An introduction to the mechanics of performance assessment using examples of calculations done for the Waste Isolation Pilot Plant between 1990 and 1992. Revision

This document provides an overview of the processes used to access the performance of the Waste Isolation Pilot Plant (WIPP). The quantitative metrics used in the performance-assessment (PA) process are those put forward in the Environmental Protection Agency`s Environmental Standards for the Management and Disposal of Spent Nuclear Fuel, HIgh-LEvel and transuranic radioactive Wastes (40 CFR 191)

Historical Background on Assessment the Performance of the Waste Isolation Pilot Plant

In 1979, six years after selecting the Delaware Basin as a potential disposal area, Congress authorized the US Department of Energy to build the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico, as a research and development facility for the safe management, storage, and disposal of waste contaminated with transuranic radioisotopes. In 1998, 19 years after authorization and 25 years after site selection, the US Environmental Protection Agency (EPA) certified that the WIPP disposal system complied with its regulations. The EPA's decision was primarily based on the results from a performance assessment conducted in 1996. This performance assessment was the culmination of four preliminary performance assessments conducted between 1989 and 1992. This report provides a historical setting and context for how the performance of the deep geologic repository at the WIPP was analyzed. Also included is background on political forces acting on the project. For example, the federal requirement to provide environmental impact statements and negotiated agreements with the State of New Mexico influenced the type of scientific areas that were investigated and the engineering analysis prior to 1989 for the WIPP

Milestones for disposal of radioactive waste at the Waste Isolation Pilot Plant (WIPP) in the United States

Since its identification as a potential deep geologic repository in about 1973, the regulatory assessment process for the Waste Isolation Pilot Plant (WIPP) in New Mexico has developed over the past 25 years. National policy issues, negotiated agreements, and court settlements over the first half of the project had a strong influence on the amount and type of scientific data collected. Assessments and studies before the mid 1980s were undertaken primarily (1) to satisfy needs for environmental impact statements, (2) to develop general understanding of selected natural phenomena associated with nuclear waste disposal, or (3) to satisfy negotiated agreements with the State of New Mexico. In the last third of the project, federal compliance policy and actual regulations were sketched out, but continued to evolve until 1996. During this eight-year period, four preliminary performance assessments, one compliance performance assessment, and one verification performance assessment were performed

Performance assessment of the direct disposal in unsaturated tuff of spent nuclear fuel and high-level waste owned by U.S. Department of Energy. Volume 1: Executive summary

This assessment studied the performance of high-level radioactive waste and spent nuclear fuel in a hypothetical repository in unsaturated tuff. The results of this 10-month study are intended to help guide the Office of Environment Management of the US Department of Energy (DOE) on how to prepare its wastes for eventual permanent disposal. The waste forms comprised spent fuel and high-level waste currently stored at the Idaho National Engineering Laboratory (INEL) and the Hanford reservation. About 700 metric tons heavy metal (MTHM) of the waste under study is stored at INEL, including graphite spent nuclear fuel, highly enriched uranium spent fuel, low enriched uranium spent fuel, and calcined high-level waste. About 2,100 MTHM of weapons production fuel, currently stored on the Hanford reservation, was also included. The behavior of the waste was analyzed by waste form and also as a group of waste forms in the hypothetical tuff repository. When the waste forms were studied together, the repository was assumed also to contain about 9,200 MTHM high-level waste in borosilicate glass from three DOE sites. The addition of the borosilicate glass, which has already been proposed as a final waste form, brought the total to about 12,000 MTHM. A source term model was developed to study the wide variety of waste forms, which included radionuclides residing in 10 different matrices and up to 8 nested layers of material that might react with water. The possibility and consequences of critical conditions occurring in or near containers of highly enriched uranium spent nuclear fuel were also studied

An introduction to the mechanics of performance assessment using examples of calculations done for the Waste Isolation Pilot Plant between 1990 and 1992

This document provides an overview of the process used to assess the performance of the Waste Isolation Pilot Plant (WIPP), a proposed repository for transuranic wastes that is located in southeastern New Mexico. The quantitative metrics used in the performance-assessment (PA) process are those put forward in the Environmental Protection Agency`s Environmental Standards for the Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive flasks (40 CFR 191). Much has been written about the individual building blocks that comprise the foundation of PA theory and practice, and that WIPP literature is well cited herein. However, the present approach is to provide an accurate, well documented overview of the process, from the perspective of the mechanical steps used to perform the actual PA calculations. Specifically, the preliminary stochastic simulations that comprise the WIPP PAs of 1990, 1991. and 1992 are summarized

Euler buckling of geothermal well casing

Geothermal well operators have expressed concern over the vulnerability of unsupported casing to buckling from thermal elongation. Preliminary numerical and theoretical calculations are presented, which indicate the buckling phenomenon should not be serious in N-80 casing if the string is tension preloaded. Buckling would be detrimental for K-55 casing. The effect of wall contact was found to be beneficial for closely confined pipe strings and of no detriment when hole gaps are large. The weakness of API screw joints in bending appears to be the structural limitation. The analysis assumed stresses above yield constituted failure, that thermal expansion was strain controlled, and that the casing was continuous. Excessive internal pressure instability was ignored. The temperature variation considered was between cementing conditions of 100 to 200/sup 0/F (40 to 95/sup 0/C) and shut-in conditions of 425 to 450/sup 0/F (220 to 230/sup 0/C)

CLADDING DEGRADATION COMPONENT IN WASTE FORM DEGRADATION MODEL IN TSPA-SR

The U.S. Department of Energy (DOE) has prepared a total system performance assessment for a site recommendation (TSPA-SR), if suitable, on Yucca Mountain for disposal of radioactive waste. Discussed here is the Cladding Degradation Component of the Waste Form Degradation Model (WF Model), of the TSPA-SR. The Cladding Degradation Component determines the degradation rate of the Zircaloy cladding on commercial spent nuclear fuel (CSNF) and, thereby, the CSNF matrix exposed and radioisotopes available for dissolution in any water present. Since the 1950s, most CSNF has been clad with less than 1 mm (usually between 600 and 900 {micro}m) of Zircaloy, a zirconium alloy. Zircaloy cladding is not a designed engineered barrier of the Yucca Mountain disposal system, but rather is an existing characteristic of the CSNF that is important to determining the release rate of radioisotopes once the waste package (WP) has breached. Although studies of cladding degradation from fluoride [F] began at Lawrence Livermore National Laboratory as early as 1984, cladding as a characteristic of the waste was not considered in TSPAs, conducted in the early 1990s. However, enough information on cladding performance has accumulated in the literature such that cladding was considered in 1993 when examining the performance of DOE spent nuclear (DSNF) and most recently in TSPA for the viability assessment (TSPA-VA). The Nuclear Regulatory Commission (NRC) currently uses cladding data as the basis for extending the period of wet storage, for licensing dry storage facilities, and for licensing shipping casks for CNSF

Permeability change near instrumentation holes in jointed rock: implications for the tuff radionuclide-migration field experiment

In order to assess in situ joint permeability near waste repositories, it has been proposed that instrumentation holes with axes parallel to the joint plane be drilled. However, the drill holes after the normal stress across the joint. The resultant stress concentration decreases the joint aperture and can significantly affect the joint permeability. Different intersections of the hold axis relative to the joint plane were examined utilizing a plane-strain, elastic analysis. It was found that a tangential joint intersection minimized the normal stress change. Stress along the joint increased by 10 to 15 percent and the permeability-aperture product decreased to 65 to 70 percent of its original flow

Permeability change near instrumentation holes in jointed rock: implications for the tuff radionuclide-migration field experiment

In order to assess in situ joint permeability near waste repositories, it has been proposed that instrumentation holes with axes parallel to the joint plane be drilled. However, the drill holes after the normal stress across the joint. The resultant stress concentration decreases the joint aperture and can significantly affect the joint permeability. Different intersections of the hold axis relative to the joint plane were examined utilizing a plane-strain, elastic analysis. It was found that a tangential joint intersection minimized the normal stress change. Stress along the joint increased by 10 to 15 percent and the permeability-aperture product decreased to 65 to 70 percent of its original flow