keyword(s):  
Test Data ◽  
Nuclear Regulatory Commission ◽  
High Level Waste ◽  
Geologic Repository ◽  
Waste Package ◽  
Performance Predictions ◽  
Technical Issues ◽  
Regulatory Commission ◽  
High Level ◽  
General Standard

AbstractPrograms intended to provide supporting information for the high-level radioactive waste (HLW) repository program must consider the licensing requirements and the technical issues involved with extrapolation of short-term test data to periods of up to 10,000 years. The licensing requirements of the Nuclear Regulatory Commission (NRC), and the issues the NRC staff considers important for the development of predictive methods, are described. Because performance predictions of the geologic repository and particular components of the waste package must largely be based upon inference, a reasonable assurance, on the basis of the record before the Commission, is the general standard that will be required.

scholarly journals The initial dissolution rates of simulated UK Magnox–ThORP blend nuclear waste glass as a function of pH, temperature and waste loading

2015 ◽  
Vol 79 (6) ◽  
pp. 1529-1542 ◽  
Author(s):  
N. Cassingham ◽  
C.L. Corkhill ◽  
D.J. Backhouse ◽  
R.J. Hand ◽  
J.V. Ryan ◽  
...  
Keyword(s):  
Dissolution Kinetics ◽  
Intrinsic Rate ◽  
Waste Glass ◽  
High Level Waste ◽  
Dissolution Mechanism ◽  
Dissolution Rates ◽  
Forward Rates ◽  
Glass Dissolution ◽  
High Level ◽  
Kinetics Of

AbstractThe first comprehensive assessment of the dissolution kinetics of simulant Magnox–ThORP blended UK high-level waste glass, obtained by performing a range of single-pass flow-through experiments, is reported here. Inherent forward rates of glass dissolution were determined over a temperature range of 23 to 70°C and an alkaline pH range of 8.0 to 12.0. Linear regression techniques were applied to the TST kinetic rate law to obtain fundamental parameters necessary to model the dissolution kinetics of UK high-level waste glass (the activation energy (Ea), pH power law coefficient (η) and the intrinsic rate constant (k0)), which is of importance to the post-closure safety case for the geological disposal of vitreous products. The activation energies based on B release ranged from 55 ± 3 to 83 ± 9 kJ mol–1, indicating that Magnox–THORP blend glass dissolution has a surface-controlled mechanism, similar to that of other high-level waste simulant glass compositions such as the French SON68 and LAW in the US. Forward dissolution rates, based on Si, B and Na release, suggested that the dissolution mechanism under dilute conditions, and pH and temperature ranges of this study, was not sensitive to composition as defined by HLW-incorporation rate.

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.

The Effect of Fuel Burnup and Dispersed Water Intrusion on the Criticality of Spent High-Level Nuclear Fuel in a Geologic Repository

1993 ◽  
Vol 333 ◽  
Author(s):  
William G. Culbreth ◽  
Paige R. Zielinski
Keyword(s):  
Fuel Burnup ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Chain Reaction ◽  
Worst Case ◽  
Mixture Ratio ◽  
Waste Package ◽  
Radiological Data ◽  
A Chain ◽  
High Level

ABSTRACTStudies of the spent fuel waste package have been conducted through the use of a Monte-Carlo neutron simulation program to determine the ability of the fuel to sustain a chain reaction. These studies have included fuel burnup and the effect of water mists on criticality. Results were compared with previous studies.In many criticality studies of spent fuel waste packages, fresh fuel with an enrichment as high as 4.5% is used as the conservative (worst) case. The actual spent fuel has a certain amount of “burnup” that decreases the concentration of fissile uranium and increases the amount of radionuclides present. The LWR Radiological Data Base from OCRWM has been used to determine the relative radionuclide ratios and KENO 5.a was used to calculate values of the effective multiplication factor, keff.1Spent fuel is not capable of sustaining a chain reaction unless a suitable moderator, such as water, is present. A completely flooded container has been treated as the worst case for criticality. Results of a previous report that demonstrated that keff actually peaked at a water-to-mixture ratio of 13% were analyzed for validity. In the present study, these results did not occur in the SCP waste package container.

scholarly journals Sorption-Capacity Limited Retardation of Radionuclides Transport in Water-Saturated Packing Materials

1984 ◽  
Vol 44 ◽  
Author(s):  
C. Pescatore ◽  
T. Sullivan
Keyword(s):  
Sorption Capacity ◽  
Retardation Factor ◽  
High Level Waste ◽  
Radionuclide Transport ◽  
Breakthrough Time ◽  
Packing Materials ◽  
Sorption Data ◽  
Waste Package ◽  
High Level ◽  
Water Saturated

AbstractRadionuclides breakthrough times as calculated through constant retardation factors obtained in dilute solutions are non-conservative. The constant retardation approach regards the solid as having infinite sorption capacity throughout the solid. However, as the solid becomes locally saturated, such as in the proximity of the waste form-packing materials interface, it will exhibit no retardation properties, and transport will take place as if the radionuclides were locally non-reactive. The magnitude of the effect of finite sorption capacity of the packing materials on radionuclide transport is discussed with reference to high-level waste package performance. An example based on literature sorption data indicates that the breakthrough time may be overpredicted by orders of magnitude using a constant retardation factor as compared to using the entire sorption isotherm to obtain a concentrationdependent retardation factor.

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.

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