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1. INTRODUCTION

The U.S. Department of Energy (DOE) has developed a comprehensive set of technical documents to support a decision to be made by the Secretary of Energy on whether to recommend the Yucca Mountain site as the location for development of a geologic repository. This Yucca Mountain Site Suitability Evaluation is one of these technical documents; it specifically addresses the provisions of the DOE site suitability guidelines set out in 10 CFR Part 963 (66 FR 57298). Suitability evaluations have been completed for both preclosure and postclosure periods. The methodology and criteria used in both the preclosure and postclosure evaluations are discussed in this report in Sections 2 and 3, respectively. Results of the site suitability evaluations are summarized in Section 4.

The Nuclear Waste Policy Act of 1982 (42 U.S.C. 10101 et seq.) sets out a process for the recommendation and approval of a site for development of a repository, and it calls for the Secretary of Energy to provide a comprehensive statement of the basis for any site recommendation. With release of the initial version of the Yucca Mountain Science and Engineering Report (DOE 2001a) in May 2001, the DOE opened a public comment period on the Secretary of Energy's consideration of the possible recommendation of the Yucca Mountain site to the President. The report described the results of scientific and engineering studies completed to date, the waste forms to be disposed, the repository and waste package designs, and the results of assessments of the long-term performance of the potential repository. The scientific investigations included site characterization studies of the geologic, hydrologic, and geochemical environment and an evaluation of how conditions might evolve over time. These analyses considered a range of processes that would operate in and around the potential repository. The initial version of the Yucca Mountain Science and Engineering Report (DOE 2001a) was issued to provide information for public review and comment in advance of public hearings to inform area residents that recommendation of the site is being considered and to receive their comments. It also supported the process of informing the public, elected officials, affected units of government or Indian tribes, regulatory agencies, review groups, and other interested parties of the Secretary's consideration of a possible recommendation of the site.

Yucca Mountain Preliminary Site Suitability Evaluation (DOE 2001b), issued in August 2001, provided additional information to support public comment on whether the Secretary should recommend the Yucca Mountain site for development of a repository. It provided results of a preliminary evaluation of the site's performance against the standards set out in the DOE's proposed suitability guidelines in 10 CFR Part 963 (64 FR 67054). Descriptions of the scientific and engineering studies that served as the basis for the preliminary suitability evaluation were provided in the initial version of the Yucca Mountain Science and Engineering Report (DOE 2001a) and in its comprehensive set of scientific and technical integrating documents. FY01 Supplemental Science and Performance Analyses (BSC 2001a; BSC 2001b) provided additional information and analyses that were developed following completion of the initial version of the Yucca Mountain Science and Engineering Report (DOE 2001a) to better understand the performance attributes of a lower-temperature repository operating mode, the potential for quantification of uncertainties, and the conservatisms in earlier performance analyses.

This Yucca Mountain Site Suitability Evaluation describes information that supports an updated evaluation of the suitability of the Yucca Mountain site for the location of a geologic repository for disposal of spent nuclear fuel and high-level radioactive waste. Information from the following sources was used in updating what was presented in the Yucca Mountain Preliminary Site Suitability Evaluation (DOE 2001b):

The DOE's site suitability guidelines establish the criteria for assessing suitability during the preclosure and postclosure periods and the methodology for evaluating those criteria. The criteria were derived considering physical processes pertinent to understanding the performance of a disposal system at Yucca Mountain. The criteria and methodology for this suitability evaluation are structured to be consistent with NRC performance objectives (10 CFR 63.111 and 10 CFR 63.113 [66 FR 55732]). Table 1-1 lists preclosure suitability evaluation methods and suitability criteria set out in 10 CFR Part 963 (66 FR 57298) and where they are discussed in this site suitability evaluation. Table 1-2 lists the postclosure suitability evaluation methods and suitability criteria set out in 10 CFR Part 963 and where they are discussed in this site suitability evaluation. Discussions on applications of the methodologies for the preclosure and postclosure suitability evaluations are presented in Sections 2.2 and 3.2, respectively. The evaluation of the preclosure suitability criteria is presented in Section 2.4. Information summarizing the considerations of the postclosure suitability criteria is provided in Section 3.3.

The evaluation of the site pursuant to the site suitability guidelines in 10 CFR Part 963 (66 FR 57298) begins with an evaluation of preclosure suitability. This assessment entails the use of a preclosure safety evaluation, which examines the ability of the preliminary design of surface and subsurface facilities and the waste package to contain and limit releases of radioactive materials under both normal and accident conditions, in accordance with the standard contained in 10 CFR 63.111(a) and (b) and 10 CFR 63.204 (66 FR 55732), which are referenced in 10 CFR 963.2. The elements at 10 CFR 963.13(a) and (b) describe what is contemplated in the preclosure suitability evaluation method. The evaluation using this method is described in Sections 2.2 and 2.3. The assessment of preclosure suitability pursuant to the preclosure suitability guidelines of 10 CFR Part 963 entails consideration of the preclosure suitability criteria at 10 CFR 963.14(a) through (d). Those evaluations are described in Section 2.4. Results of the preclosure suitability evaluation are summarized in Section 4.1.

The evaluation of the site pursuant to the site suitability guidelines in 10 CFR Part 963 (66 FR 57298) also entails an evaluation of postclosure suitability. Consistent with the risk-informed, performance-based approach contained in the NRC licensing regulations, the suitability guidelines call for a total system performance assessment (TSPA) which, according to the 10 CFR 963.2 guidelines, involves

    1. identify the features, events and processes (except human intrusion) that might affect the Yucca Mountain disposal system and their probabilities of occurring during 10,000 years after disposal;

    2. examine the effects of those features, events, processes, and sequences of events and processes (except human intrusion) on the performance of the Yucca Mountain disposal system; and

    3. estimate the dose incurred by the reasonably maximally exposed individual, including associated uncertainties, as a result of releases caused by all significant features, events, processes, and sequences of events and processes, weighted by their probability of occurrence.

The methodology for the postclosure suitability evaluation, which is discussed in Section 3.2, has produced a TSPA that conforms to the provisions of the guidelines and comprises computer models and sensitivity analyses that forecast radiation exposure and radioactive releases. The TSPA evaluates the ability of the natural and engineered barriers to isolate and contain releases of radioactive materials.

10 CFR 963.16 (66 FR 57298) sets out the types of information that will be considered in the performance assessment models. The results of the assessments of total system performance are described in Section 3.1 and summarized in Section 4.2. The evaluation of the site using the postclosure suitability guidelines of 10 CFR Part 963 entails consideration of the suitability criteria at 10 CFR 963.17(a). These considerations are described in Section 3.3.

Section 4 provides a summary of the results of the evaluations provided in Sections 2 and 3. The results of the preclosure and postclosure suitability evaluations are shown in comparison with applicable radiation protection standards, which the DOE's site suitability guidelines (10 CFR 963.2 [66 FR 57298]) have established as the NRC preclosure dose limits in 10 CFR 63.204 and 10 CFR 63.111(a) and (b), the NRC postclosure dose limits in 10 CFR 63.311 and 10 CFR 63.321, and the NRC radionuclide concentration limits in 10 CFR 63.331 (66 FR 55732).

1.1 REGULATORY FRAMEWORK FOR A SUITABILITY EVALUATION

This section summarizes the regulatory framework for assessing the suitability of the Yucca Mountain disposal system (alternatively referred to as the geologic repository) for high-level radioactive waste and spent nuclear fuel. Sections 2 and 3 of this report present the results of the DOE preclosure and postclosure suitability evaluations, respectively. These evaluations address whether the Yucca Mountain site is suitable for the location of a geologic repository, based on the criteria and methodologies set forth in 10 CFR Part 963 (66 FR 57298). The suitability evaluations follow the methods and criteria for the preclosure and postclosure periods, as set out in the site suitability guidelines.

Results of the evaluation will be part of the bases for any site recommendation decision. In this evaluation, Subpart B of 10 CFR Part 963 (66 FR 57298) is used to assess whether a Yucca Mountain disposal system is likely to meet applicable radiation protection standards. As discussed earlier, 10 CFR 963.2 has established applicable radiation protection standards as those established by the NRC in 10 CFR Part 63 (66 FR 55732). The NRC, in 10 CFR Part 63, adopted the concepts and standards established by the EPA in 40 CFR Part 197 for both the preclosure and postclosure periods.

Under these standards, during the preclosure period, no member of the public in the general environment can receive more than an annual dose of 15 mrem (10 CFR 63.204 [66 FR 55732]). During the postclosure period of regulatory compliance (10,000 years), the reasonably maximally exposed individual in the accessible environment can receive no more than an annual dose of 15 mrem in the case where there is no human intrusion into the repository (10 CFR 63.311). If exposure from an inadvertent human intrusion is projected to occur within 10,000 years, the DOE must demonstrate that the reasonably maximally exposed individual would receive no more than an annual dose of 15 mrem as a result of the human intrusion (10 CFR 63.321). If exposure from a human intrusion is projected to occur after 10,000 years, the analysis of the human intrusion scenario must be presented in the environmental impact statement, and the dose limits for the human intrusion standard would not apply. The standards also restrict the concentration of selected radionuclides in groundwater.

If the Secretary recommends the site and the President approves the Yucca Mountain site for development of a repository, the Secretary will apply to the NRC for an authorization to construct a Yucca Mountain disposal system. If the NRC authorizes construction, after construction is substantially completed, the Secretary will apply for an NRC license to receive and possess radioactive waste. If the NRC issues the license, waste disposal could begin in 2010 and continue for up to 100 years, based on a repository capacity of 70,000 MTHM. The Nuclear Waste Policy Act of 1982 (42 U.S.C. 10101 et seq.) directs the NRC to prohibit emplacement of more than 70,000 MTHM until such time as a second repository, if authorized by Congress, is in operation.

The NRC licensing rules require that the DOE begin a performance confirmation program during site characterization and continue the program until the Yucca Mountain disposal system is permanently closed. The program includes in situ monitoring, laboratory and field testing, and in situ experiments (10 CFR 63.131(c) [66 FR 55732]). The NRC licensing regulations also require that the geologic repository operations area be designed so that any or all of the emplaced waste could be retrieved on a reasonable schedule starting at any time up to 50 years after waste emplacement operations are initiated (10 CFR 63.111(e)).

1.2 BACKGROUND INFORMATION

The United States has evaluated methods for the safe storage and disposal of radioactive waste for more than 40 years. The DOE (including the national laboratories) and the U.S. Geological Survey (USGS) began studying Yucca Mountain in the late 1970s. Detailed investigations of the geology, hydrology, geochemistry, and other characteristics of the site have been performed since 1986 to determine whether it is a suitable place to build a geologic repository.

The concept of disposing of waste in desert regions of the Southwest was first proposed by the USGS in the 1970s. In 1976, the director of the USGS identified a number of positive attributes in and around the Nevada Test Site that would make positive contributions to geologic disposal, including multiple natural barriers, remoteness, and an arid climate (McKelvey 1976). In 1981, a USGS scientist documented that water tables in the desert Southwest are among the deepest in the world, and the geologic setting includes multiple natural barriers that could isolate waste for "tens of thousands to perhaps hundreds of thousands of years" (Winograd 1981). In contrast to the strategy for isolating waste in salt or sites below the water table, waste could be disposed near the Nevada Test Site well below the ground surface, but still well above the water table. Following the initial USGS recommendation, the DOE sponsored investigations of the feasibility of disposal above the water table. Formal site characterization at Yucca Mountain began in 1986.

Yucca Mountain is a flat-topped, north–south trending ridge that rises approximately 300 m (1,000 ft) above the adjacent valleys. The geologic setting is composed of a layered sequence of alternating welded and nonwelded tuffs. The geologic setting has changed little over the last several million years. Investigations of potential natural hazards at the site reveal that a repository would be unlikely to be disrupted by volcanoes, earthquakes, erosion, or other geologic occurrences. The geologic repository operations area would be designed to withstand vibratory ground motion from seismic events as required by NRC licensing regulations. Earthquakes would not compromise postclosure performance of the Yucca Mountain disposal system because of their limited effect underground and because underground facilities would also be designed to accommodate the effects of seismic events. Waste packages would not be placed where substantial fault movement could occur. The likelihood and consequences of volcanic activity on the performance of a Yucca Mountain disposal system are considered in the evaluations of the suitability of the site.

The arid environment of Yucca Mountain results in a relatively small total amount of water available to transport radionuclides. Because the transport of radionuclides by water is the primary means by which mobile radionuclides could leave the site, the geologic repository and the waste packages would be designed to keep the waste dry for as long as practicable. This could be achieved through a number of different approaches, including either keeping emplacement drift temperatures high enough to drive water away, thereby limiting the amount of water that could come in contact with the drip shields during early times, or keeping temperatures low enough near the waste packages to enhance their corrosion resistance (DOE 2002a, Section 2.1.5). The Yucca Mountain disposal system design used as the basis of this suitability evaluation is flexible such that it could be operated over a range of thermal conditions.

The repository design and thermal operating mode also would take advantage of natural processes, such as matrix diffusion and capillary pressure, that would divert water flow around or away from the emplacement drifts and the waste packages. Solubility limits, retardation, diffusion, and other properties affecting particle transport through the geologic setting also are barriers to the movement of radionuclides (DOE 2002a, Section 4.5). The combination of natural and engineered barriers working in a complementary manner provides a high level of confidence that the overall, integrated Yucca Mountain disposal system would perform in a manner that protects public health and safety.

On average, the Yucca Mountain site receives about 190 mm (7.5 in.) (CRWMS M&O 2000a, Table 3.5-2) of rain and snow per year. Nearly all the precipitation (above 95 percent) runs off, evaporates, or is taken up by plants. Geological information indicates that the regional climate has changed over the past million years and that long-term average precipitation is higher than the current average. Even at higher precipitation rates, most of the water would still run off the mountain or evaporate rather than soak into the ground and reach the repository.

The water table at the Yucca Mountain site is approximately 500 to 800 m (1,600 to 2,600 ft) below the surface of the mountain. The underground facility would be excavated about 200 to 500 m (660 to 1,600 ft) below ground surface and about 210 to 390 m (690 to 1,300 ft) or, on average, about 300 m (1,000 ft) above the water table (DOE 2002a, Section 4.2.1.2.2). Any precipitation that does not run off, evaporate, or get taken up by plants at the surface would have to seep through approximately 300 m (1,000 ft) of rock before reaching the geologic repository. Most of the water moving downward past the elevation of the underground facility would move through fractures in the rock pillars between emplacement drifts.

Natural hydrologic processes, including capillary forces and the tendency of water to flow along drift walls, would limit the amount of water that could encounter waste. Any water that could move through the natural barrier system above the underground facility would reach multiple engineered barriers before it could contact the waste. Even after the drip shield and waste package are breached, water would have to move through the package, dissolve the waste form, and carry the radionuclides out of the breached waste packages. The water would carry a small fraction of radionuclides that are mobile through the engineered barrier system and downward approximately another 300 m (1,000 ft) through the natural barriers beneath the geologic repository, including the unsaturated zone, before reaching the water table. After reaching the water table, the water moves laterally to the accessible environment where it could be used or consumed by the reasonably maximally exposed individual (i.e., a hypothetical receptor) as defined at 10 CFR 63.312 (66 FR 55732) and referenced in 10 CFR 963.2 (66 FR 57298). As the water travels through the unsaturated and saturated zones, numerous geologic units would capture and slow the radionuclides' transport and reduce their concentration.

The Yucca Mountain site is remote from population centers. It is about 160 km (100 mi) northwest of Las Vegas, Nevada, and the nearest residents live approximately 20 km (12 mi) away. In addition, the Yucca Mountain site is located on land controlled by three federal agencies: the U.S. Department of Defense (Nellis Air Force Range), the U.S. Department of the Interior (Bureau of Land Management), and the DOE (Nevada Test Site). Because of its history as the site of more than 900 nuclear weapons tests, access to the Nevada Test Site will be controlled for the foreseeable future (DOE 1998a, Section 2.2) and thereby reduce the likelihood of inadvertent human intrusion.

1.3 APPROACH TO A SITE SUITABILITY EVALUATION

The 10 CFR Part 963 guidelines (66 FR 57298) set out methodologies for evaluating the site, considering specific criteria related to the natural and engineered systems, particularly those important to the performance of the Yucca Mountain disposal system. The guidelines include preclosure and postclosure methods at 10 CFR 963.13 and 10 CFR 963.16 and preclosure and postclosure criteria at 10 CFR 963.14 and 10 CFR 963.17. The methods and criteria describe the scope of technical information that will support a suitability evaluation. The preclosure and postclosure evaluations contemplated by 10 CFR 963.12 and 10 CFR 963.15 reference the preclosure and postclosure radiation protection standards of the NRC.

The site suitability evaluation presented in this report is supported by a comprehensive set of scientific and technical integrating documents, which explain in detail the basis for the analyses and results presented. There are six primary reference documents, each with numerous supporting documents, that present the technical information that was used in this site suitability evaluation. These six documents and their relationships are:

Since completion of the performance analyses that provided the technical basis for the preclosure and postclosure suitability evaluations, additional information has been collected through the ongoing testing program and other sources. Technical Update Impact Letter Report (BSC 2001d) documents this additional information and evaluates the potential impacts it could have on the results of the TSPA analyses and the preclosure safety evaluation. While this information is not included in this site suitability evaluation, it provides supplemental information that, for the most part, confirms the results (and in some cases reduces the uncertainty) of the postclosure and preclosure technical analyses used as the technical basis for this suitability evaluation. Where this is not the case, the impacts of the additional information on the results of the analyses are not significant. Results of the impact reviews and conclusions drawn from the evaluation of the additional information are presented in Sections 3 and 4, respectively, of Technical Update Impact Letter Report (BSC 2001d).

1.3.1 Technical Approach for the Suitability Evaluation

The DOE has developed a flexible design for a Yucca Mountain disposal system that will give future generations the choice of either closing and sealing the underground facility as early as allowable under applicable regulations or keeping it open and monitoring it for a longer time period. The Yucca Mountain disposal system design concept also provides flexibility for operation over a range of thermal operating modes. The range of these thermal operating modes was examined to identify the potential performance benefits of different environmental conditions (for example, a range of lower temperatures and associated humidity conditions) in the emplacement drifts.

A key aspect of design flexibility for the purposes of this evaluation of the suitability of the Yucca Mountain site is the ability to support a range of thermal operating modes. These operating modes may have different attributes, such as a larger waste emplacement area or different ventilation durations and rates. Other aspects of design flexibility include (1) the ability of the repository design to support a range of construction approaches (e.g., change in emplacement drift spacing, modular or sequential construction of surface and subsurface facilities), (2) the capability to dispose of a range of radioactive waste forms and waste package sizes, and (3) the ability to continue to enhance the design to best achieve performance-related benefits identified through ongoing analyses. Monitored Geologic Repository Project Description Document (Curry 2001) describes a flexible design and operational concepts that could accommodate higher- or lower-temperature operations.

The TSPA model presented in Total System Performance Assessment for the Site Recommendation (CRWMS M&O 2000b) (i.e., the TSPA-SR model) is based on the repository design operated in a higher-temperature operating mode. In that operating mode, heat from the emplaced waste (mostly from the spent nuclear fuel) would keep the temperature of the rock around the waste packages higher than the boiling point of water after the geologic repository is closed (DOE 2002a, Section 4.2.2.3.2). The emplacement drifts were designed so the zones with temperatures above boiling extend only partway into the rock pillars between the emplacement drifts. This higher-temperature operating mode is intended to promote water condensation and drainage through the cooler portions of the pillars and away from the emplacement drifts. Analyses of the effects of the higher-temperature operating mode indicate that higher temperatures in the repository environment would effectively limit the potential for contact between water and waste packages. Such an approach could minimize the amount of water that might eventually be available to enter the drifts and contact the waste packages. However, during that period, the interaction of rock, water, and heat in and near emplacement drifts may affect rock properties and water chemistry in complex ways (through coupled processes). It has not been possible to capture these processes fully in the current models. The complexity of these coupled processes introduces uncertainty into the analyses.

The design, however, is flexible, and the repository could be operated in a range of modes that would allow the temperature and humidity in the geologic repository to be managed to reduce uncertainty. Analyses described in FY01 Supplemental Science and Performance Analyses (BSC 2001a; BSC 2001b), the TSPA Report for Final Environmental Impact Statement and Suitability Evaluation (Williams 2001a), and the TSPA Sensitivity Analyses for NRC Regulations (Williams 2001b) evaluated the effects of both higher and lower temperatures on long-term performance of the Yucca Mountain disposal system. The lower-temperature operating modes would keep the average waste package surface temperature below 85°C (185°F). These lower-temperature conditions could result in less complexity in the coupled processes and a concomitant reduction in uncertainty in modeling the processes.

FY01 Supplemental Science and Performance Analyses (BSC 2001a; BSC 2001b) describes analyses of the quantification of conservatisms and analyses undertaken to determine whether the complexity of the associated process models could be reduced. These investigations involved analyzing performance over a range of thermal operating modes to ensure that the performance evaluations appropriately considered the potentially detrimental and potentially beneficial aspects of the system performance over a range of operating modes, encompassing above- and below-boiling conditions. Temperature-sensitive parameters and coupled thermal-hydrologic-mechanical-chemical processes were considered in analyses of the effects of lower temperatures on the in-drift environment. These analyses supported sensitivity studies (i.e., studies designed to determine which models and parameters most significantly affect overall system performance) to determine if lower-temperature operating modes have the potential to reduce uncertainties in long-term performance assessments.

The evaluation of site suitability presented in this report considers the data, analyses, and results provided in Total System Performance Assessment for the Site Recommendation (CRWMS M&O 2000b), FY01 Supplemental Science and Performance Analyses (BSC 2001a; BSC 2001b), the TSPA Report for Final Environmental Impact Statement and Suitability Evaluation (Williams 2001a), and the TSPA Sensitivity Analyses for NRC Regulations (Williams 2001b).

1.3.2 Documentary Bases for the Preclosure Suitability Evaluation

The DOE guidelines identify separate suitability evaluations for the preclosure and postclosure periods. Both evaluations rely on a thorough understanding of the natural system at Yucca Mountain, including the geology and hydrology, on a regional and a site-specific scale. This information is provided in Yucca Mountain Site Description (CRWMS M&O 2000c) and in a number of process model reports (see Section 1.3.3).

The preclosure suitability evaluation method and criteria at 10 CFR 963.13 and 10 CFR 963.14 (66 FR 57298), respectively, consider the safe operation of the geologic repository operations area while waste is being received, staged, emplaced, monitored, and potentially retrieved. Waste emplacement is anticipated to last for approximately 24 to 50 years. The NRC requires that the DOE be able to retrieve the waste from the repository starting at any time up to 50 years after waste emplacement begins, unless a different time period is established by the NRC. In addition, an extended ventilation period may extend the preclosure period to 325 years following initial waste emplacement should the DOE decide to ventilate and cool the waste prior to closing the geologic repository (see Section 2.2.1).

The DOE guidelines set out a preclosure safety evaluation method that is consistent with a preclosure safety analysis that is required by the NRC at 10 CFR Part 63.112 (66 FR 55732) for licensing. It includes performance standards, analytical bases, technical justifications, and evaluations that show that safety functions will not be compromised.

The preclosure guidelines at 10 CFR 963.13 and 963.14 (66 FR 57298) provide a method and criteria for a preclosure safety evaluation. The evaluation uses the preliminary designs and operational concepts to assess the adequacy of the geologic repository operations area to perform its intended functions and prevent or mitigate the effects of postulated event sequences. The bases for the evaluation are discussed in the following paragraphs.

Preliminary Preclosure Safety Assessment for Monitored Geologic Repository Site Recommendation (BSC 2001c) presents an evaluation of preclosure safety. It is based on preliminary system, structure, and component designs and operational concepts. Information in the safety assessment, together with supporting design documents and system description documents, supports this preclosure suitability evaluation. Results of the preclosure evaluation are discussed in Section 2.

Preliminary Preclosure Safety Assessment for Monitored Geologic Repository Site Recommendation (BSC 2001c) identifies, categorizes, and analyzes design basis events2 (events that the geologic repository operations area is designed to withstand) and their consequences. Events are identified by first evaluating potential hazards that could adversely affect the site and the geologic repository operations area. Then detailed site-specific and design-specific event scenarios are developed to define the sequence of events and establish their frequency of occurrence during the preclosure time frame, which could last up to 325 years. Shorter preclosure time frames generally would apply to a higher-temperature operating mode, while longer time frames may be needed for lower-temperature operating modes (i.e., the longer the repository stays open and ventilates the heat from the waste, the cooler the system becomes). Preliminary Preclosure Safety Assessment for Monitored Geologic Repository Site Recommendation (BSC 2001c) identifies the event sequences that would be relevant during a range of preclosure time frames. Based on the frequency of occurrence, events are categorized as Category 1 or Category 2 event sequences. Events that have less than one chance in 10,000 of occurring before permanent closure do not need further analysis. This method is consistent with the site suitability guidelines' definition (10 CFR 963.2) of event sequence (66 FR 57298).

Preliminary Preclosure Safety Assessment for Monitored Geologic Repository Site Recommendation (BSC 2001c) also presents the strategy for criticality safety, radiation protection, and fire protection, along with a description of the provisions for control and management of low-level radioactive waste. Descriptions of site characteristics and facility design are provided to support the identification of hazards and the evaluation of design basis events.

Retrievability Strategy Report (CRWMS M&O 1998a), Retrieval Equipment and Strategy for WP on Pallet (CRWMS M&O 2000d), Waste Emplacement/Retrieval System Description Document (CRWMS M&O 2000e), and other relevant system description documents describe the requirements and strategy for retrievability and define the preliminary bases and criteria for design of a geologic repository operations area to permit the retrieval of any or all high-level radioactive waste and spent nuclear fuel, should it prove necessary or desirable to do so.

Repository Safety Strategy: Plan to Prepare the Safety Case to Support Yucca Mountain Site Recommendation and Licensing Considerations (CRWMS M&O 2001a, Volume 1) guides the development of design bases and criteria for structures, systems, and components important to radiological safety. The strategy helps ensure that appropriate information and documentation are available to support an evaluation against the preclosure method and criteria set out in 10 CFR 963.13 and 10 CFR 963.14 (66 FR 57298).

Figure 1-1 summarizes the preclosure suitability evaluation process. Section 4.1 provides a summary of the results of the preclosure suitability evaluation. Results are shown in comparison with applicable radiation protection standards.

1.3.3 Documentary Bases for the Postclosure Suitability Evaluation

The regulations issued by the DOE (10 CFR Part 963 [66 FR 57298]), the NRC (10 CFR Part 63 [66 FR 55732]), and the EPA (40 CFR Part 197) all involve the DOE's evaluation of the disposal system's postclosure performance using a total system performance assessment. The TSPA method set out in 10 CFR 963.16(b)(1) through (12) is consistent with the method required by the NRC at 10 CFR 63.114(a) through (g) and 10 CFR 63.115. For example, both the DOE and the NRC call for an account of uncertainties and alternative models. Both regulations require the same methods for screening out features, events, and processes (FEPs) based on their low probability or low consequence in the expected (or mean) dose.

Under the DOE's methodology set forth at 10 CFR 963.16(b)(7) (66 FR 57298), multiple lines of evidence are part of this evaluation against the postclosure guideline. Laboratory testing, field investigations, and natural analogues provide the technical basis for process models and confirm the capability of natural and engineered barriers to contribute to meeting the applicable radiation protection standards. Volume 1 of FY01 Supplemental Science and Performance Analyses (BSC 2001a) provides an evaluation of multiple lines of evidence.

The TSPA methods set forth in the regulations at 10 CFR 963.16(a)(1) (66 FR 57298) involve evaluation of postclosure performance of the Yucca Mountain disposal system in the case where there is no human intrusion into the repository. 10 CFR 963.16(a)(2) also involves evaluation of the performance of the system in the case where there is a human intrusion, in accordance with the NRC licensing regulations. The TSPA evaluation method in both cases is the same, except that the TSPA method for human intrusion must use prescribed assumptions about the human intrusion scenario (10 CFR 63.322 [66 FR 55732]). The NRC licensing regulations would require compliance with the 15-mrem/yr dose limit for individual protection if the DOE determines that, within the 10,000-year regulatory compliance period, the waste package would degrade sufficiently that a human intrusion could occur without recognition by the drillers (as a result of exploratory drilling for groundwater). If exposure from a human intrusion is projected to occur more than 10,000 years after disposal, the analysis of the human intrusion scenario must be presented in the environmental impact statement, and the dose limits for the human intrusion standard would not apply.

The guidelines for the TSPA method (10 CFR 963.16(b)(1) through (12) [66 FR 57298]) provide for a thorough TSPA, but they are not necessarily set out in the guidelines in the order in which analysts would conduct a TSPA. That is, the guidelines do not dictate the sequence in which a TSPA must be conducted. Figure 1-2 illustrates the sequence of steps in the TSPA method, relating each step in the method to the corresponding postclosure guideline in 10 CFR Part 963. The double arrows indicate that some steps of the evaluation processes are iterative and feed into subsequent evaluations. Section 3.2 presents a discussion of the postclosure evaluation method.

Twelve guidelines for the TSPA criteria are provided at 10 CFR 963.17 (66 FR 57298). Nine guidelines at 10 CFR 963.17(a)(1) through (9) relate to processes and models of the total system performance, and the remaining three guidelines, at 10 CFR 963.17(b)(1) through (3), consider specific disruptive events and processes. Separate evaluations for these guidelines have been completed in support of the TSPA criteria, which are discussed in Section 4.2 of Yucca Mountain Science and Engineering Report (DOE 2002a) and summarized in Section 3.3 of this evaluation.

The nine postclosure suitability criteria (10 CFR 963.17(a)(1) through (9) [see Table 1-2] [66 FR 57298]) consider the conditions expected at the Yucca Mountain site and correspond to the physical processes acting on the natural and engineered components of the geologic repository. These processes also form the basis for the numerical models used in the assessment of total system performance. The remaining three criteria (10 CFR 963.17(b)(1) through (3)) are disruptive events that are generally treated as effects imposed on the system at a particular time that reflect the probability of occurrence of disruptive events. In contrast to the processes considered in the first nine criteria, disruptive events represent comparatively rapid processes that could interrupt or permanently alter the Yucca Mountain disposal system and affect its performance.

The TSPA method begins with empirical observations from sources such as laboratory testing, field investigations, and studies of natural analogues. From these observations, over 120 analysis model reports were prepared, and from them a suite of nine process model reports was developed that summarize the information called for by the postclosure suitability criteria at 10 CFR 963.17 (66 FR 57298). The primary documentary basis for this part of the TSPA method is presented in Section 4.2 of Yucca Mountain Science and Engineering Report (DOE 2002a) and is discussed in more detail in the following process model reports3:

  1. Integrated Site Model Process Model Report (CRWMS M&O 2000f) describes how the output of the three components of the integrated site model—the geologic framework model, the rock properties model, and the mineralogical model—are combined to form the integrated site model. It characterizes the geology of the geologic setting and provides information needed for hydrologic and transport modeling activities during repository design. It also provides a common basis for rock properties and mineralogical information needed for other process models and for performance assessment modeling.

  2. Unsaturated Zone Flow and Transport Model Process Model Report (CRWMS M&O 2000a) summarizes the models that describe the processes affecting the amount of water entering the unsaturated zone above the underground facility that could be available to contact waste in the underground facility. It also describes the potential movement of water through the unsaturated zone below the underground facility, which could transport dissolved radionuclides or radionuclides attached to colloidal particles. The purpose of the model is to describe (1) the spatial and temporal distribution of water flow through the unsaturated zone above the underground facility, (2) the potential for water seepage into the underground openings of the underground facility, (3) the flow and transport characteristics of water flowing from the underground facility to the saturated zone, and (4) the effect of heat on mountain-scale flow and transport processes.

  3. The models summarized in Near Field Environment Process Model Report (CRWMS M&O 2000g) describe the evolution of the near-field environment (the portion of the geologic environment that would be significantly affected by excavation and affected by heat associated with emplaced waste) and how that evolution relates to the amount and chemical composition of water that could contact waste. The processes and impacts modeled include the effects of heat from the waste on the hydrologic properties of the rock mass (such as effects on water flow through the unsaturated zone), the thermodynamic environment, chemical reactions and products, and the mechanical interactions in the near-field host rock surrounding the emplacement drifts.

  4. Engineered Barrier System Degradation, Flow, and Transport Process Model Report (CRWMS M&O 2000h) summarizes the conceptual and mathematical models of the processes that affect the performance of the engineered barrier system. It documents the role of the system in controlling the waste package and drip shield environment, particularly with respect to water flow and geochemical conditions. The process models in this report describe variations in drift water chemistry and gas compositions. They also document the thermal, mechanical, hydrologic, chemical, and radionuclide transport processes arising from the interaction of the emplaced waste with the engineered barrier system and the surrounding host rock.

  5. Waste Package Degradation Process Model Report (CRWMS M&O 2000i) summarizes the models of processes involved in the degradation of the drip shield and waste package, such as the corrosion of the waste package materials in the near-field environment. It provides information about important environmental factors affecting waste package lifetime, including the thermal, hydrologic, and geochemical processes acting on waste package surfaces.

  6. Waste Form Degradation Process Model Report (CRWMS M&O 2000j) summarizes the analysis of the processes of waste form degradation. It also describes those characteristics of the waste forms that may retard the release of radionuclides. Discussions include cladding degradation and waste form dissolution. The report describes the physical and chemical processes of waste form degradation and how the nature of degradation is expected to affect the release of radionuclides to the surrounding environment.

  7. Saturated Zone Flow and Transport Process Model Report (CRWMS M&O 2000k) summarizes the models of flow and transport in the saturated zone and describes the processes that control the movement of water containing dissolved radionuclides or radionuclides attached to colloidal particles through the saturated zone below the underground facility. The processes described retard the movement and reduce the concentrations of those radionuclides that may escape the Yucca Mountain disposal system. The model describes the spatial and temporal distribution of water flow and transported radionuclides through the saturated zone to the point of uptake by the critical exposure group (now called the reasonably maximally exposed individual).

  8. Biosphere Process Model Report (CRWMS M&O 2000l) documents how the models of biosphere processes describe the movement of those radionuclides that could eventually be released from the repository to the human environment. The models also develop radionuclide-specific dose conversion factors used to calculate doses to a member of the receptor group (now called the reasonably maximally exposed individual) living and farming according to the current typical lifestyle of people living in Amargosa Valley, Nevada.

  9. Disruptive Events Process Model Report (CRWMS M&O 2000m) summarizes analyses describing the probability and consequences of volcanic, seismic, and structural deformation events that may affect the postclosure performance of the Yucca Mountain disposal system. It also presents a condensed description of analyses supporting conceptual models and discusses hazard characterizations resulting from expert elicitations.

The process models described in these reports were developed to represent processes and conditions over the temperature range from ambient (25°C [77°F]) to over 200°C (392°F) at the drift wall and back to ambient. The abstractions of these models were initially focused on a higher-temperature operating mode.

After completing the process model reports, the DOE prepared FY01 Supplemental Science and Performance Analyses (BSC 2001a; BSC 2001b), which considers updated information related to guidelines concerning uncertainties (10 CFR 963.16(b)(2) [66 FR 57298]) and the technical basis for the TSPA models (10 CFR 963.16(b)(7)). Volume 1 of FY01 Supplemental Science and Performance Analyses (BSC 2001a) presents the new information, and Section 1.3.1 of Volume 1 identifies uncertainties that were previously treated as point estimates, bounds, or conservatively biased distributions. The uncertainties related to thermal effects on potential repository performance are included. As noted earlier, Volume 1 also examines multiple lines of evidence (e.g., natural analogues, field experiments and observations, calculations) that as part of the technical basis for the TSPA models. Information presented in Volume 1 has been considered in Section 3.3 of this evaluation.

In addition to the reports previously described, Yucca Mountain Site Description (CRWMS M&O 2000c) provides a comprehensive description of the natural system. This report summarizes the results of scientific investigations carried out as part of site characterization, the studies performed at the site and in the surrounding region before formal site characterization began, and relevant information obtained by independent investigators.

The TSPA method identifies the FEPs that will be analyzed by the models used in the assessment of total system performance. FEPs include features observed today, events that could suddenly and randomly occur, and processes that gradually and continuously evolve. The relevant FEPs are identified from a variety of sources, including an international database (NEA 1999) and specific consideration of the Yucca Mountain site by various technical experts. Those identified FEPs are then screened for inclusion in or exclusion from TSPA analyses. FEPs may be excluded if their probability of occurrence falls below one chance in 10,000 over 10,000 years, as provided in 10 CFR 963.16(b)(4) (66 FR 57298) (see Section 3.2.4). FEPs may also be excluded on the basis of their consequences expressed as a change to the calculated annual dose, as provided in 10 CFR 963.16(b)(5) (see Section 3.2.5). For example, criticality events and radiolysis (corrosion) processes were screened out for their low probability or consequences. Otherwise, FEPs are considered in the analyses of system performance. The technical bases for the FEPs included in the TSPA are discussed in Section 3.2.5, and the screening process for inclusion or exclusion of FEPs is more completely documented in a suite of analysis model reports. The results—the identification of FEPs important to isolating waste—are presented in Section 3.2.7. The barriers associated with the FEPs are described in Sections 3.2.8 and 3.2.9.

Because of the volume and diversity of relevant information and the need to perform multiple calculations to address uncertainty and variability in the data fed into the TSPA analyses, not every aspect of every process model is analyzed directly in the assessment of total system performance. Information, including simplified versions of process models, process model results, direct data feeds, and documented assumptions, is incorporated as input into the TSPA models to analyze the performance of a potential Yucca Mountain disposal system (see Section 3.2.1). This input process, known as abstraction, captures the essential aspects of the information from the process models in a manner that the TSPA computer models can analyze. The abstraction process for each of the major process models is discussed further in Section 4.2 of Yucca Mountain Science and Engineering Report (DOE 2002a) and in Section 3.2.6 of this report.

The central step in the TSPA method is the analysis of the TSPA models. This part of the TSPA method is documented primarily in Total System Performance Assessment—Site Recommendation Methods and Assumptions (CRWMS M&O 2000n), Total System Performance Assessment for the Site Recommendation (CRWMS M&O 2000b), FY01 Supplemental Science and Performance Analyses (BSC 2001a; BSC 2001b), and the TSPA Report for Final Environmental Impact Statement and Suitability Evaluation (Williams 2001a). These reports also consider the effects of alternative models (see Section 3.2.3), discuss the sensitivity studies performed (see Section 3.2.11), and address uncertainty and variability in the data and models (see Section 3.2.2).

Additional information regarding the significance and applicability of TSPA results is documented in Repository Safety Strategy: Plan to Prepare the Safety Case to Support Yucca Mountain Site Recommendation and Licensing Considerations (CRWMS M&O 2001a). That information, in general, addresses defense in depth and barrier importance.

Figure 1-2 summarizes the postclosure suitability evaluation process. Section 4.2 provides a summary of the results of the postclosure suitability evaluation. Results are shown in comparison with applicable radiation protection standards.

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