scholarly journals Initial specifications for nuclear waste package external dimensions and materials

10.2172/59284 ◽  
1983 ◽  
Author(s):  
D.W. Gregg ◽  
W.C. O`Neal
Keyword(s):  
Nuclear Waste ◽  
Waste Package

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.

Design by Analysis of Waste Packages at Yucca Mountain for Impact Loads

2009 ◽  
Author(s):  
Randy J. James ◽  
Kenneth Jaquay ◽  
Michael J. Anderson
Keyword(s):  
Nuclear Waste ◽  
Impact Loading ◽  
Structural Integrity ◽  
Structural Response ◽  
Yucca Mountain ◽  
Impact Loads ◽  
Geologic Repository ◽  
Dynamic Simulations ◽  
Nuclear Waste Storage ◽  
Waste Package

The proposed geologic repository under development at Yucca Mountain, Nevada, will employ multiple shell metallic containers (waste packages) for the disposal of nuclear waste. The waste packages represent a primary engineered barrier for protection and containment of the radioactive waste, and the design of these containers must consider a variety of structural conditions to insure structural integrity. Some of the more challenging conditions for structural integrity involve severe impact loading due to hypothesized event sequences, such as drops or collisions during transport and placement. Due to interactions between the various components leading to complex structural response during an impact sequence, nonlinear explicit dynamic simulations and highly refined models are employed to qualify the design for these severe impact loads. This paper summarizes the Design by Analysis methodologies employed for qualification of waste package design under impact loading and provides several illustrative examples using these methods. Example evaluations include a collision of a waste package by the Transport and Emplacement Vehicle (TEV) and two scenarios due to seismic events, including WP impact within the TEV and impact by falling rock. The examples are intended to illustrate the stringent Design by Analysis methods employed and also highlight the scope of structural conditions included in the design basis for waste packages to be used for proposed nuclear waste storage at Yucca Mountain.

Impact of Phase Stability on the Corrosion Behavior of the Austenitic Candidate Materials For NNSWI.

1987 ◽  
Vol 112 ◽  
Author(s):  
Daniel B. Bullen ◽  
Gregory E. Gdowski ◽  
R. Daniel McCright
Keyword(s):  
Corrosion Behavior ◽  
Nuclear Waste ◽  
Phase Stability ◽  
Lawrence Livermore National Laboratory ◽  
National Laboratory ◽  
Intergranular Stress Corrosion ◽  
Technical Literature ◽  
Nuclear Waste Storage ◽  
Waste Package ◽  
The Impact

AbstractThe Nuclear Waste Management Program at Lawrence Livermore National Laboratory is responsible for the development of the waste package design to meet the Nuclear Regulatory Commission licensing requirements for the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. The metallic container component of the waste package is required to assist in providing substantially complete containment of the waste for a period of up to 1000 years. Long term phase stability of the austenitic candidate materials (304L and 316L stainless steels and alloy 825) over this time period at moderate temperatures (100–250°C) can impact the mechanical and corrosion behavior of the metal barrier.A review of the technical literature with respect to phase stability of 304L, 316L and 825 is presented. The impact of martensitic transformations, carbide precipitation and intermediate (σ. χ, and η) phase formation on the mechanical properties and corrosion behavior of these alloys at repository relevant conditions is discussed. The effect of sensitization on intergranular stress corrosion cracking (IGSCC) of each alloy is also addressed. A summary of the impact of phase stability on the degradation of each alloy in the proposed repository environment is included.

237Np aND 239Pu Solution Behavior During Hydrothermal Testing of Simulated Nuclear Waste Glass with Basalt and Steel

1984 ◽  
Vol 44 ◽  
Author(s):  
Janet A. Schramke ◽  
Scott A. Simonson ◽  
David G. Coles
Keyword(s):  
Nuclear Waste ◽  
Cast Steel ◽  
Size Fraction ◽  
Weight Ratio ◽  
Reaction Products ◽  
Solid Materials ◽  
Solution Behavior ◽  
Waste Package ◽  
Hydrothermal Experiments ◽  
Nuclear Waste Glass

AbstractA series of hydrothermal experiments were carried out on 237Np- and 239Pu-doped PNL 76–68 glass, synthetic basalt groundwater, basalt, and cast steel. These hydrothermal experiments are part of the Basalt Waste Isolation Project investigation of the interactions of waste package components in a basalt repository. Experiments of three months duration were conducted in Dickson-type rocking autoclaves at 200°C and 30 MPa, with an initial fluid to solid weight ratio of 10:1. All solid materials were ground and sieved to a narrow size fraction. The experiments carried out were: glass and groundwater; glass, basalt, and groundwater; glass, steel, and groundwater; and glass, steel, basalt, and groundwater. Unfiltered, 4000 Â filtered, and 18 Å filtered solutions were analyzed to determine the concentrations of radionuclides in solution and those associated with colloids.The quantities of 237Np and 239Pu in solution were very small. Worstcase calculations indicate that 0.01% or less of the total radionuclide inventories were present in solution. The highest solution concentrations of the actinide dopants were observed in the experiments with basalt, even though smaller amounts of glass were dissolved than in the experiments without basalt. The observed differences in the solution concentrations of 237Np and 239Pu were probably controlled by differences in the reaction products, which were clinoptilolite in the experiments with basalt, and an Fe-Zn smectite clay in the experiments without basalt.

scholarly journals Preclosure analysis of conceptual waste package designs for a nuclear waste repository in tuff

10.2172/59344 ◽  
1984 ◽  
Author(s):  
W.C. O`Neal ◽  
D.W. Gregg ◽  
J.N. Hockman ◽  
E.W. Russell ◽  
W. Stein
Keyword(s):  
Nuclear Waste ◽  
Nuclear Waste Repository ◽  
Waste Package ◽  
Waste Repository

Effect of the Residual Heat Release of the Nuclear Waste Stored in an Unsaturated Zone on Radionuclide Release

2010 ◽  
Author(s):  
Lubna K. Hamdan ◽  
John C. Walton ◽  
Arturo Woocay
Keyword(s):  
Heat Release ◽  
Nuclear Waste ◽  
Unsaturated Zone ◽  
Yucca Mountain ◽  
Minimum Length ◽  
General Corrosion ◽  
Waste Package ◽  
Flow Through ◽  
Radionuclide Release ◽  
Residual Heat

Over time, nuclear waste packages disposed in geological repositories are expected to fail gradually due to localized and general corrosion. As a result, water will have access to the nuclear waste and radionuclides will be transported to the accessible environment by ground water. In this paper we consider a serious failure case in which penetrations at the top and bottom of the waste package will allow water to flow through it (flow-through model). We introduce a new conceptual model that examines the effect of the residual heat release of the nuclear waste stored in an unsaturated environment on radionuclide release. This model predicts that the evaporation of water at the hotter sheltered areas (from condensate and seepage) inside the failed waste package will create a capillary pressure gradient that drives water to wick with its dissolved and suspended contents toward these relict areas, effectively preventing radionuclides release. We drive a dimensionless group to estimate the minimum length of the sheltered areas required to sequester radionuclides and prevent their release. The implications of this model on the performance of the proposed repository at Yucca Mountain or unsaturated zone geological repositories in general are explored.

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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