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.

Thermal Conductivity of Bentonite/Quartz High-Level Waste Package Backfill

1982 ◽  
Vol 15 ◽  
Author(s):  
Marvin Moss ◽  
Martin A. Molecke
Keyword(s):  
Thermal Conductivity ◽  
Geometric Mean ◽  
Bentonite Clay ◽  
Weight Percent ◽  
High Level Waste ◽  
Package Design ◽  
Waste Package ◽  
Heating And Cooling ◽  
High Level ◽  
The Impact

ABSTRACTA mixture of bentonite clay and quartz sand is being considered for use as a waste-package backfill, the material placed between a radioactive-waste canister and the repository host rock. Compacts of bentonite/quartz with weight-percent ratios of 100/0, 70/30, 50/50 and 30/70 were made at room temperature under a pressure of 100 MPa (15 ksi). Upon initial heating, the thermal conductivity of the 70/30 compact rose from 1.20 W/m·K at 298 K to 1.32 W/m·K at 373 K. After further heating to 473 K, it fell to 1.10 W/m·K, reflecting the loss of interlamellar water from the bentonite. The conductivity of the now-dehydrated compact was reproducible through several heating and cooling cycles, varying from 1.15 W/m·K at 573 K to 1.03 W/m·K at 298 K. The other mixtures were qualitatively similar to the 70/30: the 100/0, 50/50 and 30/70 dehydrated compacts displayed conductivities of 0.59, 1.06 and 0.83 W/m.K,respectively, at 298 K. Measured densities ranged from 1.98 to 2.12 g/cc. Combined geometric-mean and Maxwell models for thermal conduction in composite systems predict the measured results reasonably well. An analysis of the impact of backfills on high-level waste (HLW) package design indicates that no significant thermal penalty is imposed.

Waste Package Reliability / Nalysis

1983 ◽  
Vol 26 ◽  
Author(s):  
C. Pescatore ◽  
C. Sastre
Keyword(s):  
Ordinary Sense ◽  
Regulatory Requirements ◽  
Package Design ◽  
Waste Package ◽  
Qualitative And Quantitative ◽  
Future Performance ◽  
High Level Nuclear Waste ◽  
High Level ◽  
Package Reliability ◽  
General Method

ABSTRACTProof of future performance of a complex system such as a high-level nuclear waste package over a period of hundreds to thousands of years cannot be had in the ordinary sense of the word. The general method of probabilistic reliability analysis could provide an acceptable framework to identify, organize, and convey the information necessary to satisfy the criterion of reasonable assurance of waste package performance according to the regulatory requirements set forth in 10 CFR 60. General principles which may be used to evaluate the qualitative and quantitative reliability of a waste package design are indicated and illustrated with a sample calculation of a repository concept in basalt.

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.

Radionuclide Uptake and Transport on Microbes in Potential Repository Drifts at Yucca Mountain, Nevada

2002 ◽  
Vol 713 ◽  
Author(s):  
Darren M. Jolley
Keyword(s):  
Source Term ◽  
Spent Nuclear Fuel ◽  
High Level Waste ◽  
Multiple Model ◽  
Data Set ◽  
Model Calculations ◽  
Waste Package ◽  
Uranium Uptake ◽  
High Level ◽  
Dry Cell

ABSTRACTRadionuclide adsorption onto microbes, microbial retention in the engineered barrier system (EBS), and their potential release from the EBS as microbial colloids have been investigated. The microbial source term for these calculations was derived using MING V 1.0 software code [1]. Multiple model calculations from MING representing variations in possible microbial communities in the EBS were abstracted into two equations representing one meter segments of potential repository drift containing either commercial spent nuclear fuel (CSNF) or defense high level waste (HLW) packages. These two equations (Equations 1 and 2) represent the average cumulative microbial biomass generated in the EBS at any given time. A distribution for uranium uptake onto microbes (162.88 ± 133.05 mg U/gm dry cell) was applied to the microbial source term. The distribution was derived from the data set in Suzuki and Banfield [2] representing 45 different species of bacteria and fungus, covering uranium uptake at optimum pH values of 1 to 7. The mass of uranium sorbed onto the biomass was either sequestered in the EBS or transported as a microbial colloid based on a regression of data from Jewett et al. [3] representing microbial sorption onto air-water interfaces in unsaturated column experiments. Over one million years, it is estimated that EBS microbes may adsorb from 77 to 2302 kg of uranium [2302 kg U > 100% of the uranium available in a one meter segment of a CSNF waste package] per meter of waste package depending on the saturation of the invert and type of waste package. Over the same time, microbial colloids may transport from 8 to 1250 kg of adsorbed uranium per meter of waste package from the EBS.

Evolution of Chemistry and Its Effects on the Corrosion of Engineered Barrier Materials

2006 ◽  
Vol 985 ◽  
Author(s):  
Darrell Dunn ◽  
Yi-Ming Pan ◽  
Xihua He ◽  
Lietai Yang ◽  
Roberto Pabalan
Keyword(s):  
Environmental Conditions ◽  
Environmental Influence ◽  
Near Field ◽  
Yucca Mountain ◽  
High Level Waste ◽  
Seepage Water ◽  
Waste Package ◽  
Barrier Materials ◽  
Engineered Barrier ◽  
High Level

ABSTRACTThe evolution of environmental conditions within the emplacement drifts of a potential high-level waste repository at Yucca Mountain, Nevada, may be influenced by several factors, including the temperature and relative humidity within the emplacement drifts and the composition of seepage water. The performance of the waste package and the drip shield may be affected by the evolution of the environmental conditions within the emplacement drifts. In this study, tests evaluated the evolution of environmental conditions on the waste package surfaces and in the surrounding host rock. The tests were designed to (i) simulate the conditions expected within the emplacement drifts; (ii) measure the changes in near-field chemistry; and (iii) determine environmental influence on the performance of the engineered barrier materials. Results of tests conducted in this study indicate the composition of salt deposits was consistent with the initial dilute water chemistry. Salts and possibly concentrated calcium chloride brines may be more aggressive than either neutral or alkaline brines.

Evaluation of the Susceptibility of Astm A216 Grade Wca Mild Steel to Stress Corrosion Cracking in Simulated Salt Repositopy Environments

1986 ◽  
Vol 84 ◽  
Author(s):  
S. G. Pitman
Keyword(s):  
Mild Steel ◽  
Stress Corrosion ◽  
Nuclear Waste ◽  
Dynamic Tests ◽  
Static Tests ◽  
Waste Package ◽  
Test Conditions ◽  
High Level Nuclear Waste ◽  
High Level ◽  
Container Material

AbstractIn current conceptual designs, a mild steel (ASTM A?16 Grade WCA) is the relerence container material for use in high level nuclear waste packages intended for emplacement in a salt repository. The resistance of the steel to stress corrosion crackinq (SCC) is being investigated as part of the effort underway to verify the suitability of the material for waste package applications. Static tests (U-bend and bolt-loaded fracture toughness specimens) and dynamic tests (slow strain rate and corrosion fatigue) were conducted on both as-cast and weldment specimens of the material, in both low-Mg and high-Mg halite-saturated brines, in the temperature range of 90 to 200°C. The investigations indicate that the steel is not susceptible to SCC under the test conditions employed.

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