Characterization of subjective uncertainty in the 1996 performance assessment for the Waste Isolation Pilot Plant

The 1996 performance assessment (PA) for the Waste Isolation Pilot Plant (WIPP) maintains a separation between stochastic (i.e. aleatory) and subjective (i.e. epistemic) uncertainty, with stochastic uncertainty arising from the possible disruptions that could occur at the WIPP over the 10,000 yr regulatory period specified by the US Environmental Protection Agency (40 CFR 191, 40 CFR 194) and subjective uncertainty arising from an inability to uniquely characterize many of the inputs required in the 1996 WIPP PA. The characterization of subjective uncertainty is discussed, including assignment of distributions, uncertain variables selected for inclusion in analysis, correlation control, sample size, statistical confidence on mean complementary cumulative distribution functions, generation of Latin hypercube samples, sensitivity analysis techniques, and scenarios involving stochastic and subjective uncertainty.     

Summary discussion of the 1996 performance assessment for the Waste Isolation Pilot Plant

The Waste Isolation Pilot Plant (WIPP) is under development by the US Department of Energy (DOE) for the geologic disposal of transuranic waste. The construction of complementary cumulative distribution functions (CCDFs) for total radionuclide release from the WIPP to the accessible environment is described. The resultant CCDFs (i) combine releases due to cuttings and cavings, spallings, direct brine release, and long-term transport in flowing groundwater; (ii) fall substantially to the left of the boundary line specified by the US Environmental Protection Agency's (EPA's) standard 40 CFR 191 for the geologic disposal of radioactive waste; and (iii) constitute an important component of the DOE's successful Compliance Certification Application to the EPA for the WIPP. Insights and perspectives gained in the performance assessment (PA) that led to these CCDFs are described, including the importance of: (i) an iterative approach to PA; (ii) uncertainty and sensitivity analysis; (iii) a clear conceptual model for the analysis; (iv) the separation of stochastic (i.e. aleatory) and subjective (i.e. epistemic) uncertainty; (v) quality assurance procedures; (vi) early involvement of peer reviewers, regulators, and stakeholders; (vii) avoidance of conservative assumptions; and (viii) adequate documentation.     

Regulatory basis for the Waste Isolation Pilot Plant performance assessment

The Waste Isolation Pilot Plant (WIPP) is the first operational repository designed for the safe disposal of transuranic (TRU) radioactive waste from the defense programs of the US Department of Energy (DOE). The US Environmental Protection Agency (EPA) is responsible for certifications and regulation of the WIPP facility for the radioactive components of the waste. The EPA has promulgated general radioactive waste disposal standards at 40 CFR Part 191, and WIPP-specific criteria to implement and interpret the generic disposal standards at 40 CFR Part 194. In October 1996, the DOE submitted its Compliance Certification Application (CCA) to the EPA to demonstrate compliance with the disposal standards at Subparts B and C of 40 CFR Part 191. This paper summarizes the development of the overall legal framework for radioactive waste disposal at the WIPP, the parallel development of the WIPP performance assessment (PA), and how the EPA disposal standards and implementing criteria formed the basis for the CCA WIPP PA. The CCA resulted in a certification in May 1998 by the EPA of the WIPP's compliance with the EPA's disposal standard, thus enabling the WIPP to begin radioactive waste disposal.     

Uncertainty and sensitivity analysis for two-phase flow in the vicinity of the repository in the 1996 performance assessment for the Waste Isolation Pilot Plant: undisturbed conditions

Uncertainty and sensitivity analysis results obtained in the 1996 performance assessment for the Waste Isolation Pilot Plant are presented for two-phase flow in the vicinity of the repository under undisturbed conditions. Techniques based on Latin hypercube sampling, examination of scatterplots, stepwise regression analysis, partial correlation analysis and rank transformations are used to investigate brine inflow, gas generation, repository pressure, brine saturation, and brine and gas outflow. Of the variables under study, repository pressure is potentially the most important due to its influence on spallings and direct brine releases, with the uncertainty in its value being dominated by the extent to which the microbial degradation of cellulose takes place, the rate at which the corrosion of steel takes place, and the amount of brine that drains from the surrounding disturbed rock zone into the repository.     

Review and perspectives on spallings release models in the 1996 performance assessment for the Waste Isolation Pilot Plant

The Waste Isolation Pilot Plant was licensed for disposal of transuranic wastes generated by the US Department of Energy. The facility consists of a repository mined in a bedded salt formation, approximately 650 m below the surface. Regulations promulgated by the US Environmental Protection Agency require that performance assessment calculations for the repository include the possibility that an exploratory drilling operation could penetrate the waste disposal areas at some time in the future. Release of contaminated solids could reach the surface during a drilling intrusion. One of the mechanisms for release, known as spallings, can occur if gas pressures in the repository exceed the hydrostatic pressure of a column of drilling mud. Calculation of solids releases for spallings depends critically on the conceptual models for the waste, for the spallings process, and assumptions regarding driller parameters and practices. This paper presents a review of the evolution of these models during the regulatory review of the Compliance Certification Application for the repository. A summary and perspectives on the implementation of conservative assumptions in model development are also provided.     

Assignment of probability distributions for parameters in the 1996 performance assessment for the Waste Isolation Pilot Plant. Part 1: description of process

A managed process was used to consistently and traceably develop probability distributions for parameters representing epistemic uncertainty in four preliminary and the final 1996 performance assessment (PA) for the Waste Isolation Pilot Plant. Between 67 probability density functions (PDFs) in the 1989 PA and 236 PDFs in the 1996 PA were assigned by a parameter development team, using a process described in a companion paper. In the five iterative PAs conducted, the most commonly used distributions were the uniform PDF and piecewise-uniform PDF (also referred to as a piecewise-linear cumulative distribution function (CDF)). Other distributions used included the truncated normal, truncated Student-t, and triangular PDFs. In a few instances, a discrete delta (piecewise-uniform CDF), beta, and exponential PDF were also used. The PDFs produced for the 24 most important parameters observed in the five iterative PAs are presented. As background, the list of 194 parameters documented in the first 1989 PA through the 1471 parameters documented in the 1996 compliance PA are also provided.     

Radioactive and nonradioactive waste intended for disposal at the Waste Isolation Pilot Plant

Transuranic (TRU) waste generated by the handling of plutonium during research on or production of U.S. nuclear weapons will be disposed of in the Waste Isolation Pilot Plant (WIPP). This paper describes the physical and radiological properties of the TRU waste that will be deposited in the WIPP. This geologic repository will accommodate up to 175,564 m³ of TRU waste, corresponding to 168,485 m³ of contact-handled (CH-) TRU waste and 7079 m³ of remote-handled (RH-) TRU waste. Approximately 35% of the TRU waste is currently packaged and stored (i.e. legacy) waste, with the remainder of the waste to be packaged or generated and packaged in activities before the year 2033, which is the closure time for the repository. These wastes were produced at 27 U.S. Department of Energy (DOE) sites in the course of generating defense nuclear materials. The radionuclide and nonradionuclide inventories for the TRU wastes described in this paper were used in the 1996 WIPP Compliance Certification Application (CCA) performance assessment calculations by the Sandia National Laboratories/New Mexico (SNL/NM).     

Latin hypercube sampling and the propagation of uncertainty in analyses of complex systems

The following techniques for uncertainty and sensitivity analysis are briefly summarized: Monte Carlo analysis, differential analysis, response surface methodology, Fourier amplitude sensitivity test, Sobol' variance decomposition, and fast probability integration. Desirable features of Monte Carlo analysis in conjunction with Latin hypercube sampling are described in discussions of the following topics: (i) properties of random, stratified and Latin hypercube sampling, (ii) comparisons of random and Latin hypercube sampling, (iii) operations involving Latin hypercube sampling (i.e. correlation control, reweighting of samples to incorporate changed distributions, replicated sampling to test reproducibility of results), (iv) uncertainty analysis (i.e. cumulative distribution functions, complementary cumulative distribution functions, box plots), (v) sensitivity analysis (i.e. scatterplots, regression analysis, correlation analysis, rank transformations, searches for nonrandom patterns), and (vi) analyses involving stochastic (i.e. aleatory) and subjective (i.e. epistemic) uncertainty.     

Conceptual basis for the definition and calculation of expected dose in performance assessments for the proposed high-level radioactive waste repository at Yucca Mountain, Nevada

A deep geologic repository for high-level radioactive waste is under development by the US Department of Energy (DOE) at Yucca Mountain (YM), Nevada. As mandated in the Energy Policy Act of 1992, the US Environmental Protection Agency has promulgated public health and safety standards (i.e., 40 CFR Part 197) for the YM repository, and the US Nuclear Regulatory Commission has promulgated licensing standards (i.e., 10 CFR Parts 2, 19, 20, etc.) consistent with 40 CFR Part 197 that the DOE must establish are met in order for the YM repository to be licensed for operation. Important requirements in 40 CFR Part 197 and 10 CFR Parts 2, 19, 20, etc. relate to the determination of expected (i.e., mean) dose to a reasonably maximally exposed individual (RMEI) and the incorporation of uncertainty into this determination. This paper is the first part of a two-part presentation and describes how general and typically nonquantitative statements in 40 CFR Part 197 and 10 CFR Parts 2, 19, 20, etc. can be given a formal mathematical structure that facilitates both the calculation of expected dose to the RMEI and the appropriate separation in this calculation of aleatory uncertainty (i.e., randomness in the properties of future occurrences such as igneous and seismic events) and epistemic uncertainty (i.e., lack of knowledge about quantities that are imprecisely known but assumed to have constant values in the calculation of expected dose to the RMEI). The second part of this presentation is contained in the following paper, “Computational Implementation of Sampling-Based Approaches to the Calculation of Expected Dose in Performance Assessments for the Proposed High-Level Radioactive Waste Repository at Yucca Mountain, Nevada,” and both describes and illustrates sampling-based procedures for the estimation of expected dose and the determination of the uncertainty in estimates for expected dose.     

Computational implementation of sampling-based approaches to the calculation of expected dose in performance assessments for the proposed high-level radioactive waste repository at Yucca Mountain, Nevada

A deep geologic repository for high-level radioactive waste is under development by the US Department of Energy (DOE) at Yucca Mountain (YM), Nevada. As mandated in the Energy Policy Act of 1992, the US Environmental Protection has promulgated public health and safety standards (i.e., 40 CFR Part 197) for the YM repository, and the US Nuclear Regulatory Commission has promulgated licensing standards (i.e., 10 CFR Parts 2, 19, 20, etc.) consistent with 40 CFR Part 197 that the DOE must establish are met in order for the YM repository to be licensed for operation. Important requirements in 40 CFR Part 197 and 10 CFR Parts 2, 19, 20, etc. relate to the determination of expected (i.e., mean) dose to a reasonably maximally exposed individual (RMEI) and the incorporation of uncertainty into this determination. This paper is the second part of a two-part presentation on the determination of expected dose to the RMEI in the context of 40 CFR Part 197 and 10 CFR Parts 2, 19, 20, etc. The first part of this presentation is contained in the preceding paper, “Conceptual Basis for the Definition and Calculation of Expected Dose in Performance Assessments for the Proposed High-Level Radioactive Waste Repository at Yucca Mountain, Nevada”, and describes how general and typically nonquantitative statements in 40 CFR Part 197 and 10 CFR Parts 2, 19, 20, etc. can be given a formal mathematical structure that facilitates both the calculation of expected dose to the RMEI and the appropriate separation in this calculation of aleatory uncertainty (i.e., randomness in the properties of future occurrences such as igneous and seismic events) and epistemic uncertainty (i.e., lack of knowledge about quantities that are poorly known but assumed to have constant values in the calculation of expected dose to the RMEI). The present paper describes and illustrates sampling-based procedures for the estimation of expected dose and the determination of the uncertainty in estimates for expected dose.     

The geologic and hydrogeologic setting of the Waste Isolation Pilot Plant

The Waste Isolation Pilot Plant (WIPP) is a mined repository for the permanent disposal of transuranic wastes. It has been constructed by the United States Department of Energy (DOE) in semiarid, sparsely inhabited rangeland in southeastern New Mexico. The disposal area is 655 m below the land surface, in bedded salt of the Late Permian (approximately 255 million-years-old) Salado Formation. The extremely low permeability of the halite and other evaporite rocks provide the primary geologic barrier assuring long-term (10,000-year-plus) isolation of the radioactive waste from the accessible environment. Extensive geologic investigations during site characterization have provided information on the stratigraphy, structure, and natural resources of the region (including hydrocarbons, potash, and groundwater), and have investigated the potential for disruption by processes such as dissolution, salt deformation, tectonic activity, and climate change. Hydrogeologic investigations have documented the physical properties of the evaporite rocks, and have identified the Culebra Dolomite Member of the overlying Rustler Formation as the most transmissive water-bearing unit in the region. If the evaporite barriers are breached by accidental drilling intrusion, the Culebra would provide the most likely pathway for radionuclide transport away from the site. Although water in the Culebra is of poor quality and none is currently used for human consumption, groundwater flow and potential radionuclide transport in the unit have been studied in detail. Results of geologic and hydrogeologic studies of the WIPP region indicate that the geologic and hydrogeologic features of the site will provide effective long-term containment of radionuclides. Geologic and hydrogeologic information is used in the performance assessment that supported the DOE’s compliance certification application to the United States Environmental Protection Agency.