Rob P. Rechard

Historical Relationship Between Performance Assessment for Radioactive Waste Disposal and Other Types of Risk Assessment

This article describes the evolution of the process for assessing the hazards of a geologic disposal system for radioactive waste and, similarly, nuclear power reactors, and the relationship of this process with other assessments of risk, particularly assessments of hazards from manufactured carcinogenic chemicals during use and disposal. This perspective reviews the common history of scientific concepts for risk assessment developed until the 1950s. Computational tools and techniques developed in the late 1950s and early 1960s to analyze the reliability of nuclear weapon delivery systems were adopted in the early 1970s for probabilistic risk assessment of nuclear power reactors, a technology for which behavior was unknown. In turn, these analyses became an important foundation for performance assessment of nuclear waste disposal in the late 1970s. The evaluation of risk to human health and the environment from chemical hazards is built on methods for assessing the dose response of radionuclides in the 1950s. Despite a shared background, however, societal events, often in the form of legislation, have affected the development path for risk assessment for human health, producing dissimilarities between these risk assessments and those for nuclear facilities. An important difference is the regulators interest in accounting for uncertainty.

Historical background on performance assessment for the Waste Isolation Pilot Plant

Abstract 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 EPAs decision was primarily based on the results from a performance assessment conducted in 1996, which is summarized in this special issue of Reliability Engineering and System Safety . This performance assessment was the culmination of four preliminary performance assessments conducted between 1989 and 1992. This paper 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.

Transport modeling in performance assessments for the Yucca Mountain disposal system for spent nuclear fuel and high-level radioactive waste.

This paper summarizes modeling of radionuclide transport in the unsaturated and saturated zone conducted between 1984 and 2008 to evaluate feasibility, viability, and assess compliance of a repository for spent nuclear fuel and high-level radioactive waste at Yucca Mountain, Nevada. One dimensional (1-D) transport for a single porosity media without lateral dispersion was solved in both the saturated zone (SZ) and unsaturated zone (UZ) for the first assessment in 1984 but progressed to a dual-porosity formulation for the UZ in the second assessment in 1991. By the time of the viability assessment, a dual-permeability transport formulation was used in the UZ. With the planned switch to a dose performance measure, individual dose from a drinking water pathway was evaluated for the third assessment in 1993 and from numerous pathways for the viability assessment in 1998 and thereafter. Stream tubes for transport in the SZ were initially developed manually but progressed to particle tracking in 1991. For the viability assessment, particle tracking was used to solve the transport equations in the 3-D UZ and SZ flow fields. To facilitate calculations, the convolution method was also used in the SZ for the viability assessment. For the site recommendation in 2001 and licensing compliance analysis in 2008, the 3-D transport results of the SZ were combined with 1-D transport results, which evaluated decay of radionuclides, in order to evaluate compliance with groundwater protection requirements. Uncertainty in flow within the unsaturated and saturated zone was generally important to explaining the spread in the individual dose performance measure.

Hazards and scenarios examined for the Yucca Mountain disposal system for spent nuclear fuel and high-level radioactive waste

This paper summarizes various hazards identified between 1978 when Yucca Mountain, located in arid southern Nevada, was first proposed as a potential site and 2008 when the license application to construct a repository for spent nuclear fuel and high-level radioactive waste was submitted. Although advantages of an arid site are many, hazard identification and scenario development have generally recognized fractures in the tuff as important features; climate change, water infiltration and percolation, and an oxidizing environment as important processes; and igneous activity, seismicity, human intrusion, and criticality as important disruptive events to consider at Yucca Mountain. Some of the scientific and technical challenges encountered included a change in the repository design from in-floor emplacement with small packages to in-drift emplacement with large packages without backfill. This change, in turn, increased the importance of igneous and seismic hazards.

Waste package degradation from thermal and chemical processes in performance assessments for the Yucca Mountain disposal system for spent nuclear fuel and high-level radioactive waste

Abstract This paper summarizes modeling of waste container degradation in performance assessments conducted between 1984 and 2008 to evaluate feasibility, viability, and assess compliance of a repository for spent nuclear fuel and high-level radioactive waste at Yucca Mountain, Nevada. As understanding of the Yucca Mountain disposal system increased, modeling of container degradation evolved from a component of the source term in 1984 to a separate module describing both container and drip shield degradation in 2008. A thermal module for evaluating the influence of higher heat loads from more closely packed, large waste packages was also introduced. In addition, a module for evaluating drift chemistry was added in later PAs to evaluate the potential for localized corrosion of the outer barrier of the waste container composed of Alloy 22, a highly corrosion-resistant nickel–chromium–tungsten–molybdenum alloy. The uncertainty of parameters related to container degradation contributed significantly to the estimated uncertainty of performance measures (cumulative release in assessments prior to 1995 and individual dose, thereafter).

Consideration of Nuclear Criticality When Directly Disposing Highly Enriched Spent Nuclear Fuel in Unsaturated Tuff - I: Nuclear Criticality Constraints

Abstract This paper presents the mass, concentration, and volume required for a critical event to occur in homogeneous mixtures of fissile material and various other geologic materials. The fissile material considered is primarily highly enriched uranium spent fuel; however, 239Pu is considered in some cases. The non-fissile materials examined are those found in the proposed repository area at Yucca Mountain, Nevada: volcanic tuff, iron rust, concrete, and naturally occurring water. For 235U, the minimum critical solid concentration for tuff was 5 kg/m3 (similar to sandstone), and in goethite, 45 kg/m3. The critical mass of uranium was sensitive to a number of factors, such as moisture content and fissile enrichment, but had a minimum, assuming almost 100% saturation and >20% enrichment, of 18 kg in tuff as Soddyite (or 9.5 kg as UO2) and 7 kg in goethite. For 239Pu, the minimum critical solid concentration for tuff was 3 kg/m3 (similar to sandstone); in goethite, 20 kg/m3. The critical mass of plutonium was also sensitive to a number of factors, but had a minimum, assuming 100% saturation and 80-90% enrichment, of 5 kg in tuff and 6 kg in goethite.

Conceptual structure of performance assessments conducted for the Waste Isolation Pilot Plant

The Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico is being developed by the US Department of Energy as a disposal facility for transuranic waste. In support of this project, Sandia National Laboratories is conducting an ongoing performance assessment (PA) for the WIPP. The ordered triple representation for risk proposed by Kaplan and Garrick is used to provide a clear conceptual structure for this PA. This presentation describes how the preceding representation provides a basis in the WIPP PA for (1) the definition of scenarios and the calculation of scenario probabilities and consequences, (2) the separation of subjective and stochastic uncertainties, (3) the construction of the complementary cumulative distribution functions required in comparisons with the US Environmental Protection Agency`s standard for the geologic disposal of radioactive waste (i.e., 40 CFR Part 191, Subpart B), and (4) the performance of uncertainty and sensitivity studies. Results obtained in a preliminary PA for the WIPP completed in December of 1991 are used for illustration.

General Approach Used in the Performance Assessment for the Waste Isolation Pilot Plant

This paper discusses the technical approach and rationale of the performance assessments (PAs) conducted for the Waste Isolation Pilot Plant that contributed to the success of the certification in 1998. The PAs were iterated: there were four preliminary PAs between 1989 and 1992 and the certification PA in 1996. Although many changes occurred between the 1992 PA and 1996 PA, the preliminary iterations provided a strong foundation on quality assurance, parameter and model databases, documentation, and peer review. The seven general steps of a PA are used to provide a brief overview of their history. The paper then delves into the rationale used for the most computationally intense step of a PA, the consequence evaluation. For this evaluation, Sandia National Laboratories mostly used detailed models when evaluating the probabilistic performance of the repository under the stylized conditions specified by the U.S. Environmental Protection Agency (EPA). A computational tool, CAMCON, was specifically tailored for this approach. Many advantages were gained by using detailed models directly in the PA, rather than using simplified results of these models. One important advantage was that models and parameters remained fairly unbiased (i.e., the analysis had only a few major conservative assumptions rather than many, unquantified conservatisms). This approach allowed Sandia to faithfully provide a “reasonable expectation” of repository performance, as specified by the EPA.

The use of formal and informal expert judgments when interpreting data for performance assessments

This paper discusses the general process by which data and information are compiled and used for defining modeling parameters. These modeling parameters are input for the mathematical models that are used in performance assessments of the Waste Isolation Pilot Plant (WIPP), near Carlsbad, NM, which is designed to safely manage, store, and dispose of transuranic radioactive wastes. The physical and temporal scales, and the difficulty of obtaining measurements in geologic media make interpretation of measured data, including the difference between laboratory experiment scale and repository scale, important tasks. In most instances, standard scientific practices can ensure consistency of data use. To illustrate this point, an example is provided of the interpretation of field measurements of intrinsic permeability for use in computational models using the bootstrap technique. In some cases, sufficient information can never be collected, interpretation of the information is controversial, or information from diverse disciplines must be used. A procedure that formalizes the standard scientific practices under these conditions has been developed. An example is provided of how this procedure has been applied in eliciting expert judgments on markers to deter inadvertant human intrusion into the WIPP.

Consideration of Criticality when Directly Disposing Highly Enriched Spent Nuclear Fuel in Unsaturated Tuff: Bounding

Although idealized calculations of the potential for an atomic explosion within a repository can make headlines, a more technically useful assessment is a systematic, multidisciplinary, integrated analysis that uses a set of consistent assumptions of disposal system performance. The analysis described here, called a performance assessment, employs the same general approach to study the potential of a critical mass assembly as has been used to examine other potentially disruptive sce- narios in a nuclear waste disposal system. This report presents one of two approaches—bounding calculations-which were used in a major study in 1994 to examine the possibility of a criticality in a repository. The bounding probabilities in this study are rough and do not entirely dismiss the pos- sibility of a critical condition; however, they do point to the difficulty of creating conditions under which a critical mass could be assembled (i.e., corrosion of containers, separation of neutron absorbers from the fissile material, and collapse or precipitation of the fissile material) and, more important, how significant the geochemical and hydrologic phenomena are in examining this criti- cality issue. Furthermore, the study could not conceive of a mechanism that was consistent with conditions under which an atomic explosion could occur, i.e., first, the manner in which. fissile material could be collected and, then, how it would be assembled (or diffused outward) within microseconds. In addition, should a criticality occur in or near a container in the future, the bound- ing consequence calculations in this study showed that fissions from one critical event (<-l&O fis- sions, if similar to aqueous and metal accidents and experiments) are quite small compared to the amount of fissions represented by the spent nuclear fuel itself. Also, if it is assumed that the con- tainers necessary to hold the highly enriched spent nuclear fuel in this study went critical once per day for 1 million years, creating an energy release of about l02Ofissions, the number of fissions equals about l028, which corresponds to only 1% of the fission inventory in a repository containing 70,000 metric tons of heavy metal (MTHM) (the expected size for the proposed repository at Yucca Mountain, Nevada).