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2. DESCRIPTION OF THE POTENTIAL REPOSITORY

Section 114(a)(1)(A) of the Nuclear Waste Policy Act of 1982 (NWPA), as amended (42 U.S.C. 10134(a)(1)(A)), requires "a description of the proposed repository, including preliminary engineering specifications for the facility." Refining the design and the choice of operating mode for the potential repository is an ongoing, iterative process involving scientists, engineers, and decision-makers. The goal of this process is to develop a design that works with the natural system to enhance containment and isolation of spent nuclear fuel and high-level radioactive waste.

The U.S. Department of Energy (DOE) completed a major milestone in December 1998 with the publication of Viability Assessment of a Repository at Yucca Mountain (DOE 1998). That report included a repository design referred to as the Viability Assessment (VA) design. Based on an improved understanding of the interactions of potential repository features with the natural environment and the addition of engineered features for enhanced waste containment and isolation, the VA design has evolved into the design described in this report. As the program moves forward, the design will continue to evolve.

Figure 2-1 depicts a cutaway view of a proposed layout for the repository facilities. Construction of the emplacement drifts and subsurface facilities would be accomplished in phases. In the current plan, about 10 percent of the emplacement drifts would be completed during the initial construction phase (prior to initiation of waste emplacement), with the remainder of the emplacement drifts being completed during the operation phases. This phased construction would allow the DOE flexibility to develop the repository based on future deliveries of spent nuclear fuel.

The potential repository facilities have been designed to be fully integrated, using a systems engineering approach to identify and then fulfill requirements by providing adequate design solutions. This systems engineering approach to design and development is explained in Section 2.1.1. An important aspect of the design solutions is that they provide the flexibility to accommodate developing operational scenarios, including associated thermal environment characteristics. They provide a basis to refine the design as it evolves in response to increased understanding of the performance of its components. They can also accommodate unanticipated underground conditions that may be encountered during construction with minimal interruptions and no need for expensive retrofits.

The description of the potential geologic repository at Yucca Mountain is organized into five parts. Section 2.1 presents a general overview of the engineering and design process common to all design disciplines in the Yucca Mountain project. It also discusses the evolution of the design, design and operating flexibility, and an assessment of a broader range of thermal operating modes. Design descriptions for the surface and subsurface facilities are covered in Sections 2.2 and 2.3, respectively. Section 2.4 describes the emplacement drift design features that are part of the engineered barriers. Although the definition of engineered barriers includes the waste package, the descriptions of waste package designs and the different waste forms that would be contained within waste packages are provided separately in Section 3. Section 2.5 describes the surface and subsurface facilities that support the performance confirmation program for the potential repository.

The design and operating mode that was analyzed to assess long-term performance of the repository system presented in Total System Performance Assessment for the Site Recommendation (CRWMS M&O 2000a) was based on a higher-temperature operating mode. Subsequent analyses that considered the performance of lower-temperature operating modes are described in FY01 Supplemental Science and Performance Analyses (BSC 2001a; BSC 2001b).

2.1 ENGINEERING AND DESIGN ANALYSIS

Design development for the potential repository follows a structured approach that links statutory, regulatory, and derived requirements to the final design products. Design work has been performed in accordance with a quality assurance program (DOE 2000a). This program has been reviewed and accepted by the U.S. Nuclear Regulatory Commission (NRC). Figures 2-2, 2-3, and 2-4 discussed in the following section, illustrate the steps in the design process.

2.1.1 Design Process

This section describes the design process, including how requirements are identified and passed down to individual systems through an established document hierarchical system and how systems are analyzed and then classified according to their importance to preclosure radiological safety.

2.1.1.1 Allocation of Yucca Mountain Site Characterization Project Requirements

The framework for the design of the major repository structures, systems, and components is consistent with the following:

 The DOE has published a comprehensive hierarchy of design documents for a monitored geologic repository. The primary document is the Civilian Radioactive Waste Management System Requirements Document (DOE 2000b). Requirements in that document applicable to the repository site are allocated to the Monitored Geologic Repository Requirements Document (YMP 2000a). The hierarchy then branches into specific areas of scope, becoming more detailed with each level of document, from the Monitored Geologic Repository Project Description Document (Curry 2001) to a set of System Description Documents. The Monitored Geologic Repository Requirements Document (YMP 2000a) captures top-level functions and requirements. Monitored Geologic Repository Project Description Document (Curry 2001) documents the functions, requirements, criteria, and assumptions for a potential repository while allocating each to the appropriate systems, as detailed in the System Description Documents. For instance, the guidance for blending spent nuclear fuel assemblies to achieve a maximum thermal output of 11.8 kW per waste package at the time of emplacement is contained in the Monitored Geologic Repository Project Description Document (Curry 2001). This document also includes controlled assumptions and captures performance criteria and design constraints. In this way, it supplements the higher-level approach of the Monitored Geologic Repository Requirements Document (YMP 2000a). Figure 2-2 illustrates this transfer of requirements from higher-level to lower-level documents, as well as the allocation of functions, requirements, and criteria by the Monitored Geologic Repository Project Description Document (Curry 2001) to individual systems.

The requirements documents also include references to codes and standards applicable to specific structures, systems, and components. These codes and standards are generally developed by professional organizations, such as the American Society of Mechanical Engineers or the American Nuclear Society, in cooperation with the American National Standards Institute.

2.1.1.2 Safety Classification of Structures, Systems, and Components

The design components of the repository system would contribute to performance in varying degrees. Certain design components of the system are more important to safety of preclosure operations, as described in Preliminary Preclosure Safety Assessment for Monitored Geologic Repository Site Recommendation (BSC 2001f). Other design components are more important to postclosure performance, as shown by performance assessment sensitivity studies (e.g., Section 4.4.5). Considerations of safety margin and defense in depth, insights from analogues, and expert judgments also help identify the importance of repository components to performance (see Section 4.1). Knowledge of how components of the system affect performance can provide insights on how design or operating mode features could be developed in a manner that could contribute to long-term performance or mitigate potentially adverse conditions.

The definitions of design criteria and requirements are influenced by a consideration of the importance of each system, its structures, and its components in the overall safety strategy for the potential repository. That safety strategy has been developed over the years based on determinations of critical factors in design with respect to preclosure safety and postclosure performance. The structures, systems, and components important to preclosure safety are identified through engineering analyses and relate directly to the health and safety of facility workers, the health and safety of the public, and the environment. The factors important to postclosure performance are determined through evaluations of the importance of the components to overall system performance. These evaluations include total system performance assessment (TSPA), considerations of safety margin and defense in depth, and independent, multiple lines of evidence. These evaluations integrate the performance of natural barriers (i.e., the geologic environment) and man-made barriers (e.g., the drip shield and waste package outer barrier) with respect to their complementary attributes in containing and isolating radioactive waste over long periods of time.

For the preclosure period, the importance of design features is defined in terms of their role in preventing or controlling radiological exposure to repository workers and the public. The design of these features must address the health and safety requirements associated with radiological work. The more important a design feature is to ensuring radiological safety, the more process controls are imposed on that design. To the maximum extent practicable, the design of potential repository structures, systems, and components has been developed from the design of structures, systems, and components already in use at other licensed nuclear facilities. The standards used in the designs of such facilities are well developed.

Following the process summarized in Figure 2-3, structures, systems, and components are classified to define their importance to preclosure safety. Event sequences form the basis of these safety classifications. This process is integrated in the sense that all structures, systems, and components important to safety are analyzed for event sequences that represent a complete set of bounding conditions. This set of events results from scenario analysis and grouping of the internal and external hazards. The preclosure safety assessment integrates safety evaluations through a joint consideration of safety measures that otherwise might conflict, including but not limited to integration of fire protection, radiation safety, criticality safety, and chemical safety measures. Event sequences include natural or human-induced events that are reasonably likely to occur and that could lead to exposure of individuals to radiation. Event sequences also include other natural or human-induced events that are unlikely but sufficiently probable to warrant consideration (i.e., that have at least 1 chance in 10,000 of occurring before permanent closure of the repository).

Event sequences could be internal (e.g., collision, loss of power, or fire) or external (e.g., earthquake, tornado, or flood). A potential bounding event sequence (i.e., the one resulting in the worst failure) is identified for analysis of a particular structure, system, or component. The frequency of occurrence of that event sequence determines its event credibility (probability) and category. An engineering analysis is then performed to determine whether a radiological release could result from the failure of a structure, system, or component because of a credible event. These analyses are performed with mathematical and analytical models of processes and events. The consequences of the release are evaluated to determine whether the resulting doses are within regulatory limits and, if not, what preventive or mitigating measures are required to bring the radiological consequences within compliance limits. Structures, systems, and components required to meet regulatory limits are classified as important to safety. Once the safety classifications are defined, individual systems are ranked based on the importance to safety of the performance of their structures and components. The process depicted in Figure 2-3 is iterative in nature and results in a safety analysis that is integrally tied to the facility design (BSC 2001f, Section 5.2).

This report focuses on the most important systems, those that could adversely affect worker and public safety if they failed. To this end, potential systems have been evaluated for their importance to safety and classified into four groups of quality levels, as defined in the Preliminary Preclosure Safety Assessment for Monitored Geologic Repository Site Recommendation (BSC 2001f, Section 4.4.1):

The preclosure safety analysis process is shown in Figure 2-3. Event sequences are classified as Category 1 or Category 2, based on the frequency of the entire event sequence (also known as the scenario frequency). The frequency ranges for each event sequence category, given in Table 2-1, correlate with the probability-based definitions from 10 CFR 63.2 (66 FR 55732), assuming for the purpose of analysis a period of 100 years before repository closure (BSC 2001f, Section 5.2).

2.1.1.3 System Description Documents

Monitored Geologic Repository Project Description Document (Curry 2001, Section 4) captures, by logical groupings, the hierarchical arrangement of the repository design documents. In that hierarchy, the repository is divided into three major systems, each with several secondary systems, as follows:

The secondary systems are discussed further in this section, since they relate to the surface and subsurface design descriptions. The complete arrangement for all the potential repository systems is described in Section 4 of Monitored Geologic Repository Project Description Document (Curry 2001).

Each repository system would eventually have a complete System Description Document. Figure 2-4 illustrates the relationship of a System Description Document to the design process. Section 1 of a System Description Document defines the system functions, criteria, requirements, constraints, and interface requirements with other systems, as applicable to that system's structures and components.

Given this set of information, the design engineer develops design concepts that fulfill the system needs; those concepts are subjected to additional evaluations, including technical feasibility, functionality, constructibility, cost, and operations and maintenance efficiency analyses. These analyses identify the selected design option. Selected design options may be subjected to further scrutiny such as ALARA evaluations for radiological exposures.

The design work is documented in calculations, technical reports, and design analyses that are submitted for independent reviewers to check and for impacted organizations to review, following a rigorous NRC-approved quality assurance program. Design documents are revised to incorporate checking and review comments, then rechecked for conformance to checking and design review comment resolutions before approval.

Section 2 of a System Description Document summarizes the design information contained in the approved design documents with a description of the system, its structures, and its components. Section 2 of each System Description Document also includes a demonstration of how the system criteria and requirements are fulfilled by the selected design for systems important to radiological safety.

2.1.2 Design and Operational Mode Evolution

Refining the design for the potential repository and the mode in which the design is operated has been an ongoing, iterative process involving scientists, engineers, and decision-makers. The design and mode of operations of a potential Yucca Mountain repository has evolved as more has been learned about the site and the performance contribution of design attributes and operational objectives.

Previous studies have investigated repository operating modes, layout, and performance considerations for a range of thermal conditions. Mansure and Ortiz (1984) developed evaluations of areas that could be used for expansion of the repository conceptual design footprint based on uncertainty of rock characteristics. This information was used in Environmental Assessment Yucca Mountain Site, Nevada Research and Development Area, Nevada (DOE 1986c) and the Site Characterization Plan Yucca Mountain Site, Nevada Research and Development Area, Nevada (DOE 1988). Building on this information, alternative design concepts were presented in Viability Assessment of a Repository at Yucca Mountain (DOE 1998, Volume 2, Section 8.3.2). Viability Assessment of a Repository at Yucca Mountain (DOE 1998) described the evolution of the repository design from the Site Characterization Plan Conceptual Design Report (SNL 1987) to the publication of the Viability Assessment. This section briefly discusses the evolution of the repository design and the range of thermal operations, from the publication of the Viability Assessment in December 1998 to the present. Descriptions of the design and mode of operations used in performance evaluations are covered in more detail in Sections 2.2 through 2.5 [2.2, 2.3, 2.4, 2.5] and in Section 3 to assist in understanding the design.

The DOE initiated an effort to evaluate a range of alternative design concepts and operational performance objectives during the License Application Design Selection process. This process culminated with the selection of what was referred to as the "license application design" (CRWMS M&O 1999c) and the subsequent analyses and designs for the selected design alternative, Enhanced Design Alternative II (CRWMS M&O 1999d). The Enhanced Design Alternative II concept was modified to remove backfill from the emplacement drifts and add operational performance objectives for thermal loading and ventilation (Wilkins and Heath 1999; Stroupe 2000). This design selection process considered a range of thermal conditions. The Enhanced Design Alternative II design concept and mode of operations provide a moderate repository environment compared to the design described in the Viability Assessment. The Enhanced Design Alternative II design would keep the spent nuclear fuel cladding temperature below 350°C (662°F), a design requirement established to maintain the integrity of the cladding, and keep the boiling fronts from coalescing in the rock pillars between the emplacement drifts. To achieve the operational objective of keeping the boiling fronts from coalescing in the rock pillars, the ventilation rate in the emplacement drifts was increased, which allowed the spacing between waste packages to be reduced.

The design of structures, systems, and components and the methods used to operate and maintain these structures, systems, and components are distinct but interrelated in their developmental and evolutionary processes. Sensitivity analyses have been performed to identify design features and operational objectives that contribute to the isolation of waste or reduce uncertainty in performance analyses. In the design evaluation process described above, the design features and operational objectives were evaluated together to identify the combination of design features and the range of operational objectives that might enhance the performance of the potential repository system.

The draft environmental impact statement (EIS) for the potential Yucca Mountain repository recognized the potential for operating modes spanning a range of lower temperatures and presented results of analyses of the impacts associated with the range of thermal loadings (DOE 1999a, Appendix I, Section I.4.2). The conceptual layouts analyzed for the draft EIS for the different thermal load scenarios and inventories were developed considering the TSPA model domains and a number of rock areas within which the emplacement drifts would be located. The repository and emplacement areas for the higher- and lower-temperature operating modes with a 70,000-MTHM inventory are illustrated in Figures 2-5 and 2-6. Repository layouts for the draft EIS high, intermediate, and low thermal load scenarios are illustrated in Figure 2-7. The figure also illustrates, for comparative purposes, the layout for the higher-temperature operating mode described in this report.

A more detailed analysis of the lower-temperature operating modes resulted in a repository layout (see Figure 2-10) capable of accommodating the 70,000 MTHM and 97,000 MTHM inventories within the primary and lower blocks (BSC 2001g). This analysis also included parametric studies and layout sensitivity analyses for a wide range of repository thermal operating modes. The postclosure performance assessment for this new layout was discussed in Volume 2 of FY01 Supplemental Science and Performance Analyses (BSC 2001b). The layout is also described in the final EIS.

The various components of the repository system would contribute to performance in different ways. Certain components of the system are more important to performance than others, as shown by performance assessment sensitivity studies (e.g., Section 4.4.5). Considerations of safety margin and defense in depth, insights from analogues, and expert judgments also help identify the importance of repository components to performance. Knowledge of how components of the system affect performance can provide insights on how design features or the operating mode could be developed in a manner that contributes to long-term performance or could mitigate potentially adverse conditions.

2.1.2.1 Summary of Evolution of Design Features

From the issuance of Viability Assessment of a Repository at Yucca Mountain (DOE 1998) to the present, some design features in surface, subsurface, and engineered barrier systems and in the waste package have evolved. This section summarizes the evolution of these features.

Surface Facilities—The design of the surface facilities has evolved in two areas since the publication of the Viability Assessment (DOE 1998). The design of the Waste Handling Building has evolved to include an expanded-capacity spent nuclear fuel blending pool. The number of assembly and canister transfer lines was reduced to two assembly transfer lines and one canister transfer line; the design described in the Viability Assessment included three assembly transfer lines and two canister transfer lines. The fuel blending strategy is presented in Section 2.2.1. Also, a solar power generating facility was added to supplement power from the site, which gets its power from the southern Nevada grid.

Subsurface Facilities and the Emplacement Drift Portion of the Engineered Barriers—The evolution of the subsurface facilities design has introduced several changes since the VA design, mainly to permit operating the repository to accommodate a range of thermal conditions. The most significant design change relates to thermal loading (the allowable amount of introduced heat per unit of subsurface emplacement area). The higher-temperature operating mode design described in this report uses a spacing between emplacement drifts of approximately 81 m (266 ft), compared to a spacing of approximately 28 m (92 ft) in the VA design (DOE 1998). Operational parameters, such as waste package spacing and ventilation, can be varied to support flexible thermal goals (see Section 2.1.2). In addition, the design and operating mode used in the performance analyses achieves a more uniform distribution of rock temperatures along the drift, limiting potentially complex thermal-mechanical effects resulting from a varying thermal gradient along the drift axis.

The emplacement drifts were reoriented to increase drift stability. Based on the dominant rock mass joint orientation and a minimum offset criterion of 30° between the joint strike and drift orientation, a reorientation of the emplacement drifts to an azimuth of 252º was established for the design described in this report (BSC 2001d, Section 6.2.2.2).

Emplacement drift backfill and drip shields were added to further limit the possibility of water contacting the waste packages. The backfill concept also offers the waste package protection against rockfall (CRWMS M&O 1999c, Section 4.4.2). The backfill concept was later eliminated because of its potential adverse impact on spent nuclear fuel cladding temperatures. The backfill inhibits heat dissipation into the rock, resulting in higher waste package temperatures. Granular backfill placed around the waste packages as a layered system could divert moisture away from the waste package by capillary action (CRWMS M&O 1999c, Section 4.4.2). The drip shield is a metal structure placed over the waste package that serves as a diversion shield for water entering the drift from the rock strata above. The drip shield design concept is described in Section 2.4.

The ground control design concept for emplacement drifts was changed from precast concrete liners with a concrete invert to a combination of steel or ballast invert with steel sets and welded wire fabric, with and without grouted rock bolts. Concerns with the long-term impact of concrete on the alkalinity of the drift environment, along with its implications for corrosion of the engineered barrier and waste package components and for the possible enhancement of colloidal radionuclide transport, motivated this change in the design for ground control in the emplacement drifts. Section 2.3.4.1 describes the ground support system.

The evolution to line-loading of waste packages resulted in modifications in the design and function of the waste package support assembly, from two independent post supports in the VA design to a more substantial emplacement pallet for the entire waste package. This necessitated a change in transporter and emplacement gantry design. The support design simplifies and enhances the reliability of transporting and emplacing waste packages.

Waste Package—Two waste package design attributes have evolved since the publication of the Viability Assessment. The first involves the metallurgy of the inner and outer shells of the waste package. The design described in the Viability Assessment utilized a corrosion-resistant Alloy 22 inner shell and a structurally strong, carbon-steel outer shell. The design described in this report utilizes a corrosion-resistant Alloy 22 outer shell and a structurally strong, stainless steel inner shell. In addition to the enhanced shell design, the waste package has a modified top lid design. A third lid has been added and the lid design has been modified to accommodate stress mitigation techniques in the closure weld area.

2.1.2.2 Summary of Evolution of Operational Parameters

This section presents a summary of the evolution of the operational mode from issuance of Viability Assessment of a Repository at Yucca Mountain (DOE 1998) to this report for a range of thermal conditions.

Emplacement Drift Ventilation—The TSPA described in Viability Assessment of a Repository at Yucca Mountain (DOE 1998, Volume 3) utilized a mode of operations where the emplacement drift ventilation rate was 0.1 m3/s (3.5 ft3/s) after waste package emplacement. This low ventilation rate was not designed to remove heat from the emplacement drifts. This operating mode had temperatures in the host rock throughout the emplacement area above the boiling temperature of water. This zone of above-boiling temperatures extended into the host rock up to approximately 100 m (330 ft) from the emplacement drifts.

To keep the boiling fronts from coalescing in the rock pillars between the emplacement drifts, the ventilation system has to remove approximately 70 percent of the heat generated by the waste packages. Meeting this requirement demands ventilation of the emplacement drifts at an estimated rate of 15 m3/s (530 ft3/s) (see Section 2.3.4.3).

Waste Package Spacing—The spacing between the waste packages that was used in the VA evaluations varied depending on both the thermal environment in the emplacement drift and the time-dependent heat output for each individual waste package. The concept behind the spacing was to maintain a fairly uniform linear thermal loading along the length of the emplacement drifts. Since the thermal output of the waste packages varied with the waste form and the age of the waste enclosed, the spacing between the waste packages varied from approximately 1.3 to 9.3 m (4.3 to 30.5 ft). With this spacing, the waste packages could act as point sources for heat in the emplacement drifts.

With the requirement for the ventilation system to remove approximately 70 percent of the heat generated by the waste packages, the concept of uniform linear thermal loading evolved into a line-loading concept for waste package emplacement. Line loading, that is, placing the waste packages end to end, with a separation of 10 cm (4 in.), achieves a more uniform thermal profile along the length of the emplacement drifts with the ventilation conditions described. The concentration of heat sources resulting from line loading required that the emplacement drifts be spaced farther apart to distribute the thermal load over a larger area.

2.1.2.3 Design and Operating Mode Evolution

The repository design concept described in this document is a flexible design that can be operated over a range of thermal conditions. The future evolution of the design and operating mode will follow a process that includes (1) refining specific design requirements and performance goals to recognize performance-related benefits that could be realized through design and (2) enhancing components of the design to best achieve the performance-related benefits. The iterative design process has focused on improving the understanding of the contribution of design features to the performance of a potential repository. The emphasis of the approach using a flexible design and modes of operation for a potential repository at Yucca Mountain is to understand the impact that a design attribute or operational performance objective has on the performance of the site across a range of environmental conditions. This approach examines the sensitivity of a design parameter to a range of operating temperatures and environmental conditions, and it evaluates the performance and uncertainties associated with the temperature variation. If a design attribute is shown to have a significant impact on the performance of the repository, then the attribute undergoes further evaluation to fully develop the positive contribution or minimize the negative contribution to the performance of the repository.

Sensitivity analyses that consider the performance of a design over a range of thermal operating modes provide information that will guide the future development of the repository design. The evolution of the design will take advantage of insights gained through the performance analyses. Furthermore, if the performance evaluations indicate benefits to be gained by refinement of the basic design concept on which the performance analyses over the range of operating modes were based, the evolution of the design will take advantage of those insights as well. This flexible design, which will continue to evolve for license application if the site is recommended, will complement a set of operational parameters that can be managed to accommodate thermal characteristics of a waste stream with potentially evolving characteristics. Adjustments can continue based on updated information on waste stream and other repository variables during the very long emplacement period.

The DOE has expanded the assessment of performance of the repository design to better understand how to use reductions in uncertainty to improve the design. The DOE is further expanding the range of thermal operating modes by assessing the performance of a potential repository operated to achieve lower temperatures. The TSPA described in Section 4 of this report utilizes an operating mode where the walls of the emplacement drifts are above the boiling temperature of water. Results of preliminary engineering evaluations show that the design can also be operated in a lower-temperature mode (CRWMS M&O 2000k). The results of these preliminary evaluations have been further corroborated with more detailed calculations and analyses presented in the following documents:

The effect of heat on the performance of the repository and the associated uncertainties are subjects of ongoing studies. These studies are considering ranges of drift wall temperatures from as high as 200°C (390°F) to below the boiling point of water (96°C [205°F] at the elevation of the emplacement horizon). The performance assessment in the Viability Assessment (DOE 1998, Volume 3) was based on a design that allowed drift wall temperatures to exceed 200°C (390°F). An objective of the operating mode described in this report is to maintain temperatures in a portion of the rock between the emplacement drifts below the boiling point of water. More recent design studies include sensitivity analyses that evaluate repository performance limiting all drift wall temperatures below the boiling point of water (BSC 2001g; BSC 2001h; BSC 2001i; BSC 2001j; CRWMS M&O 2000l). Lower-temperature operating modes to reduce uncertainty about corrosion rates associated with waste package performance have been evaluated in FY01 Supplemental Science and Performance Analyses (BSC 2001a; BSC 2001b). These evaluations have considered waste package temperatures as low as 85°C (185°F).

The evaluation of the performance characteristics of the proposed repository design over a range of operating conditions is an important step in the evolution of the design of a potential repository at Yucca Mountain. While certain details of design are needed to understand the environmental conditions as input to the performance analyses, future evaluations will recognize and be based on the performance implications of the design and accompanying range of operating modes.

2.1.3 Design Flexibility

The design described in this report is part of an overall waste management system that is flexible, continues to evolve, and can adapt to various construction and operational conditions.

In this report, the phrase "flexibility in design" refers to capabilities inherent in the repository design to accommodate changing conditions, such as unanticipated underground conditions and new design requirements. The need for flexibility in the repository design evolves from operational requirements, such as:

 Flexibility in the design allows this evolutionary process to continue and to take advantage of the interrelationship between the design and operating mode. Because the specific thermal criteria (e.g., the waste package and drift wall temperature limits) that will be imposed on future design enhancements have not been selected, the repository design must be sufficiently flexible to allow operation under a wide range of thermal conditions. Assessment of performance over a range of thermal conditions will support the definition of specific thermal criteria upon which future design enhancements will be based. There is a possibility that spent nuclear fuel with higher burnup than presently projected would be received. This could lead to a higher integrated heat over the first 1,000 years of emplacement and could change the predicted temperature profile. The repository would have the flexibility to accept spent nuclear fuel with higher burnup rates.

Key aspects of design flexibility are (1) the ability of the repository design to support a range of construction approaches; (2) the capability to dispose of a wide range of waste container sizes; (3) the ability to support a range of thermal operating modes; and (4) the ability to continue to enhance the design to best achieve performance-related benefits identified through ongoing analyses. The following discussion describes the design's flexibility to accommodate sequential and modular construction of surface and subsurface facilities. Section 2.1.4 describes the design's flexibility to support a range of thermal operating modes.

Sequential and Modular Repository Development—The design of the repository can accommodate modular or sequential construction of surface and subsurface facilities. The phrase "modular or sequential implementation" describes a process in which decisions concerning repository development are made in a stepwise manner: at each step in the process, a decision whether to proceed would be made based on the licensing and regulatory requirements, the funding profile, and operating experience. The next stage of construction would proceed informed by the experience gained from the previous stage.

The DOE has assessed possible benefits of a modular or stepwise approach, which range from the incorporation of lessons learned after each stage of construction to the leveling of annual construction costs (CRWMS M&O 1998a). However, the potential benefits and impacts from modular and sequential construction have not been fully assessed. The DOE has requested that the National Research Council continue the study of possible repository development strategies (Itkin 2000). A report on the results of this study will be provided at a later date, and the DOE will continue to assess this concept for construction activities.

Sequential and modular construction would not be expected to change the design or operational concepts of a potential repository at Yucca Mountain and would not, therefore, impact an evaluation of repository performance.

2.1.4 Operating Flexibility to Achieve a Range of Thermal Operating Modes

Preliminary engineering evaluations, including Operating a Below-Boiling Repository: Demonstration of Concept (CRWMS M&O 2000k) and Natural Ventilation Study: Demonstration of Concept (CRWMS M&O 2000m), show that the design described in this report can be operated to support a range of thermal operating modes. For example, the potential repository could be operated in a mode that keeps temperatures below the boiling point of water (96°C [205°F] at the repository elevation); or the potential repository could be operated such that the host rock reaches a temperature that is above the boiling point of water. Drift wall and waste package temperatures and relative humidity can be managed by altering several operational features of the design: (1) varying the thermal load to the repository by managing the thermal output of the waste packages; (2) managing the period and rate of drift ventilation prior to repository closure; and (3) varying the distance between waste packages in emplacement drifts (CRWMS M&O 2000k, Section 3.2). These factors are described in the following paragraphs. Other parameters, such as postemplacement natural ventilation, could also be used to reduce long-term repository temperatures (CRWMS M&O 2000m). Altering design features, such as emplacement drift spacing, could also be used in conjunction with variations in operational parameters to achieve a lower-temperature repository environment.

Recently completed calculations and analyses (BSC 2001j; CRWMS M&O 2000l; BSC 2001h; BSC 2001i; BSC 2001g) support the preliminary findings on the feasibility of operating the repository to meet varying thermal goals by adjusting the three operational features listed above (i.e., controlling the thermal output of the waste packages, adjusting the period of ventilation, or varying the emplacement distance between waste packages).

Since completion of the TSPA-SR analyses, the DOE has performed additional analyses supporting the effects of a range of thermal operating modes on projected system performance. These analyses were described in FY01 Supplemental Science and Performance Analyses (BSC 2001a; BSC 2001b). Evaluations of higher- and lower-temperature operating modes in the supplemental TSPA model have provided insight into system performance at the subsystem level over a range of thermal operating modes. While some subsystem models indicated significant differences, results from changing the thermal operating mode showed only a minor impact on overall repository performance. The supplemental model also showed that the repository could be operated at either a higher- or lower-temperature operating mode with a low probability of the development of aggressive chemistries on the waste package and the subsequent potential for localized corrosion.

If the site is recommended for development of a repository, the DOE will utilize the enhanced understanding of repository performance to further improve the repository design for any license application.

Thermal Output of the Waste Packages—The major contributor of heat in the repository would be commercial spent nuclear fuel, which would likely have a wide range of thermal outputs. The thermal load of a repository is directly related to the amount of thermal energy contained in the fuel emplaced in it. The thermal energy contained in the fuel, in turn, is directly related to its age. The age and burnup rate of spent nuclear fuel received for emplacement in a repository would vary considerably, so the current operational plan for the potential repository specifies that the DOE will manage the fuel inventory by one or more of the following features: (1) fuel blending (i.e., placing low heat output fuel with high heat output fuel within a waste package); (2) de-rating (i.e., limiting the number of spent fuel assemblies to less than the waste package design capacity); (3) placing high heat output fuel in smaller waste packages; or (4) aging (i.e., placing hotter fuel into the fuel blending inventory to be emplaced later). Managing the average thermal output of the waste packages through any of these means can control drift wall temperatures of the repository (CRWMS M&O 2000k, Section 3.2; BSC 2001a; BSC 2001b).

Duration and Rate of Forced Ventilation—During active repository operations, some of the heat generated by the waste and the moisture in the surrounding rock would be removed from the repository by forced ventilation of the loaded emplacement drifts. The amount of energy transferred from the waste to the host rock can be managed by varying the duration and the rate of emplacement drift ventilation (CRWMS M&O 2000k, Section 3.2.3).

Distance Between Waste Packages—The distance between waste packages in emplacement drifts is another design feature that can be modified to manage the temperature in the potential repository. As waste packages are spaced farther apart, the linear thermal density in the drift (measured in kilowatts of heat output per meter of drift length) decreases, delivering less heat per unit volume of the host rock when the drift-to-drift spacing remains fixed (CRWMS M&O 2000k, Section 3.2.2).

Natural Ventilation—Postemplacement natural ventilation could be employed in several lower-temperature operating scenarios. The subsurface ventilation system described in Section 2.3 could support both forced and natural ventilation. To facilitate natural ventilation, the ventilation system would be enhanced through a combination of air balancing techniques, such as adjusting the size of the openings, location and number of intake/exhaust openings, and flow controls (CRWMS M&O 2000m).

2.1.5 Assessing The Performance of a Lower-Temperature Operating Mode

The basic operational features to achieve a lower-temperature repository were described in Operating a Below-Boiling Repository: Demonstration of Concept (CRWMS M&O 2000k). That report outlines the relationships between operational parameters such as ventilation and linear thermal load. In subsequent design evaluations (BSC 2001j; CRWMS M&O 2000l; BSC 2001h; BSC 2001i; BSC 2001g), the DOE has used a range of lower-temperature operating mode characteristics to investigate the performance attributes of possible alternative lower-temperature designs and operating modes.

This section describes features of some of these lower-temperature operating modes. These features have been used to assess the performance of a potential lower-temperature repository and the sensitivities of projected performance to different parameters and design features (BSC 2001b). Evaluations of a range of operating temperatures will support the development of a design and operating mode to support any license application. Features of lower-temperature operating modes have also been assessed to support the development of the ranges of design-related environmental conditions required as input to the TSPA analyses.

Design and performance evaluations could include varying the separation between the emplacement drifts (drift-to-drift spacing) to adjust the areal mass loading; increasing the number and diameter of ventilation shafts to improve efficiency; or zone emplacement, which tailors ventilation and design attributes to the thermal output of waste packages emplaced in specific drifts (or zones). The DOE will also examine strategies to reduce ventilation times and meet thermal goals for a lower-temperature operating mode. This could involve lengthening the time to emplace all of the wastes, further reducing areal thermal densities, rearranging emplaced fuel, or combining the aging of fuel with enhanced fuel blending inventory strategies.

Lower-temperature operating modes are being considered for mitigating some of the potential uncertainties in assessing long-term repository performance. There are many ways of combining operational parameters to achieve lower repository temperatures. Figure 2-8 illustrates the variables affecting the thermal performance of the repository, from waste forms to emplacement drifts. Within the constraints imposed by the physical system, the potential repository can be operated in lower-temperature modes while also meeting other technical, policy, regulatory, schedule, and operational objectives. Some of these objectives are described in the following paragraphs.

Reduced Uncertainty in Corrosion Rates—The corrosion susceptibility of Alloy 22 may be reduced by either keeping waste package surface temperatures at or below 85°C (185°F) or maintaining in-drift relative humidity below 50 percent (see Figure 2-9) (Dunn et al. 1999, p. xvi; CRWMS M&O 2000n, Section 3.1.3.1). Operating the repository such that the combination of in-drift temperature and relative humidity does not enter the window of corrosion susceptibility of Alloy 22 may result in better overall performance.

Keeping Drift Wall Temperatures Below Boiling—In the higher-temperature operating mode, the rock temperatures within the first several meters around the emplacement drifts exceed the boiling point of water. A possible lower-temperature objective would be to keep all the rock in the repository below the boiling point of water to reduce uncertainties associated with thermal-hydrologic and thermal-mechanical processes.

Capacity for Waste Inventory—Potential objectives are (1) to ensure that 70,000 MTHM can be emplaced within the characterized area and (2) to maintain the flexibility to accommodate up to 119,000 MTHM within the repository area. However, the smaller the space that is used for inventory, the more reliance must be placed on other means of meeting lower-temperature goals.

Duration of the Ventilation Period—Potential objectives are (1) to provide for closure and sealing of the repository within a prescribed time limit determined by the Secretary of Energy or (2) to allow for natural ventilation to continue after permanent closure of the repository. Allowing some dependence on extended ventilation reduces the area required for disposal of a given amount of waste. However, a design based on centuries of ventilation through air passages connecting the repository to the surface could involve technical issues that would have to be explored.

Thermal Output of the Waste Package—The thermal output of the waste packages can be managed by combining one or more of the following features: (1) fuel blending to produce a thermal loading at or below 11.8 kW per waste package; (2) de-rating the waste packages to produce a thermal loading below 11.8 kW per waste package; (3) placing high heat output fuel in smaller waste packages to produce a thermal loading below 11.8 kW per waste package; or (4) aging spent nuclear fuel assemblies to reduce their heat output.

2.1.5.1 Lower-Temperature Operating Mode—Coupled Operational Parameters

Figure 2-10 illustrates a layout of potential development areas (or blocks) within the characterized area of Yucca Mountain that may be utilized for the emplacement of waste. The shaded areas denote portions of the upper and lower blocks that could be utilized for the emplacement of waste under different operating modes, ranging from higher-temperature to lower-temperature cases corresponding to the examples given in Table 2-2.

The TSPA-SR evaluated the performance of a higher-temperature operating mode. In this mode, as rock temperatures increase above the boiling point of water, moisture around the emplacement drifts evaporates and is driven away from the drifts as water vapor. As described in Section 4.2.2, this may have performance benefits because it would delay the time at which liquid water could begin to seep into the emplacement drifts. However, there are heat-related uncertainties about the long-term impact of thermal, chemical, and mechanical effects on the hydrologic properties of the potential repository host rock. Maintaining the repository at lower temperatures (below the boiling point of water) may reduce these uncertainties. Additional analyses of lower-temperature operating modes were performed and are presented in FY01 Supplemental Science and Performance Analyses (BSC 2001a; BSC 2001b).

To demonstrate how the DOE can vary operational parameters of the design to manage the thermal load of the repository, a set of preliminary calculations was performed to assess the effects on repository temperatures of the parameters described in the previous section: (1) the age of the spent fuel emplaced; (2) the duration of ventilation; and (3) the distance between waste packages (CRWMS M&O 2000k; CRWMS M&O 2000m).

The family of operational parameters that would produce average maximum drift wall temperatures below the boiling temperature of water (96°C [205°F]) is shown in Figure 2-11. This figure plots these parametric curves using postloading ventilation duration as the horizontal axis and distance between waste packages as the vertical axis. Each of the labeled curves in this plot represents a different duration of additional aging beyond the average age of the waste stream in the year it is received. The curve with 0 years of aging corresponds to the case where the fuel is emplaced at the rate it is received, without additional aging. The curve for 5 years of aging corresponds to the case where the waste is allowed, on average, an additional 5 years of aging.

An example of how these curves are used can be shown by examining the point on the 10-year aging curve for 50 years of ventilation and a waste package spacing of about 2.3 m (7.5 ft). For a drift operated this way, the postclosure emplacement drift wall would reach a peak temperature of just at or below 96°C (205°F). If the drift were to receive fewer years of ventilation, keeping the age of the fuel loaded in the drift and the waste package spacing constant (to the left of the point in question), then the average maximum temperature of the host rock walls would increase above the boiling point of water during the postclosure period. On the other hand, if an emplacement drift were ventilated longer (to the right of the point in question), then the peak temperatures in the emplacement drift rock would be farther below the boiling point of water.

The trade-offs between the operational parameters that can be changed to achieve a below-boiling repository operating mode can be determined from Figure 2-11. For example, if an emplacement drift is loaded with waste packages spaced around 2.3 m (7.5 ft) apart without fuel aging, 75 years of post-loading ventilation is required. If the age of the waste stream is increased by 10 years, a waste package spacing of 2.3 m (7.5 ft) can be used with only 50 years of ventilation. For this case, aging the waste for 10 years saves 25 years of ventilation. Increasing the duration of aging, however, yields diminishing benefits. As the waste ages appreciably, the rate of decay in heat generation diminishes greatly. Consequently, whether the waste is aged further on the surface or underground in the potential repository makes less and less difference in keeping the drift wall below boiling. The flattening shape of the curves in Figure 2-11 beyond 75 years of ventilation also shows that increasingly longer ventilation periods would be required to remove the same quantity of decay heat from the repository.

2.1.5.2 Example Lower-Temperature Operating Scenarios

Lower-temperature operating modes have been developed to provide options for mitigating some of the potential uncertainties in assessing long-term repository performance. Analyses of the potential performance of lower-temperature operating modes are described in FY01 Supplemental Science and Performance Analyses (BSC 2001a; BSC 2001b).

This section describes several possible operating scenarios, each of which meets the primary goals for lower-temperature operating modes described in Section 2.1.5 while achieving one set of possible operating objectives. Table 2-2 compares operational parameters used in the TSPA-SR with operational parameters that could be used to achieve a lower-temperature operating mode.

The following sections compare five examples of lower-temperature operating scenarios. These examples provide a basis for understanding the technical issues associated with developing a lower-temperature operating mode.

2.1.5.2.1 Lower Waste Package Temperature Achieved through Extended Ventilation and Minimal Increase in Disposal Area

By extending the time during which loaded emplacement drifts are ventilated, the repository could be operated at lower temperatures with minimal increase in the disposal area. This section describes two example lower-temperature operating mode scenarios that are likely to satisfy the following three objectives:

  1. Maintain average waste package surface temperatures below 85°C (185°F)

  2. Ensure that 70,000 MTHM of waste fits mostly within the upper block, as shown in Figure 2-10

  3. Close and seal the repository within approximately 300 years.

Example scenarios 1 and 2 could achieve a lower-temperature operating mode through extended forced and natural ventilation. Example scenario 2 also includes a de-rated waste package.

Example Scenario 1: Increased Waste Package Spacing and Extended Ventilation—In this example, loaded waste packages would be emplaced an average of about 2 m (6.6 ft) apart to create a 1-kW/m drift thermal load at emplacement. The drift-to-drift spacing would remain at 81 m (266 ft), as specified in the design described in this report. For the first 75 years after the start of emplacement, fans would actively ventilate the drifts with an airflow rate of 15 m3/s per drift. Because of the time required for emplacement, the drifts loaded last would be actively ventilated for 50 years. The repository would be allowed to ventilate naturally for 250 years. Other operational parameters would be unchanged.

Figure 2-12 contains plots of waste package surface and drift wall temperatures over time for the drifts loaded in the final year of emplacement operations under this scenario, as projected by two-dimensional models (BSC 2001h). In the plots shown in Figure 2-12 (also in Figures 2-13, 2-14, and 2-15), the time axis begins after the last emplacement drift is loaded; this results in bounding projections of the maximum waste package and drift wall temperatures in drifts that are loaded earlier in the emplacement period. Descriptions of the repository layout and the changes to the ventilation system that would be required to implement this scenario are discussed later in this section.

Some of the implications of this scenario are (1) the flexibility to readily adjust to a higher-temperature operating mode in drifts loaded later by moving waste packages closer together; (2) a requirement for additional drift excavation to accommodate more widely spaced waste packages; (3) increased complexities in projecting the thermal-hydrologic response of the repository because the widely spaced waste packages would act more like point heat sources within drifts; and (4) the programmatic uncertainty associated with the protracted period of natural ventilation.

Example Scenario 2: De-Rated or Smaller Waste Packages—In this scenario, the thermal output of the waste packages is reduced by limiting waste package loading. This can be achieved by limiting the number of spent nuclear fuel assemblies to less than the waste package design capacity (de-rating) or replacing the large waste packages (e.g., ones containing 21 pressurized water reactor fuel assemblies) with smaller waste packages (e.g., ones that would contain 12 pressurized water reactor fuel assemblies) that have a lower thermal output.

All waste packages would be placed approximately end-to-end within the drifts to create a 1-kW/m linear thermal load at emplacement. Other operational parameters of this scenario are identical to those in Scenario 1. Two-dimensional projections of waste package surface and drift wall temperatures over time are the same as those shown in Figure 2-12.

Figures 2-12 through 2-15 [2-12, 2-13, 2-14, 2-15] show variations in the waste package and drift wall temperature responses with time. Figure 2-12 illustrates a representative process, as follows.

  1. During the forced ventilation period (0 to 50 years), the heating effect of the waste is steadily reduced until the fans are turned off in year 50.

  2. From that point, natural ventilation flow continues to remove heat; however, the reduced flow rate (3 m3/s) takes several years to turn the temperature response to a downward trend (approximately years 70 to 100).

  3. The natural ventilation flow rate gradually decreases over time as the waste decays, since the heat from the waste is the main driver in inducing convective currents in airflow. The simplified calculation methods used could not simulate a steady decline in the natural ventilation rate, so this decline was simulated by stepping down the rate in years 100 to 300 from 3 m3/s to 1.5 m3/s.

  4. The temperature rises from years 100 to approximately 130 due to this abrupt reduction in the natural ventilation flow rate. After year 130, the reduced flow rate starts reducing temperatures until year 300, when the repository would be sealed and closed.

  5. At that point, the lack of ventilation induces a steep rise in the drift temperatures until the host rock's capacity to transfer heat away from the waste packages starts to have a regulating effect on drift temperatures and establishes a very slow downward trend in thermal response over a period of 1,000 years or longer.

Note that in the more natural situation of a steady decrease in the natural ventilation flow rate, the calculated thermal response curves would actually show a steady decrease in drift temperatures across most of the natural ventilation period, rather than the variations shown around the 100-year period.

The primary difference between this scenario and Scenario 1 is a potential reduction in the complexity of modeling the thermal-hydrologic response of the repository because the thermal loading would more closely resemble a line load. A second difference is the increase in the number of waste packages needed, which could result in an increase in the total excavated drift length and the total area required for emplacement.

2.1.5.2.2 Lower Waste Package Temperature Achieved through Increased Disposal Area and Limited Ventilation Period

Lower-temperature operating goals can also be achieved with a limited increase in the ventilation period by increasing the area used for emplacement. The three objectives for this set of examples are:

  1. Maintain average waste package surface temperatures below 85°C (185°F)

  2. Ensure that 70,000 MTHM of waste fits within the upper and lower blocks (Figure 2-10)

  3. Close and seal the repository within approximately 125 years.

Example scenarios 3 and 4 could achieve lower temperatures through extended forced ventilation. Example scenario 4 also includes extended surface aging of spent nuclear fuel.

Example Scenario 3: Increased Spacing and Duration of Forced Ventilation—Lower temperatures could be achieved with a limited increase in the preclosure period by emplacing waste packages an average of about 6 m (20 ft) apart. This would create a drift thermal load at emplacement of approximately 0.7 kW/m. The drift-to-drift spacing would remain at 81 m (266 ft). The loaded drifts would be actively ventilated for 125 years from the start of waste emplacement, with the drifts that are loaded in the last year of emplacement operations receiving 100 years of forced ventilation. Figure 2-13 displays preliminary projections of waste package surface and drift wall temperatures over time for this scenario.

The implications of this approach include: (1) a preclosure period comparable to the current preclosure period of approximately 100 years; (2) an increase in waste package spacing, which would likely result in point thermal loading of drifts and may give rise to increased complexities in modeling the thermal-hydrologic response of the host rock; (3) increases in the total excavated drift length and the total area required for emplacement; and (4) requirements for additional years of forced ventilation, maintenance, and other site and operations support.

Example Scenario 4: Extended Surface Aging with Forced Ventilation—In this example, surface aging of the hotter portion of the commercial spent nuclear fuel inventory, combined with the spacing of waste packages approximately 2 m (6.6 ft) apart within the drifts, reduces the linear thermal load to about 0.5 kW/m at emplacement. Surface aging of the hottest wastes would extend the total emplacement period from approximately 24 years to 50. However, initiation of repository operations would not be delayed because the cooler commercial spent nuclear fuel, along with the generally cooler DOE waste forms, could be emplaced immediately while the hotter commercial spent nuclear fuel cools through aging.

To meet the goal of a maximum waste package surface temperature of 85°C (185°F), forced drift ventilation would continue for approximately 125 years from the start of waste emplacement, with the last drifts loaded receiving 75 years of forced ventilation. At the end of the operating period, the repository would be closed and sealed, with no provision for extended natural ventilation. Figure 2-14 shows preliminary projections of waste package surface and drift wall temperatures over time for this scenario.

The implications of this scenario include: (1) the ability to accommodate the waste packages, with drift-to-drift spacing of 81 m (266 ft), in the areas currently characterized for a repository; (2) a preclosure period comparable to that of the higher-temperature operating mode; (3) a smaller increase in total drift length and disposal area than in Scenario 3; (4) an increase in the spacing of waste packages, which may result in thermal point loading that introduces additional complexities in modeling the thermal-hydrologic response of the potential repository; and (5) a longer emplacement period and additional fuel handling activities, which could increase the preclosure safety risk.

2.1.5.2.3 Lower Rock Temperature and In-Drift Relative Humidity through Indefinite Natural (Passive) Ventilation

A third approach for meeting the goals of lower-temperature operating modes is to keep the temperature of the host rock below the boiling point of water and maintain in-drift relative humidity below 50 percent by incorporating passive natural ventilation into repository operations. A repository that maintains a low relative humidity is possible because of the natural characteristics of the Yucca Mountain site, including the arid environment and thick unsaturated zone.

An example lower-temperature scenario using this approach is framed around the following objectives:

  1. Maintain in-drift relative humidity below 50 percent

  2. Maintain the host rock temperature below the boiling point of water

  3. Ensure that 70,000 MTHM of waste fits within the upper block (Figure 2-10).

Example Scenario 5: Extended Natural Ventilation—One concept for creating a lower-temperature, dry repository is to increase the ventilation duration of the current operation to approximately 75 years after the start of emplacement or 50 years after the last drift is loaded, followed by an indefinite period of natural ventilation. At the end of the forced ventilation period, the repository would stay open to allow natural ventilation to circulate cooler air for an extended period of time. Figure 2-15 shows projections of waste package surface temperatures, drift wall temperatures, and in-drift relative humidity over time (BSC 2001h).

Implications of this approach include: (1) a total excavated drift length and emplacement area comparable to the high-temperature operating mode layout; (2) the same thermal line loading, waste package spacing, and drift-to-drift spacing as the higher-temperature operating mode; and (3) the programmatic uncertainty associated with the indefinite period of natural ventilation.

2.1.5.3 Comparative Analysis of Alternative Lower-Temperature Operating Scenarios

Thermal management involves the consideration of complex, nonlinear relationships among many parameters of the repository system. The major determinants of peak temperatures are the age of the fuel at emplacement, the linear heat load along each drift, and the ventilation period after emplacement. Figure 2-11 shows the relationships among preemplacement aging, ventilation period, and waste package spacing for an average thermal goal of 96°C (205°F) at the drift wall (BSC 2001h). For this case, the thermal goal can be achieved with a total preclosure period of about 100 years. The results change as the temperature goal is lowered to 85°C (185°F) or below on the waste package surface; in particular, the preclosure period might be extended. With longer periods of ventilation after waste emplacement, aging spent nuclear fuel before emplacement has less impact, and the key parameters become the ventilation period and the overall areal mass loading. The areal mass loading can be varied by changing the linear thermal load in each drift, the drift spacing, or both.

The examples in Section 2.1.5.2 were selected for discussion because they illustrate the effects that varying parameters can have in achieving the objectives of lower temperature and humidity. This section summarizes the main issues associated with the principal trade-offs in the example scenarios.

2.1.5.3.1 Reducing Peak Temperatures through Extended Ventilation Period or Reduced Linear Thermal Load

Analyses of ways to achieve a below-boiling repository indicate that reducing peak temperatures even further, so that waste package surface temperature does not exceed 85°C (185°F), would require a significant reduction in the linear thermal load in each drift when the total preclosure period is about 100 years, which is the period now contemplated for open operation of a repository. Scenarios 1, 2, 3, and 4 would require a substantially larger repository area. If ventilation can be extended well beyond 100 years, meeting the goal of temperatures at or below 85°C (185°F) would not require as much expansion of the disposal area. This extended ventilation period also adds flexibility in choosing between reduced waste package loading or increased waste package spacing. The added flexibility results because, with extended ventilation, the most important factor is total areal mass loading. Scenarios 1 and 2 in Section 2.1.5.2 assume that the ventilation period is extended to approximately 325 years after the start of emplacement. Scenarios 3 and 4 assume that forced ventilation continues for about 125 years after the start of emplacement; they rely primarily on reducing the linear thermal load. These two cases represent different ways of reducing the linear load at emplacement: increasing waste package spacing and aging waste. The following paragraphs focus on the broad trade-offs between extended ventilation and reduced linear thermal load.

2.1.5.3.2 Reducing Linear Heat Load by Using Widely Spaced Large Waste Packages or Closely Spaced Small Waste Packages

The total heat load over the entire emplacement area determines the repository temperature in the long term. However, in the short term, the amount of heat per unit length of drift is more important in determining the peak temperatures of both the rock and the waste package surface. Heat per unit length of drift is the dominant factor before the heat from each drift has time to move through the rock and influence temperatures in the adjacent drifts, and before the heat-producing radionuclides with a short half-life (e.g., strontium-90 and cesium-137) have had time to decay to comparatively low levels. As illustrated by Scenarios 1 and 2 in Section 2.1.5.2, there are two ways of reducing the linear thermal load at emplacement: moving larger waste packages farther apart or using de-rated or smaller waste packages without increasing the spacing.

2.1.5.3.3 Aging Hotter Waste or Reducing the Linear Heat Load to Allow a Short Ventilation Period

To achieve a maximum waste package temperature of 85°C (185°F) with ventilation and closure periods of 100 years or less, the linear thermal load at emplacement must be reduced substantially, requiring an increase in drift length. Preemplacement aging can have a significant effect on the magnitude of the increase needed by allowing the heat-generating radionuclides with an intermediate half-life to decay before emplacement. Scenarios 3 and 4 in Section 2.1.5.2, which use roughly the same preclosure ventilation period, show the effects of using aging to reduce drift length requirements.

2.1.5.4 Other Considerations of Lower-Temperature and Lower-Humidity Operating Modes

Additional considerations associated with the various operating modes that reduce drift temperatures and relative humidity are discussed in this section.

Drip Shields—The drip shield is intended to protect the waste package from seeping water. This aspect of corrosion protection is effective early in the postclosure period and is a component of the DOE defense-in-depth strategy. In a lower-temperature design specifically engineered to remain outside the window of corrosion susceptibility, such protection may be less important or unnecessary. The benefits of drip shields in deflecting seepage and the time of their emplacement will be evaluated in light of the environments associated with the range of thermal operating modes.

Another related design issue is that some of the lower-temperature options include waste package spacing of several meters. In such cases, it may be more cost effective to have individually placed drip shields for each waste package rather than a continuously interlocked set of drip shields (BSC 2001l). The impact of these individually placed drip shields on repository performance will be the subject of future evaluations.

Repository Layout—Changes in such variables as waste package spacing or emplacement drift spacing to achieve lower-temperature operating conditions may alter the layout of the potential repository, including the total size of the repository footprint, as identified in Table 2-2. Layouts have been developed for both the primary (upper) and lower blocks characterized as areas for a potential higher-temperature operating mode, as described in Site Recommendation Subsurface Layout (BSC 2001d), and for a lower-temperature operating mode as presented in Lower-Temperature Subsurface Layout and Ventilation Concepts (BSC 2001g). The latter analysis developed layout configurations that support lower-temperature operating modes for the 70,000 MTHM and 97,000 MTHM cases, utilizing the primary and lower blocks. The analysis also presented results of parametric studies to evaluate the sensitivity of the layout configurations against operational and design options including waste package spacing, smaller waste package sizes, waste package heat output, and emplacement drift spacing.

Nonuniform Temperatures within the Drift—When the thermal loading within drifts is adjusted by increasing the distance between waste packages, the thermal load to the rock environment becomes more discretized along the drift axis, more like point loads than a continuous line load. This could result in a nonuniform temperature distribution along the axis of the drift after the repository is sealed. A significant nonuniform temperature distribution in the drift walls or within the emplacement drift could cause water or steam to move from the high heat region toward the low heat region, resulting in water dripping from the walls onto the drip shields, or water condensing on drip shields or low heat output waste packages. The rate of moisture transport decreases with a decrease in the thermal gradient.

Natural Ventilation—Another possibility for mitigating uncertainties associated with corrosion is to keep in-drift relative humidity below 50 percent by allowing longer-term natural ventilation of the repository. By taking advantage of natural ventilation, a repository can be operated so the temperature of the host rock remains below the boiling point of water while relative humidity stays below 50 percent, the value bounding the conservatively defined window of susceptibility for localized corrosion of Alloy 22.

It is illustrative to examine a scenario in which the period of natural ventilation is extended indefinitely. In example scenario 5, the repository is operated through the periods of emplacement and for 75 years after emplacement, during which the emplacement drifts are actively ventilated, after which the repository is allowed to ventilate naturally for an extended period of time. Figure 2-15 depicts nominal waste package surface temperature, emplacement drift wall temperature, and in-drift humidity as a function of time for this scenario, as predicted by two-dimensional models (BSC 2001h). In this case, the natural ventilation period was allowed to extend indefinitely. Natural ventilation flow rates were modeled at 3 m3/s per drift for the first 50 years, at 1.5 m3/s per drift for the next 200 years, and at 1 m3/s per drift thereafter. In calculating relative humidity, an infiltration flux of 60 mm/yr (2.4 in./yr) was assumed over the emplacement drifts. (This infiltration flux is ten times greater than the present-day mean infiltration rate over the repository footprint.) As Figure 2-15 shows, the peak drift wall temperature for this scenario is projected to stay below 90°C (194°F). Because of the ample carrying capacity of the dry desert air, the average relative humidity in the drifts does not exceed 30 percent until waste package temperatures drop below 40°C (104°F).

Natural Analogues—Natural analogues support the concept that a deep geologic repository in the unsaturated zone would keep waste dry and isolated, particularly through natural ventilation. Dry environments, in which fragile cultural artifacts have been preserved for millennia, could provide a hospitable environment for more robust materials. Prehistoric paintings, for example, survive in numerous caves that are analogous to a deep geologic repository. Water tends to flow around caves and tunnels in the unsaturated zone, depending partly on the size of the openings: the smaller the size of the cave, the more likely water will flow around it. Most unsaturated zone caves with highly preserved artwork have openings larger than the 5.5-m (18-ft) diameter proposed for emplacement drifts. This suggests the drifts may remain even drier than caves, which have stayed dry enough to preserve ancient paintings (Stuckless 2000, pp. 2 to 4). A cooler repository would be more closely related to these natural analogues than a higher-temperature repository.

The oldest authenticated examples of Paleolithic (older than 10,000 years) human cave paintings are the Chauvet paintings in southeastern France. These recently discovered cave paintings include images of the mammoth, long extinct, and the rhinoceros, long absent from Europe. The intact Chauvet paintings are more than 30,000 years old and still survive in a subhumid climate with annual precipitation ranging up to roughly 80 cm (31 in.); average annual precipitation at Yucca Mountain is less than 20 cm (8 in.). On the cave's ceiling, even the soot from later human use of oil lamps has remained, despite having been deposited some 26,000 years ago (Stuckless 2000, pp. 2 to 4).

Spirit Cave, about 75 miles east of Reno, Nevada, is another example of a natural analogue to a repository. Several small burial pits were discovered in this cave; in one of these pits, archaeologists found an almost completely intact human body, mummified by the dryness of the climate. Subsequent analyses, including carbon-14 testing, have shown the body to be that of a 45- to 55-year-old male who died around 9,400 years ago. His scalp was completely intact, including a small tuft of hair. He wore a breechcloth of fiber and was wrapped in mats woven of fibers from tule leaves, all highly preserved (Barker et al. 2000, Section 4).

2.2 REPOSITORY SURFACE FACILITIES

The design described in this report focuses on protecting the health and safety of both repository workers and the public. To meet these requirements, an integrated set of surface facilities is being designed. Operation of a repository would involve a number of distinct but interrelated waste activities and functions; the major ones include receiving, handling, and packaging. The following design description discusses the surface facilities for a potential repository for spent nuclear fuel and high-level radioactive waste, based on current concepts documented in Engineering Files for Site Recommendation (CRWMS M&O 2000p) and WHB/WTB Space Program Analysis for Site Recommendation (CRWMS M&O 2000q).

The surface facilities would be located in the North Portal Repository Operations Area, the South Portal Development Area, and the Surface Shaft Areas. Figure 2-16 shows the general surface layout of the North Portal Repository Operations Area and South Portal Development Area. The Surface Shaft Areas, where the ventilation shafts and fans would be located, would be on the crest of Yucca Mountain. Together (but excluding the area for the solar power arrays), these areas would cover more than 105 acres of land, on which at least 30 structures would be built to house the operations and services needed for safe and effective repository operations (CRWMS M&O 2000p, Attachment II, Section 2.11.1).

The North Portal Repository Operations Area is logically segregated into the Radiologically Controlled Area, the balance-of-plant area, and the site services area. The Radiologically Controlled Area comprises all facilities necessary to receive, package, and emplace waste in the repository. The balance-of-plant area comprises general infrastructure facilities, such as administration, emergency management (medical and fire), and motor pool and fleet services. The site services area comprises general parking and the visitor center.

The South Portal Development Area would support continuing construction of a repository, even as the North Portal Repository Operations Area accepts and prepares waste for underground emplacement in the first emplacement drifts. This section focuses on the proposed design and development of three primary structures at the North Portal surface facility. Section 2.3 addresses other areas more closely associated with subsurface development and emplacement activities.

Consistent with 10 CFR Part 63 (66 FR 55732), a set of monuments extending at least 6.2 m (20 ft) above the surface would be constructed prior to the beginning of postclosure to identify the geologic repository area.

In addition to the design requirements driven by the fuel blending inventory strategy (described in Section 2.2.1), continual emphasis on environmental, industrial, and radiological health and safety criteria will be designed into all of the surface facilities and systems. The ALARA radiological health and safety design and operational criteria are discussed in Section 2.2.5. Operations at the North Portal Repository Operations Area are described in Section 2.2.2. Details of these surface facilities and systems, and the designers' emphasis on health and safety, are presented in Section 2.2.4.

2.2.1 Fuel Blending Inventory Strategy

Spent nuclear fuel and high-level radioactive waste arriving at the repository would be in solid form but in a variety of types and sizes. Hence, the materials would arrive in a variety of transportation casks, all certified for use by the NRC.

Commercial spent nuclear fuel would arrive as individual fuel assemblies placed directly into transportation casks, or in dual-purpose canisters, which would have to be opened to remove the fuel assemblies. DOE spent nuclear fuel would arrive in disposable canisters. Because of the variety of waste forms, a number of different designs for disposal containers (called waste packages once they are loaded, sealed, and certified) would be needed. Figure 2-17 depicts the anticipated types of transportation casks, waste forms, and disposal containers (CRWMS M&O 2000p, Attachment II, Section 1.1.3.1).

The radioactive decay process generates heat. The concentration of the particular isotopes present vary among the different waste forms, so different waste forms generate different amounts of heat. The design and operating mode described in this report were developed to meet established temperature limits for the higher-temperature operating mode (i.e., a maximum heat output of 11.8 kW for all waste packages) (CRWMS M&O 2000p, Attachment II, Section 1.1.1.4).

The limit on heat output from individual waste packages imposes special considerations for operations. The strategy for controlling heat output for the waste packages is to load low heat output waste with high heat output waste to balance total waste package heat output. This process, called "fuel blending," applies only to commercial spent nuclear fuel.

Some fuel assemblies must be placed into fuel blending inventory until they generate less heat from radioactive decay, or until additional low heat output fuel assemblies arrive for blending. Fuel assemblies would stay in inventory until they are selected for blending. By carefully planning and implementing a fuel-blending procedure, the heat output of the waste packages can be limited and optimized without increasing the number of waste packages and the amount of fuel blending inventory unnecessarily. The tracking, recording, and archiving of waste forms would be handled by the operations monitoring and control system (Section 2.2.4.2.12).

Engineers developed a computer model to estimate the design amount of fuel blending inventory space needed. Computer model simulations helped form the present configuration of the Waste Handling Building (CRWMS M&O 2000p, Attachment II, Section 1.1.1.3). Results indicate that, for the 11.8-kW heat output limit for the waste package, a fuel blending inventory capacity of approximately 5,000 MTHM, or 12,000 spent nuclear fuel assemblies, will be needed. This simulation also showed the surface facilities will need an inventory of approximately 40 small canisters (containing DOE spent nuclear fuel or high-level radioactive waste) for the canister transfer system. This small canister inventory is needed to ensure that a waste package containing five high-level radioactive canisters and one DOE spent nuclear fuel canister can be loaded using the canister inventory on hand. The inventory size accounts for the fact that these canisters will not always arrive at a ratio of five high-level waste canisters to one spent nuclear fuel canister. Additionally, the simulations established the need for two assembly transfer lines and one canister transfer line (CRWMS M&O 2000p, Attachment II, Section 1.1.4.1). The use of fuel blending inventory in the assembly transfer system ensures that any disposal container loaded with commercial spent nuclear fuel will not exceed a thermal output of 11.8 kW for the design and operating mode described in this report.

2.2.2 Operations in the North Portal Repository Operations Area

This section describes the movement of spent nuclear fuel and high-level radioactive waste forms through the North Portal Repository Operations Area. It presents the interrelationship of the systems, equipment, and facilities for receiving, preparing, packaging, and transporting these waste forms. It also provides information on how the waste would arrive at the repository, how the systems for handling the different waste forms would operate, and how secondary low-level radioactive waste would be handled.

2.2.2.1 Waste Receiving Operations

Spent nuclear fuel and vitrified high-level radioactive waste would be transported to the repository in NRC-certified transportation casks by U.S. Department of Transportation-licensed cask transportation contractors. The waste would be transported by rail and/or road to the North Portal security station, where personnel will verify the shipping manifests, then inspect and survey the cask and its carrier. After the cask and its carrier enter the Radiologically Controlled Area, they would be stationed in parking areas designated for either truck carriers or rail carriers. When the cask is scheduled for processing, a site prime mover (transporter) would move the cask and carrier to the Carrier Preparation Building. Figure 2-18 presents a simplified flow diagram of waste receiving operations.

Inside this building, workers: The casks would be taken on the carrier to a parking area to await movement to the Waste Handling Building carrier bay according to operations scheduling requirements.

2.2.2.2 Waste Handling Operations

At the carrier bay, the carrier/cask handling system would lift the transportation cask to a vertical position and place it on a cask transfer cart. Depending on the cask's contents, the cart would move to one of two transfer systems. Casks that contain disposable containers (e.g., DOE canisters that would not be opened but transferred, as is, directly into a disposal container) would go to the canister transfer system. Casks that contain commercial spent nuclear fuel in dual-purpose canisters or individual fuel assemblies would go to the assembly transfer system. Figure 2-19 is a flow diagram of Waste Handling Building operations (CRWMS M&O 2000p, Attachment II, Sections 1.1.1 through 1.1.5).

The Waste Handling Building would have one canister transfer line that moves the disposable fuel canisters through the building to prepare the waste for emplacement in the repository. The system would move arriving casks through an air lock on a transfer cart into a cask preparation area. Once a cask arrives inside the cask preparation area, workers would use remotely operated equipment to vent and sample gases from the cask, remove the lid bolts, and open the cask. An overhead crane would move the cask to a transfer cart, which would take the cask to a shielded transfer area. Inside the transfer area, machines would remove the canister from the cask. The canister may go directly into a disposal container for repository emplacement or to a holding rack for later placement in a disposal container. Another transfer cart would move loaded disposal containers to the disposal container handling system. A transfer cart would move the empty transportation casks back to the cask decontamination area, where they would be surveyed and decontaminated, if required, before return shipment. From the decontamination area, casks would be moved to the carrier/cask handling system, which would place them back on a transporter. The empty cask and cask transporter would return to the Carrier Preparation Building to be readied for offsite shipment.

The Waste Handling Building would have two assembly transfer lines. Each line would operate independently to handle waste throughput and support maintenance operations. The assembly transfer process begins by moving the cask on a transfer cart through the air lock into the cask preparation area. Once inside the cask preparation area, workers would use remotely operated equipment to inspect, vent, and cool the cask and remove the cask lid bolts. A large overhead crane would lift the casks and place them in a cask unloading pool, where fuel-handling machines would open the casks and unload the fuel assemblies. If the cask contains dual-purpose canisters, they would be removed and placed in an overpack, where the top of the canister would be cut off. The system would move the empty casks and dual-purpose containers back out through the cask decontamination area. The fuel-handling machines would transfer the fuel assemblies, one at a time, to a holding pool, where they would be placed in assembly baskets. A transfer cart would move the baskets containing the fuel assemblies underwater from the assembly holding pool through a transfer canal to a fuel-blending inventory pool. When a fuel assembly is selected from the fuel inventory pool for packaging, a transfer cart would move it underwater back through the fuel blending pool to an inclined transfer canal and onto a cart that connects to the assembly drying area.

After fuel assemblies arrive at the assembly drying area, a fuel-handling machine would transfer them into one of two drying vessels. After drying, the system would retrieve the assemblies and transfer them, one at a time, to a disposal container. The empty assembly baskets would be returned to the pool area for reuse. After loading, workers would purge the disposal container with inert gas and temporarily seal it for transfer to the disposal container handling system.

The disposal container handling system would receive loaded disposal containers from both the canister transfer system and the assembly transfer system. Each disposal container would again be purged with inert gas, after which the container's lids would be welded and the welds inspected. If the welds meet inspection criteria, the sealed disposal container would be reclassified as a waste package. A crane would transfer the waste package to the transporter loading area, where it would be decontaminated and placed on a pallet, then on a transporter for emplacement in the subsurface repository. Table 2-3 gives the preliminary performance specifications for this equipment.

2.2.2.3 Treatment of Low-Level Radioactive Waste from Repository Operations

Operations at a repository—receiving, handling, and repackaging commercial and DOE spent nuclear fuel and high-level radioactive waste—would generate secondary low-level radioactive waste. Some of this waste may be defense-related transuranic waste. In this report, any transuranic and low-level waste is referred to collectively as low-level waste. The majority would be generated in the Waste Handling Building; smaller quantities may be produced in the Waste Treatment Building, where secondary low-level waste is processed. Operations in the Carrier Preparation Building are not expected to generate any significant waste quantity. All secondary waste would be monitored at the point of generation, and administrative controls would direct its disposition. The design of the repository would include a processing system, which would minimize the volume of liquid and solid waste. To the extent applicable, it would be properly treated and packaged for shipping and disposal off the repository site.

Some hazardous wastes may also be generated during repository operations. The use of hazardous constituents will be reviewed and controlled so that the generation of hazardous wastes is minimized. When hazardous wastes are commingled with radioactive wastes, the result is mixed waste. The generation of hazardous and mixed wastes will be minimized at the repository. However, if it is generated, it would be collected and repackaged for shipment to an approved offsite location for treatment and disposal. The packaged mixed waste would then be staged in the Waste Treatment Building for transport offsite to an approved facility. No low-level or hazardous waste would be disposed in the potential repository. Hazardous wastes that are not mixed wastes would also be packaged for shipment offsite to an approved treatment disposal facility. Wet solid low-level waste (e.g., spent ion exchange resins and filtration materials) generated in the Waste Handling Building would be collected and packaged for disposal. Then it would be transferred in containers to the Waste Handling Building for shipment off the site for disposal. Any other solid low-level waste generated in the Waste Handling Building that does not exceed the radioactivity limit for the Waste Treatment Building would be collected at its point of origin and transferred to the Waste Treatment Building to be processed and packaged for shipping and disposal at an approved low-level radioactive waste facility. Solid waste that exceeds Waste Treatment Building administrative activity limits would be packaged at the source of generation for shipment and disposal off the repository site. Spent dual-purpose containers would be volume-reduced and disposed at a suitable low-level radioactive waste disposal site. Recycling of spent disposal containers for recovery of metal content will be examined in future design activities. The Waste Treatment Building would be large enough to hold all the necessary equipment for processing the proposed maximum annual secondary-waste generation rate on a regular operating schedule (CRWMS M&O 2000p, Attachment II, Section 1.2).

2.2.3 North Portal Repository Operations Area Layout

This section describes the overall orientation, configuration, and general construction features of the repository surface facilities.

The North Portal Repository Operations Area would include a Radiologically Controlled Area and a balance-of-plant area (Figure 2-20). The Radiologically Controlled Area, also known as the protected area, is where the waste would be received from offsite transportation carriers and placed in waste packages for disposal. The balance-of-plant area includes all structures and systems supporting repository operations that are not encompassed by the Radiologically Controlled Area. An additional area of 350 acres is available to stage retrieved waste, should the need arise; this area should be sufficient to stage all the waste that may be emplaced in the potential repository (CRWMS M&O 2000p, Attachment II, Section 2.11.3.3).

A buried storm drainage collection system would contain water runoff from the Radiologically Controlled Area. The drainage system would also prevent spillage over the fill slopes and runoff from the balance-of-plant area. A retention pond would be built to prevent storm water pollution (CRWMS M&O 2000p, Attachment II, Section 2.11.3.1).

Except for its north edge, the North Portal pad would be above the flood-prone area of the probable maximum flood. Two open channels around the perimeter of the pad would protect the North Portal from water flow. The operating floor of the Waste Handling Building is 0.5 m (1.5 ft) above the maximum elevation of the flood stage that intersects the building to allow for freeboard (CRWMS M&O 2000p, Attachment II, Section 2.11.3.1).

2.2.4 Surface Systems and Structures

This section includes detailed descriptions of the layout, support structures, systems, and utilities of each major surface facility. Design objectives and criteria are derived from NRC regulations, industry standards for nuclear facilities, and DOE policy. For example, NRC regulatory guides were used in preparing the System Description Document design requirements. NRC guides used in developing the System Description Document for the Waste Handling Building included:

 System Description Documents for the surface facilities would be maintained and updated as required over the life of the facility or system. Codes and standards were chosen based on a review of relevant federal laws and regulations, as well as applicable industry codes, standards, and good engineering practices. Three structures and systems in the surface facilities have been classified as QL-1 (YMP 2000b):

QL-1 systems include those structures, systems, and components whose failure could directly result in a condition adversely affecting public safety. These items have a high safety or waste isolation significance. For this reason, QL-1 systems are discussed in greater detail than systems that are not classified as QL-1. Systems that are not classified as QL-1 are defined in Section 5.2.5, where their functions important to safety are likewise described.

The Carrier Preparation Building, Waste Handling Building, and Waste Treatment Building are described in this section, as are the processes that would occur within each and the equipment that would support such activities. Some of the operations discussed, along with the equipment needed to perform them, are:

2.2.4.1 Carrier Preparation Building

The Carrier Preparation Building, to be located at the North Portal pad, would support preparation of the waste transportation casks before they enter the Waste Handling Building. Planned as a steel-framed structure, the building would be approximately 58 m (190 ft) long, 37 m (120 ft) wide, and 14 m (46 ft) high. The operations area, divided into two identical carrier operations bays, would accommodate four parallel rail tracks/roadways for passage of both rail and truck carriers. Each bay would have two rail/truck lines, separated by a dual-function work platform and equipment laydown area, a bridge crane, and a bridge-mounted manipulator. The transportation carriers would enter and exit the building through one of eight remotely operated roll-up doors (CRWMS M&O 2000p, Attachment II, Section 1.3).

2.2.4.1.1 Carrier Preparation Building: Architectural and Structural Features

The Carrier Preparation Building would be an on-grade, one-story, high-bay, steel-framed structure, enclosed with insulated steel roof and wall panels. The interior framing would be of light-gauge steel and easily decontaminated panels. The foundations would consist of reinforced concrete spread footings, to support the building's columns, and continuous reinforced concrete mat foundations, to support the railroad tracks. To mitigate vibrations from carrier movement, the spread footings would be separated from the mat foundations. The building's columns would support two bridge cranes running the length of the building, each of which would span a gantry crane for servicing the tracks (CRWMS M&O 2000p, Attachment II, Section 1.3.1.3.1).

2.2.4.1.2 Carrier Preparation Building: Material Handling System

The material handling system in the Carrier Preparation Building would receive and inspect shipping casks from the carrier/cask transport system, then prepare the casks for unloading in the Waste Handling Building (Figure 2-21). Four parallel tracks/roadways would permit the passage of both truck and rail carriers. The two outer tracks/roadways would serve incoming carriers from the rail yard or truck-parking area, and the two inner tracks/roadways would serve outgoing carriers (CRWMS M&O 2000p, Attachment II, Sections 1.3.2.1, 1.3.1.3.1).

Receiving operations include:

Shipping operations for carriers/casks leaving the repository would include:

Unloaded casks would undergo the same series of operations as loaded casks, except in reverse (CRWMS M&O 2000p, Attachment II, Section 1.3.2.1).

One 10-ton capacity, remotely operated overhead bridge crane spanning 17.7 m (58 ft) and one remotely operated manipulator would serve each pair of preparation lines. Operations would support both manual and remote handling of carrier/cask materials. Having both manual and remote handling options would improve the maintenance of facilities and equipment, permit the replacement of interchangeable components, and lower radiation doses to workers. The building's support equipment would include tools and fixtures for removing and installing personnel barriers, impact limiters, cask lifting attachments, and cask tie-downs (CRWMS M&O 2000p, Attachment II, Section 1.3).

The material handling system would interface with the cask/carrier transport system to move carriers to and from the building. The Carrier Preparation Building would house all necessary equipment and systems (e.g., facility, utility, safety, auxiliary) to support its operations and protect personnel (CRWMS M&O 2000p, Attachment II, Section 1.3.2.1). No spent nuclear fuel or high-level radioactive waste casks would be lifted in the Carrier Preparation Building. The waste forms would be protected by the transportation casks.

2.2.4.2 Waste Handling Building

The Waste Handling Building would provide the space, layout, structures, and built-in systems to support waste handling operations, loading and holding of waste packages, and inventory of unused disposal containers (CRWMS M&O 2000s). This complex would also provide a safe environment for personnel and equipment involved in waste handling operations. Figures 2-22 and 2-23 show the Waste Handling Building in plan view and sections, respectively. A 30.5-m (100-ft) elevation designation has been assigned to the finished grade elevation of the Waste Treatment Building; all other references to building elevation are measured from this assigned designation. Table 2-4 gives the preliminary performance specifications for the Waste Handling Building.

The Waste Handling Building would contain different systems to:

The building would also provide the needed space and layout for maintenance, administration, and other support operations associated with waste handling activities. It would have designed-in means of protecting its systems and subsystems from the adverse effects of any natural or man-made environment. The complex will also be designed to ensure that radiation exposure levels to workers are kept ALARA.

2.2.4.2.1 Waste Handling Building: Architectural Features

The Waste Handling Building would be located in the North Portal area. Built adjacent to the south wall of the Waste Treatment Building, the Waste Handling Building would be a multilevel, concrete and steel structure made of noncombustible materials. The exterior walls would be mainly concrete; walls that do not provide shielding for radiation protection would be constructed of metal siding panels with insulation. The building would be approximately 180 m (600 ft) wide by 210 m (700 ft) long (CRWMS M&O 2000q, Figure I-4). Figure 2-23 shows section views of the building. All personnel would enter the building through a security portal. Staff who work in contaminated or potentially contaminated areas would change into protective clothing in the change rooms before proceeding to workstations through entrance/exit corridors. All operations levels would be accessible by corridors and stairwells. To meet function and safety requirements, the operating galleries would be located outside the transfer cells. Operating galleries are shielded areas where operators can safely and remotely observe and control operations. Shielding walls, windows, and doors would protect staff operating and maintaining the five primary waste handling systems.

The Waste Handling Building would integrate the five primary systems (CRWMS M&O 2000q, Section 6.2.1) that receive, lift, unload, handle, reload, package, and deliver high-level radioactive waste to subsurface waste handling systems. Table 2-5 summarizes these preliminary engineering specifications. The primary systems in the building would be:

Carriers would move transportation casks into and out of the Waste Handling Building through vertical lift doors at the carrier bay. The casks would then be transferred through air locks to one of two assembly transfer system lines or to the one canister transfer system line. Each assembly transfer line would contain:

The assembly transfer pools would be connected, via fuel basket transfer canals, to four compartmentalized fuel blending inventory pools. The canister transfer line would consist of a cask preparation and decontamination area, a cask unloading area, and a station for loading the disposal containers with the waste. After being loaded, the disposal containers from the three transfer lines would be staged in the disposal container handling cell, where their lids would be robotically welded on. Finally, the loaded, sealed, inspected, and accepted disposal containers, now referred to as waste packages, would be transferred into the waste package transporter cell and loaded onto a transporter for subsurface emplacement.

A number of systems and structural features would support these waste handling operations. An area would be designated for preparing empty disposal containers, and a holding area would provide room for loaded and sealed waste packages waiting for emplacement. A maintenance bay would be available to maintain the handling cranes. The building would also have shops to repair and maintain instruments, robotic welders, and other equipment, along with storage areas for all necessary tools, maintenance materials, high-efficiency air particulate filters, and gas bottles.

2.2.4.2.2 Waste Handling Building: Structural System

The building's foundation would be a reinforced concrete mat (CRWMS M&O 2000p, Attachment II, Section 1.1.6.4). Before construction of the foundation, the undocumented fill of the existing North Portal pad would be improved by the DOE. The building would be designed to withstand (CRWMS M&O 1999e):

Although unlikely, ground motion from earthquakes of significant magnitude may occur at the site of the Waste Handling Building (for a discussion of site earthquake ground motion, see Section 4.3.2.2). The best way to provide confidence that the Waste Handling Building can be safely constructed and operated is to determine the response of the building to ground motion. To this end, the DOE has performed a preliminary soil-structure interaction analysis, using a simplified conceptual design of the Waste Handling Building. This analysis demonstrates that a Waste Handling Building can be designed so that its response to large earthquakes would be within acceptable levels and its structural members would not be overstressed (CRWMS M&O 2000t). The final detailed design of the Waste Handling Building will be in accordance with NRC regulations, which require the building to safely survive a site-specific earthquake with a probability of occurring once in 10,000 years.

The design of the Waste Handling Building would include features to limit worker radiation exposure to levels that are ALARA (CRWMS M&O 2000p, Attachment II, Section 1.1.6.3.4). Various areas in the Waste Handling Building have been designated to have radiation levels that either preclude human occupancy or in which occupancy would be controlled. These areas are designated as radiation access zones, which are defined as areas with radiation levels that potentially fall within boundaries that correlate to the limits in 10 CFR Part 20 and 10 CFR Part 835 if the facility were to be licensed. The object of the radiation access zones is to provide a design framework that realistically limits radiation exposures to the lowest levels that are reasonable, given the state of technology, economics, and benefits to public health and safety. Figure 2-24 shows the radiation access zones that are currently designated in the Waste Handling Building. The slab thicknesses for missile protection will be validated during detailed design. Radiological areas would have 1.5-m (5-ft) thick concrete floors that can support loads of up to 126 metric tons (140 tons) of heavy equipment that would handle casks and waste packages (CRWMS M&O 2000p, Attachment II, Section 1.1.6.4). The walls would also be concrete, with stepped decreases in thickness at the ceiling. The roof would be a concrete slab supported by steel beams and concrete walls (CRWMS M&O 2000q, Section 6.2.7). The roof structure would also be designed to withstand torn