2. PRECLOSURE SUITABILITY EVALUATION
In accordance with the U.S. Department of Energy (DOE) guidelines at 10 CFR 963.12 (66 FR 57298), the DOE has evaluated the preclosure suitability of the Yucca Mountain site using the preclosure safety evaluation method in 10 CFR 963.13 and the preclosure suitability criteria in 10 CFR 963.14. DOE guidelines explain that if the DOE finds that the results of the preclosure safety evaluation show that the Yucca Mountain site is likely to meet the applicable radiation protection standards, then the DOE may determine the site suitable for the preclosure period. The DOE guidelines at 10 CFR 963.2 reference the applicable preclosure radiation dose limits defined in the U.S. Nuclear Regulatory Commission (NRC) licensing rule at 10 CFR 63.111(a) and (b) and 10 CFR 63.204 (66 FR 55732). The NRC rule follows the U.S. Environmental Protection Agency preclosure dose standards from 40 CFR Part 197. In addition, the NRC licensing rule at 10 CFR 63.111(a)(1) incorporates NRC's 10 CFR Part 20. Section 114(a)(1)(A) of the Nuclear Waste Policy Act of 1982 (NWPA) (42 U.S.C. 10134(a)(1)(A)) requires that the basis of the Secretary's recommendation of a repository site include preliminary engineering specifications for the facility. Accordingly, the preclosure safety evaluation method described in 10 CFR 963.13 (66 FR 57298) uses preliminary design and operational concepts of systems, structures, and components to assess the adequacy of repository facilities to perform their intended functions and prevent or mitigate the effects of postulated event sequences. The preclosure safety evaluation method used to evaluate the repository is discussed in Section 2.2. 10 CFR 963.13 guidelines also state that the preclosure safety evaluation will consider (1) a preliminary description of the site characteristics, the surface facilities, and the underground operating facilities (10 CFR 963.13(b)(1)); (2) a preliminary description of the design bases for the operating facilities and a preliminary discussion of any associated limits on operations (10 CFR 963.13(b)(2)); (3) a preliminary description of potential hazards, event sequences, and their consequences (10 CFR 963.13(b)(3)); and (4) a preliminary description of the structures, systems, components, equipment, and operator actions intended to mitigate or prevent accidents (10 CFR 963.13(b)(4)). These descriptions are given in Section 2.3. DOE guidelines at 10 CFR 963.14 (66 FR 57298) contain the criteria for evaluating the preclosure suitability of the site (see Table 1-1). These criteria are (1) the ability to contain radioactive materials and limit releases (10 CFR 963.14(a)); (2) the ability to implement control and emergency systems to limit exposure to radiation (10 CFR 963.14(b)); (3) the ability to maintain a system and components that perform their intended safety functions (10 CFR 963.14(c)); and (4) the ability to preserve the option to retrieve wastes during the preclosure period (10 CFR 963.14(d)). Evaluations of the ability of the repository to meet these criteria are described in Section 2.4. The DOE has considered the potential performance of a repository at Yucca Mountain in terms of the preclosure suitability criteria to evaluate whether it is likely to meet applicable radiation protection standards. The results of the preclosure safety evaluation and the assessment of the ability of the repository to meet the preclosure suitability criteria are summarized in Section 2.5.2.1 KNOWN TECHNOLOGY AND OPERATING SYSTEMS
A repository at Yucca Mountain would use commercial and nuclear industry technologies for preclosure construction and operations. The methods these technologies use to reduce the risk of event sequences are well understood. Over the past 50 years, large nuclear facilities have been designed, constructed, and operated by the commercial nuclear industry and the U.S. Government. Incorporated into the design of these facilities are features and controls that prevent or reduce the consequences of accidents. The repository design draws upon this extensive experience, and is based on proven technology in use at nuclear installations worldwide. For example, high-efficiency particulate air filters have been used for many years to reduce atmospheric emissions from nuclear facilities. Monitoring systems have also been used for many years to measure atmospheric effluents. Computer codes to estimate exposure from effluents have been developed and are widely used. The principles of radiation shielding are well known, and computer codes are available to aid in shielding design. The principles of time, distance, and shielding are used in analyzing designs and processes to determine that they meet the NRC requirement (10 CFR 20.1101(b)) to keep radiation doses as low as is reasonably achievable (ALARA) (e.g., Health Physics Manual of Good Practices for Reducing Radiation Exposure to Levels that are As Low As Reasonably Achievable (ALARA) [Munson et al. 1988]). ALARA means making every reasonable effort to maintain radiation exposures as far below the dose limits in 10 CFR Part 20 as is practicable, consistent with the purpose for which the licensed activity is undertaken, taking into account the state of technology, the economics of improvements in relation to the state of technology, the economics of improvements in relation to the benefits to public health and safety, and other societal and socioeconomic considerations. Spent nuclear fuel transportation casks are routinely loaded and unloaded in the United States. Heavy loads are routinely moved by bridge cranes at nuclear facilities, as they would be at a repository. Across the United States, commercial nuclear power reactors currently operate spent nuclear fuel pools. At all operating nuclear plants, handling spent nuclear fuel is a routine activity. For example, from 1968 to 1994, about 105,000 spent nuclear fuel assemblies were discharged from commercial nuclear power reactors (DOE 1996a, Table 5). The lessons learned from these experiences would be incorporated into the design and concept of operations for any repository.2.2 BASIC SAFETY ASSESSMENT METHOD
The two basic elements of any safety assessment are event identification and consequence analysis. The first element involves performing a systematic review of relevant site and facility features and processes in order to define the type of events that can occur. Events identified include the full range of probable events, from normal operational events that might be anticipated to occur to very low-probability events. Events are identified by first evaluating potential hazards applicable to the site and facility design, then developing a detailed site-specific and design-specific event scenario, in which event sequences are defined and the anticipated frequency of occurrence of event sequences is established. Based on the frequency of occurrence, events are categorized as Category 1 or Category 2 event sequences. Event sequences with frequencies of occurrence less than one chance in 10,000 before permanent closure are considered beyond Category 1 and Category 2 event sequences and do not require further analysis. This method is consistent with the 10 CFR 63.2 (66 FR 55732) event categorization definition. The second element of the safety assessment involves estimating the consequences of the event sequences identified as Category 1 or Category 2 event sequences in the event identification process. The safety assessment performs an important role in the design process. It plays a key role in the identification of facility design features and controls important to safety and is a primary input to the quality assurance classification process. In some cases, alternative design approaches or additional design features may be identified based on safety assessment results, which are then considered as part of an iterative design process. Based on the insights and results obtained from the safety assessment, the acceptability of the design can be established.2.2.1 Event Identification Process
Events are identified based on a review of repository site characteristics, facility design features, and operational processes to be performed. An analysis of the internal and external hazards associated with repository preclosure operations is performed. Internal hazards are presented by the operation of the facility and associated processes. External hazards involve natural phenomena and outside man-made hazards, such as those posed by aircraft and nearby government or industrial facilities. The methodology used in the event identification analysis provides a systematic means to identify facility hazards and associated events that may result in radiological consequences to the public and repository workers during the repository preclosure period. The first step in the hazard identification process is to develop a list of generic internal and external events that could result in radiological consequences to the public or repository workers. This generic list is not facility-specific and attempts to identify potentially hazardous events by providing a comprehensive list of possible events. The generic lists developed for the internal and external hazard analyses are based upon established hazard evaluation techniques (Stephans and Talso 1997; American Institute of Chemical Engineers 1992). Tables 2-1 and 2-2 list these generic internal and external events. Once the site characteristics, facility design, and operational processes are defined, they are evaluated against specified criteria to determine the credibility of generic hazard events that could result in radiological consequences. Event applicability criteria are developed for the generic events to support the applicability determination. If the criteria are satisfied, the generic event has the potential for radiological consequence and is added to a list of specific initiating events to be considered in the design and safety analysis. The criteria used to determine the applicability of internal hazards as initiators of event sequences1 are provided in Preliminary Preclosure Safety Assessment for the Monitored Geologic Repository Site Recommendation (BSC 2001c, Section 5.1.1.3.1) and are listed below for each event category. Applicability to a functional area of design is determined by a positive response to all questions within a hazard category or subcategory, as appropriate.2.2.2 Event Sequence Categorization Process
The result of the event sequence identification process is a list of event sequences with a corresponding quantitative frequency of occurrence. The frequency of occurrence for each event sequence is determined using fault tree analysis or data from historical events. The frequency of occurrence is usually expressed in terms of the chance of the particular event sequence occurring during facility operations, for example, "3 chances in 100 of occurring before permanent closure of the repository." In this example, if the repository operates for 100 years and the event sequence frequency is uniform over the entire period, the event sequence frequency can be expressed as 0.0003 per year or 3.0 × 10-4 per year. Initially, when postulating the event sequence, no credit is given to design features that could prevent or mitigate the event (i.e., the most severe consequences are evaluated). If the radiation dose consequences of an event sequence are unacceptable, design features are added to prevent or mitigate the event. Based on frequency of occurrence, event sequences are categorized as Category 1 or Category 2 event sequences or beyond Category 1 and Category 2 event sequences, as described in 10 CFR 963.2 (66 FR 57298). Category 1 event sequences are "those event sequences that are expected to occur one or more times before permanent closure" (10 CFR 963.2). This is about equal to an annual frequency of one chance in one hundred (0.01 per year),2 based on a 100-year preclosure operational period (BSC 2001c, Section 4.4.1.2.1). Category 2 event sequences are "other event sequences that have at least one chance in 10,000 of occurring before permanent closure" (10 CFR 963.2). This is about equal to an annual frequency of one chance in 1 million (0.000001 per year), based on a 100-year preclosure operational period (BSC 2001c, Section 4.4.1.2.1). Event sequences that have less than one chance in 10,000 of occurring before permanent closure of the repository are considered beyond Category 1 and Category 2 event sequences. 10 CFR Part 963 does not suggest analyses of beyond Category 1 and Category 2 event sequences. However, in environmental impact statements and environmental assessments, event sequences with an annual frequency of about one chance in 10 million are often evaluated, based on the guidance contained in Recommendations for the Preparation of Environmental Assessments and Environmental Impact Statements (DOE 1993, Section 6.4). The consequences of these types of events are presented in Final Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada (DOE 2002c, Tables H-6 and H-7). Event sequences are developed using event trees, which are diagrams that depict the chronological sequence of events. Figure 2-1 shows an example of an event tree used to define event sequences and quantify their frequency of occurrence. In this example, Event Sequence 1 begins with an unsealed disposal container drop as the initiating event. The second event represents a breach of the spent nuclear fuel assembly. The last event in this branch represents a fully functional ventilation system and associated high-efficiency particulate air filtration. This event sequence has a frequency of 8.4 × 10-3 per year, classifying it as a Category 2 event sequence. Event Sequence 2 represents a release scenario in which the ventilation system and the high-efficiency particulate air filtration system are nonfunctional. This event sequence has a frequency of 1.4 × 10-9 per year, which is considered to be a beyond Category 1 and Category 2 event sequence; this would be the case even if the probability of the high-efficiency particulate air filtration not functioning were increased a hundredfold. Event Sequence 3, with zero probability, represents an unsealed disposal container drop that does not breach the enclosed spent nuclear fuel assemblies and does not result in a release. As illustrated in Figure 2-1, the scenario development process involves analysis of facility features or controls that can affect the progression of an event sequence, including the effects of a successful operation or failure of the heating, ventilation, and air conditioning systems with high-efficiency particulate air filters, where appropriate.2.2.3 Event Sequence Consequence Analysis Process
In accordance with the definition of applicable radiation protection standards in the DOE guidelines at 10 CFR 963.2 (66 FR 57298), separate radiation dose limits are relevant for Category 1 and Category 2 event sequences. The DOE guidelines (regarding the definition of applicable radiation protection standards) reference NRC licensing limits given in 10 CFR 63.111(a) and (b) and 10 CFR 63.204 (66 FR 55732). The NRC licensing limits are usually expressed in terms of potential radiation doses to repository workers or members of the public. NRC radiation dose limits to members of the public for Category 1 event sequences are given in 10 CFR 63.111. NRC radiation dose limits for repository workers for Category 1 event sequences (including normal operations) are given in 10 CFR 63.111(a)(1) and 10 CFR 20.1201(a)(1). For repository workers, only Category 1 event sequence radiation dose limits are applicable. Category 1 Event Sequences—Three sources that are expected to contribute to the annual radiation dose to the public or repository workers from Category 1 event sequences during the facility's preclosure operational lifetime include (1) operational effluents from the Waste Handling Building, (2) operational effluents from the subsurface areas of the repository, and (3) event sequences anticipated to occur at a frequency of 0.01 per year or higher. Section 5.3.5.4.1 in Preliminary Preclosure Safety Assessment for the Monitored Geologic Repository Site Recommendation (BSC 2001c) describes the models used to estimate the radiation doses from Category 1 event sequences. Appendix A of Preliminary Preclosure Safety Assessment for the Monitored Geologic Repository Site Recommendation (BSC 2001c) considers the influence of flexible thermal operating modes with preclosure periods of up to 325 years on Category 1 event sequence selection. Category 2 Event Sequences—The radiation dose from Category 2 event sequences comes from event sequences anticipated to occur with frequencies between 0.01 and 0.000001 per year. This frequency range assumes a 100-year preclosure period that is associated with the higher-temperature repository operating mode. The Category 2 event sequences all involve drops or collisions while handling fuel assemblies, disposal containers, and transportation casks. Section 5.3.5.4.2 in Preliminary Preclosure Safety Assessment for the Monitored Geologic Repository Site Recommendation (BSC 2001c) describes the models used to estimate the radiation doses from Category 2 event sequences. The influence on the selection of Category 2 event sequences of the flexible thermal operating modes with preclosure periods of up to 325 years is discussed in Appendix A of Preliminary Preclosure Safety Assessment for the Monitored Geologic Repository Site Recommendation (BSC 2001c). Several dosimetric quantities were calculated for Category 1 and Category 2 event sequences to show comparison to the 10 CFR 63.111 (66 FR 55732) dose limits referenced in DOE guidelines at 10 CFR 963.2 (66 FR 57298): (1) the total effective dose equivalent; (2) the radiation dose for various organs and tissues, such as the thyroid, lungs, and bone marrow; and (3) the radiation dose for the skin.2.2.4 Use of Features and Controls Important to Radiological Safety
The repository design incorporates a combination of prevention and mitigation features and operational controls. Prevention is the use of design features to reduce the frequency of events that result in radiological release. Mitigation is the use of design features to reduce the dose consequences of a postulated radiological release event sequence and includes features intended to reduce radioactive releases from routine operations that are included in Category 1 event sequences annual dose summation. The safety assessment is used to identify preventive and mitigative features needed to maintain doses within the radiation dose limits in 10 CFR 63.111 (66 FR 55732), as referenced in the DOE guidelines at 10 CFR 963.2 (66 FR 57298). The repository design emphasizes prevention features because prevention provides design and operational benefits. From an operations perspective, surveillance and maintenance of active safety features have been demonstrated to add some operational complexity to existing nuclear facilities. Prevention features are incorporated in the design by performing the safety assessment as an integral part of the design process in a manner consistent with a performance-based, risk-informed philosophy. A risk-informed approach uses risk insights, engineering analysis and judgment, and equipment performance history to focus attention on the most important facility activities and to establish design criteria and management controls based on these risk insights. This approach ensures that design features and operational controls important to radiological safety are selected in a manner that ensures safety while minimizing operational complexity through the use of proven technology. The repository would be designed, constructed, and operated to withstand external events and natural phenomena such that the Category 1 and Category 2 event sequence dose limits are met. For example, Section 2.2.4.2.2 of Yucca Mountain Science and Engineering Report (DOE 2002a) discusses the requirements for designing the surface facilities to withstand the vibratory ground motions associated with earthquakes. As an example, in the assembly transfer system and canister transfer system, overhead cranes and assembly transfer machines would be designed so that they would not become dislodged from their rails during a Category 1 or Category 2 event sequence earthquake (DOE 2002a, Table 2-6). Section 2.2.5 of that report also discusses the design processes used to keep radiation doses to workers ALARA. For accidents involving internal events, the analysis in Design Basis Event Frequency and Dose Calculation for Site Recommendation (BSC 2001e, Table 5) shows that drops of spent nuclear fuel assemblies are important contributors to event sequences. For these types of accidents, the assembly transfer system would be designed, constructed, and operated so that the probability of the dry assembly transfer machine dropping an assembly is low (CRWMS M&O 2000o, Section 1.2.2.1.1). The analyses in Design Basis Event Frequency and Dose Calculation for Site Recommendation (BSC 2001e, Section 5.2.5) shows that the availability of the Waste Handling Building heating, ventilation, and air conditioning system with high-efficiency particulate air filters plays a role in mitigating the consequences of accidents. Therefore, the ventilation system would be designed to be highly reliable. For example, it would be designed to withstand earthquakes, impacts from flying debris (referred to as missiles), fires, or loss of offsite electrical power and still perform its intended safety functions. The key prevention and mitigation methods rely on the use of:2.2.5 Safety Assessment and Quality Assurance Classification Process
Consistent with DOE guidelines at 10 CFR 963.13(b)(4) (66 FR 57298), the preliminary safety assessment provides information and descriptions of structures, systems, components, equipment, and actions to mitigate or prevent accidents. Repository features credited as event prevention or mitigation features in the safety assessment may be considered to be important to safety, and the safety assessment is useful in determining an item's functional role as part of the repository preclosure safety case. Classification is performed in accordance with formal quality assurance classification procedures. Structures, systems, and components important to safety are classified in a graded fashion to ensure quality assurance controls are implemented over the facility life cycle commensurate with an item's importance to safety. The classification process consists of establishing the configuration and function of structures, systems, and components and their effect on repository radiological safety. It is limited to structures, systems, and components procured as a part of the repository system. This information is then evaluated against criteria provided in the procedure to determine the quality assurance classification of the particular item. The following classification categories are specified by Section 3.13 of procedure QAP-2-3, Classification of Permanent Items, to meet requirements of Section 2 of Quality Assurance Requirements and Description (DOE 2000). Quality Level (QL)-1—Structures, systems, and components whose failure could directly result in a condition adversely affecting public safety are classified as QL-1. These items have a high safety or waste isolation significance. QL-1 structures, systems, and components include those items required to:2.3 CONSIDERATION OF THE PRECLOSURE SUITABILITY CRITERIA
This section contains summary descriptions specified in 10 CFR 963.13(b) (66 FR 57298). These descriptions are (1) a preliminary description of the site characteristics, the surface facilities and the underground operating facilities (10 CFR 963.13(b)(1)); (2) a preliminary description of the design bases for the operating facilities and a preliminary description of any associated limits on operation (10 CFR 963.13(b)(2)); (3) a preliminary description of potential hazards, event sequences, and their consequences (10 CFR 963.13(b)(3)); and (4) a preliminary description of the structures, systems, components, equipment, and operator actions intended to mitigate or prevent accidents (10 CFR 963.13(b)(4)). This section also includes a brief summary of the results of the analyses.2.3.1 Preliminary Description of Site Characteristics, Surface Facilities, and Subsurface Facilities
Along with the natural characteristics of the Yucca Mountain site, the design of the geologic repository would protect repository workers, the public, and the environment during the preclosure period. The repository would include the facilities, equipment, processes, and safeguards for safely receiving, handling, emplacing, and monitoring highly radioactive materials. A combination of natural and engineered features would be employed to provide an environment that is safe for repository workers and the public during the operational period (Curry 2001, Section 2.6). The DOE will accept and dispose of spent nuclear fuel from both commercial and government nuclear reactors, as well as high-level radioactive waste from commercial3 and government facilities. To protect repository workers, the public, and the environment, the DOE will comply with applicable regulations governing the handling and disposal of such materials (Curry 2001, Section 5.1.2). Site Characteristics—The potential advantages of Yucca Mountain as a repository site include its remote location, stable geologic environment, arid climate, and deep groundwater table. Section 1.3 of Yucca Mountain Science and Engineering Report (DOE 2002a) describes these site characteristics. A detailed description of site characteristics is provided in Yucca Mountain Site Description (CRWMS M&O 2000c). Preliminary Preclosure Safety Assessment for Monitored Geologic Repository Site Recommendation (BSC 2001c, Section 3.3) identifies the facilities and activities near the repository used in analyzing potential hazards. Yucca Mountain is remote from population centers. It is located about 160 km (100 mi) northwest of Las Vegas, Nevada, on unpopulated land owned by the federal government and adjacent to the Nevada Test Site (DOE 1998b, Volume 1, Section 2.2). The nearest residents are more than 20 km (12 mi) from the potential repository site. The closest year-round housing is near the location of U.S. Route 95 and Nevada Route 373 in the Amargosa Valley, about 22 km (14 mi) south of the site. There is farming, primarily grasses and legumes, about 30 km (19 mi) south of the potential repository site in the Town of Amargosa Valley (DOE 1999, Section 3.1.1.1). The population densities in the region near Yucca Mountain are between zero and four persons per square kilometer (BSC 2001c, Section 3.1.2). Yucca Mountain provides a stable geologic environment. A flat-topped ridge running six miles from north to south and rising approximately 300 m (1,000 ft) above the adjacent valleys, Yucca Mountain has changed little over the last half-million years. Although earthquakes do occur in southern Nevada and are expected to occur at Yucca Mountain, the repository surface and subsurface facilities would be designed to withstand the design basis ground motion. Ground motion from an earthquake at the repository level would be significantly less than on the ground surface. Underground repository structures would be designed for appropriate ground motion levels. Waste packages would not be placed in any underground area where substantial fault movement could occur (YMP 1997a, Section 4.3). Surface Facilities—The surface facilities would be located at the North Portal Repository Operations Area, the South Portal Development Area, the Emplacement Shaft Surface Operations Areas, and the Development Shaft Surface Operations Areas. Together these areas would cover more than 105 acres (42.7 hectares) of land, on which at least 30 structures would be built to house the operations and services needed for safe and effective repository operations (DOE 2002a, Section 2.2). The North Portal surface facilities would include the necessary buildings, equipment, and systems to support the receipt, handling, and preparation of waste before its transport underground. Also located at the North Portal would be facilities for administration, monitoring/testing, emergency response, maintenance, and other support operations. The South Portal Development Area would support construction of the repository and the operation of ventilation intake fans for subsurface development. This area would function independently of the North Portal emplacement area and would include the underground development support facilities needed for construction, maintenance, warehousing, and material staging (Curry 2001, Section 2.3). The surface shaft areas would include structures that will house ventilation and exhaust fans for the underground repository, including power supplies, headframes, and hoist systems. These structures would also support the maintenance of the ventilation and exhaust fans. In addition, air intake areas would be located at both the North and South Portals and at intake-shaft areas above the east and west main drifts. A solar power electrical generation facility would be located to the east of the repository. It will provide supplemental power for the operation of the repository. A detailed description of the surface facilities is provided in Section 2.2 of Yucca Mountain Science and Engineering Report (DOE 2002a). Subsurface Facilities—The subsurface layout would include main drifts and emplacement drifts, two inclined access/egress ramps, and vertical ventilation shafts (Figure 2-2). The drifts would be the largest features of the subsurface repository. The main drifts would be used for ventilation and to transport equipment, personnel, and waste packages. The emplacement drifts would serve as the permanent disposal locations for the waste packages. Access to and egress from the subsurface would be by two gently sloping ramps. Waste packages would be transported into the subsurface facilities through the north ramp, while the south ramp would allow continued excavation of additional drifts. Vertical shafts would be used exclusively for ventilation (Curry 2001, Section 2.5). Each emplacement drift has an entrance at each end and an exhaust raise in the middle. Access into the drift is controlled by doors at each end. The doors would have ventilation regulators (or louvers) to control the flow of ventilation air through the emplacement drift. The doors would be remotely controlled (Curry 2001, Section 2.5). Many of the drifts would be completed before receipt of the NRC license amendment to receive and emplace waste and the start of waste emplacement operations. The remaining drifts would be completed while waste is being emplaced in the repository. These concurrent operations would allow the repository to begin waste emplacement operations within five years of the start of the construction phase. To ensure repository worker safety, the excavation and emplacement operations would be physically separated from each other, and each would have its own ventilation system and ramp access (Curry 2001, Section 2.4). The DOE would actively monitor and maintain the repository from the time the first waste package is emplaced until the repository is permanently closed. Sensors would be used to monitor the waste packages, drifts, and surrounding rock while also providing data. Performance monitoring activities gather data from the waste packages, emplacement drifts, and surrounding rock. These data would be used to enhance the level of understanding of how the repository would likely perform after it is eventually closed. Testing activities would gather data from surface and subsurface facilities, materials, and equipment to ensure the repository is operating within its regulatory requirements (Curry 2001, Section 2.9). Under normal conditions, humans would be prohibited and prevented from entering any emplacement drift containing waste packages. Most monitoring and testing would be performed remotely by sensors and equipment positioned in and around the emplacement drifts. When necessary, robots would be used to investigate conditions in the emplacement drifts. This would eliminate risk to repository workers from heat and radiation emitted from the emplaced waste packages (Curry 2001, Section 2.9). Under off-normal conditions, workers could be sent into emplacement drifts for a short time to help evaluate an unexpected situation. They would only be sent in after the drift had been cooled and portable radiation shielding put in place. A detailed description of the subsurface facilities is provided in Section 2.3 of Yucca Mountain Science and Engineering Report (DOE 2002a) and in Site Recommendation Subsurface Layout (BSC 2001f). When emplacement of the waste inventory has been completed and future generations decide to close the repository, the DOE would submit an application to the NRC for an amendment to the repository license that would permit closing the facility. Closure could take place beginning 50 years after emplacement of the first waste package (unless a different period is approved) or as late as 300 years after final emplacement (Curry 2001, Section 5.1.1.1). The closure process would include installing the drip shields over the waste packages; sealing all openings to the subsurface repository; decontaminating, dismantling, and recycling or disposing of the surface facilities; restoring the surface area as closely as possible to its natural state; and protecting the repository from unauthorized intrusion (Curry 2001, Section 2.10). 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 surface area that overlies the underground repository. After closure, the repository would continue to isolate the waste for thousands of years.2.3.2 Preliminary Description of Design Bases for Facilities and Limits on Operations
For the purpose of this discussion, "design bases" are defined by 10 CFR 963.2 (66 FR 57298) as "information that identifies the specific functions to be performed by a structure, system, or component of a facility and the specific values or ranges of values chosen for controlling parameters as reference bounds for design." Design development for the potential repository follows a structured approach that links statutory, regulatory (e.g., licensing), and derived requirements to the design product. Requirements (e.g., derived licensing requirements) are identified and passed down to individual systems and components through an established document hierarchical system and become more specific and detailed with each successive level. Repository systems are designed to be fully integrated, using a systems engineering approach. In this process, systems are analyzed and then classified as to their importance to preclosure radiological safety. Design work is performed in accordance with a quality assurance program reviewed and accepted by the NRC. Section 2.1 of Yucca Mountain Science and Engineering Report (DOE 2002a) provides additional discussion of the design process. Some of the NRC regulatory guides for potential licensing that were used in the design process are discussed in Section 2.3.3 of this report. Detailed descriptions of the specific functions and controlling parameters of designed systems is provided in System Description Documents. Monitored Geologic Repository Project Description Document (Curry 2001, Section 4) captures, by logical grouping, the hierarchical arrangement of the repository design documents. In that hierarchical arrangement, the repository is divided into three major systems, each with several secondary systems. The waste handling system, which forms the bases for the following discussion, includes four secondary systems: (1) carrier/cask shipping and receiving systems, (2) waste preparation systems, (3) waste treatment systems, and (4) waste emplacement and retrieval systems. The general function of the waste handling system is to receive and process the transportation casks and the carriers, process the waste for emplacement, and emplace waste packages in the emplacement drifts (Curry 2001, Section 4.2.3). Summary descriptions of the function and associated limit of operations of the secondary systems of the waste handling system and the strategy used to prevent and mitigate radiological releases are provided in the following paragraphs. Carrier/Cask Shipping and Receiving Systems—The function of these systems is to receive and process the transportation casks and the carriers by receiving and inspecting the shipping casks, removing the impact limiters, and moving the transportation casks to the Waste Handling Building. This function covers operations from the time when a transportation cask containing waste arrives at the repository until the transportation cask is opened in the Waste Handling Building (Curry 2001, Section 4.2.4; BSC 2001c, Sections 4.3.1 to 4.3.3). The general strategy for preventing a release of radioactive material is to ensure and maintain the structural integrity of the transportation cask containment structures. This is accomplished by providing surface facility cask handling features and associated operational limits that prevent events that could result in exceeding the cask design bases for containment integrity (i.e., features and controls that limit cask lift height) (CRWMS M&O 2001a, Table 6-1). The general strategy for mitigating a release of radioactive material is to provide operations for inspecting and decontaminating transportation casks. Such operations would be located in the Carrier Preparation Building (Curry 2001, Section 3.1.3.1). Waste Preparation Systems—The function of these systems is to process the waste for emplacement by removing the waste from the transportation casks, placing the waste in disposal containers, sealing (welding) and inspecting the disposal containers (once the disposal container is sealed and inspected it is called a waste package), and placing the waste package in the shielded transporter. This function covers operations from the time when a transportation cask is opened until the waste package is placed inside a shielded transporter (BSC 2001c, Section 4.3.4 to 4.3.6). The general strategy for preventing a release of radioactive material from an unsealed transportation cask is to provide features and controls (e.g., positive locking hoisting attachments) at the surface facility that reduce the likelihood of a cask drop event. Additional waste handling operations and the associated prevention and mitigation features are based on whether the waste received is canistered fuel or individual, uncanistered assemblies. The general strategy for preventing a release of radioactive material from canistered DOE spent nuclear fuel and high-level radioactive waste is to ensure and maintain the structural integrity of the canisters' own containment features. This is accomplished by providing features and placing operational limits on the canister transfer system that would prevent events that could compromise the structural integrity of the canister (i.e., features and controls include positive latching lifting attachments and limiting the height to which a canister can be lifted). These features and controls are provided to prevent design bases drop events for some DOE canistered fuel. The general strategy for mitigating a release of radioactive material is to provide a ventilation system with high-efficiency particulate air filters to remove and contain airborne radionuclides (CRWMS M&O 2001a, Table 6-1). The general strategy for preventing a release of radioactive material from handling individual, uncanistered fuel assemblies is to ensure and maintain the structural integrity of the fuel assembly cladding's own containment features. This is accomplished by placing features and operational limits on the assembly transfer system that reduce the likelihood of an assembly drop event (i.e., positive locking hoisting attachments) and limiting the lift height of the disposal container in the transfer cells. The general strategy for mitigating a release of radioactive material from handling individual, uncanistered fuel assemblies is to provide a filtering system on the fuel blending pool and to provide a ventilation system with high-efficiency particulate air filters in the Waste Handling Building to remove and contain airborne particulates (CRWMS M&O 2001a, Table 6-1). The loaded, sealed, tested, and accepted disposal container, now referred to as a waste package, is placed into a shielded waste package transporter, along with its emplacement pallet. At this point in the operations, the waste package provides containment and serves as a preventive feature, which is augmented by surface facility handling features and operational controls that prevent events that could compromise the waste package design basis for containment integrity (i.e., features and controls that limit waste package lift height) (CRWMS M&O 2001a, Table 6-1). Waste Treatment Systems—The function of these systems is to process the waste generated from waste handling operations by managing and preparing spent resin cartridges and other low-level and industrial waste for disposal. Systems for handling and packaging low-level radioactive waste would be provided in the Waste Treatment Building. Systems for handling, packaging, transportation, and disposal of low-level and industrial waste are well established. Low-level radioactive waste would be disposed at a suitable site away from the repository site. Mixed waste is not expected to be generated by the repository operations (BSC 2001c, Section 4.3.8). Waste Emplacement and Retrieval System—The function of this system is to transport and emplace waste packages in the emplacement drifts. This function covers operations from the time the waste package and emplacement pallet are loaded into the transporter through the time the waste package and emplacement pallet are placed on the engineered invert in the emplacement drift (BSC 2001c, Section 4.3.7). This system can also, if necessary, retrieve some or all of the waste packages from the drifts and return them to the surface during the preclosure period. Retrieval is achieved by reversing the emplacement operations. The general strategy for preventing a release of radioactive material is to ensure and maintain the structural integrity of the waste package containment structures. This is accomplished by providing features and control limits on the transporter and rail system that prevent events that could result in compromising the waste package design basis for containment integrity (i.e., reliable brake and gearing system to control the descent speed). Other safety features, such as the use of dual locomotives to move the transporter, also perform preventive functions during descent. During emplacement operations the waste package structure provides containment and serves as a preventive feature, augmented by the transporter, rail system, emplacement handling, pallet, and ground support, to ensure that no Category 1 or Category 2 event sequences can compromise the waste package design basis for containment integrity (CRWMS M&O 2001a, Table 6-1).2.3.3 Preliminary Description of Potential Hazards, Event Sequences, and Consequences
This section provides preliminary descriptions of potential hazards, event sequences, and the consequences of event sequences. Generic repository hazards were described in Section 2.2.1.2.3.3.1 Preliminary Description of External Events
The general strategy for managing external initiating events, based on deterministic NRC licensing precedence, is to design those structures, systems, and components important to safety to withstand the initiating events so that no release scenarios are initiated and no loss of isolation of radioactive material results. This strategy ensures that there are no Category 1 or Category 2 radioactive material release scenarios associated with external initiating events and natural phenomena, even though the initiating events could occur. Table 2-5 lists the external events and natural phenomena initiating events considered in this evaluation. A preclosure operational period of up to 325 years does have implications to external event sequences as described in Section 2.2.2. However, Preliminary Preclosure Safety Assessment for the Monitored Geologic Repository Site Recommendation (BSC 2001c, Section 3.4) indicates that no new external events or natural phenomena would be included for preclosure periods up to 325 years. The external initiating events evaluated in Table 2-5 are appropriate for a preclosure period of about 325 years. Loss of Offsite Power—This event would result in the total loss of external alternating current power to the repository. It is postulated to occur as a result of an external event (e.g., lightning or downed power line) or an internal event (e.g., fire or random equipment failure). Loss of offsite power would temporarily halt the transfer of waste. Because loss of offsite power is assumed to occur one or more times during preclosure operations, it is a Category 1 event sequence (BSC 2001c, Section 5.2.1.1). The strategy for mitigating this event is to prevent Category 1 or Category 2 release scenarios that could exceed applicable preclosure radiation protection standards by providing reliable power through redundant standby power sources (onsite), uninterruptible power, redundant emergency equipment where needed, redundant distribution systems, and mechanical backup controls for components important to safety. Structures, systems, and components important to safety are designed to prevent load drops during a loss of offsite power. Onsite backup power sources with staged loading controls and redundant offsite power lines and sources may be used to ensure continuous power is supplied to structures, systems, and components important to safety. The repository design would also include such features as external lightning rods or grounding system to protect against a lightning-initiated loss of offsite power (BSC 2001c, Section 5.2.1.1). Earthquake—Vibratory Ground Motion—An earthquake is the result of sudden relative motions, or slip, between two adjacent rock surfaces in the earth's crust. The sudden slip results in the release of seismic energy in the form of vibratory ground motion that propagates from the earthquake's location to the earth's surface. This ground motion can impact structures, systems, and components in the surface and subsurface facilities. The possible consequences of an earthquake include a collapse of structures, cracking of structural members or shielding walls, loss of offsite power, ground displacement, and subsurface rockfall (BSC 2001c, Section 5.2.1.2). The DOE would use proven engineering techniques to design structures important to safety to withstand earthquakes in the site area. The repository surface facilities, where waste would be received, prepared for emplacement, and moved into the repository, would be subject to stronger earthquake ground shaking than subsurface facilities, where waste would be emplaced. Preclosure Seismic Design Methodology for a Geologic Repository at Yucca Mountain (YMP 1997a, Section 3.1) establishes seismic hazard probability reference values to be used in determining two levels of design basis vibratory ground motion. The two reference values correspond to Category 1 and Category 2 event sequences and are defined as mean annual exceedance probabilities of 10-3 and 10-4, respectively. The mean annual probabilities were used in the disaggregation of probabilistic seismic hazard estimates (CRWMS M&O 2000p, Section 6.5.3) to identify those earthquakes that control the seismic hazard at the reference probabilities. Ground motion inputs used for preclosure design analyses are described in Section 4.3.2.2.3 of Yucca Mountain Science and Engineering Report (DOE 2002a, Figure 4-166). These inputs are based on a mean annual exceedance probability of 10-4 and were developed for generic locations at the repository elevation (i.e., a depth of 300 m [1,000 ft]) and at a hard-rock outcrop directly above the repository. The safety strategy for the surface facilities is to design the structures, systems, and components important to safety to withstand the effects of a design basis earthquake. The design and construction attributes necessary to ensure that structures and systems are not compromised during a seismic event are well understood and would be applied to the repository facilities (BSC 2001c, Section 5.2.1.2). The following NRC licensing-related documents relevant to design basis seismic events were among the sources considered in the repository design process:2.3.3.2 Preliminary Description of Internal Event Sequences
Radiological consequences for the bounding internal event sequences were evaluated. Bounding event sequences include groups of similar event sequences that result in the maximum radiological consequences to a member of the public at the preclosure controlled area boundary or to a repository worker on site. Collectively, the bounding event sequences establish constraints on the facility design to ensure that the structures, systems, and components important to safety will perform their intended function during an event sequence and that any releases of radioactive material are likely to remain within applicable radiation protection standards referenced in DOE guidelines at 10 CFR Part 963 (66 FR 57298). Internal event sequences were screened into one of three groups based on their frequency of occurrence and potential to result in a release of radioactive material:2.3.3.2.1 Internal Event Sequences with Potential Radionuclide Releases
Events that could potentially result in a release of radionuclides have been classified as Category 1 or Category 2 event sequences, as discussed in Section 2.2.3. The Category 1 event sequences evaluated in Preliminary Preclosure Safety Assessment for Monitored Geologic Repository Site Recommendation (BSC 2001c, Section 5.3.2) occur during the handling of uncanistered commercial spent nuclear fuel assemblies or spent nuclear fuel assembly baskets in the assembly transfer system. DOE spent nuclear fuel, high-level radioactive waste, and immobilized plutonium waste forms were also considered, but it was determined that event sequences involving commercial spent nuclear fuel bounded the potential consequences of handling other waste forms. Table 2-6 identifies the Category 1 event sequences that could potentially result in releases of radioactive material.2.3.3.2.1.1 Sequences Involving Individual Spent Nuclear Fuel Assemblies
Unconfined spent nuclear fuel assemblies (i.e., assemblies not in containers) would be handled remotely, underwater, and individually, during transfer from the cask to the assembly transfer system basket staging rack. Then, during transfer from the assembly transfer system dryer to the disposal container, they would be handled in a dry environment (BSC 2001c, Section 5.3.2.1). While underwater, spent nuclear fuel assemblies could be dropped or impacted as a result of a mechanical or control system failure of the wet assembly transfer machine, or as a result of operator error. These event sequences would occur in the assembly transfer system pool area, which is a confinement area with high-efficiency particulate air filtration. Individual spent nuclear fuel assemblies event sequences that occur underwater are identified in Table 2-6 by sequence numbers 1-01 through 1-04 (BSC 2001c, Section 5.3.2.1). During transfer from the dryer to the disposal container, individual spent nuclear fuel assemblies could be dropped or impacted as a result of a mechanical or control system failure of the dry assembly transfer machine or as a result of operator error. These event sequences would occur in the assembly transfer system cell, which is a confinement area with high-efficiency particulate air filtration. Individual spent nuclear fuel assembly event sequences in the cell are identified in Table 2-6 by sequence numbers 1-12, 1-13, and 1-14. The strategy is to confine particulate releases within the Waste Handling Building and maintain offsite radiological doses ALARA by using the high-efficiency particulate air filters in the heating, ventilation, and air conditioning system (BSC 2001c, Section 5.3.2.1).2.3.3.2.1.2 Spent Nuclear Fuel Assembly Basket Event Sequences
Spent nuclear fuel assembly baskets would first be handled underwater, during transfer out of the basket staging rack. From there, the assembly baskets, which contain a maximum of four pressurized water reactor spent nuclear fuel assemblies or eight boiling water reactor spent nuclear fuel assemblies, could be transferred and staged in the pool storage area to facilitate aging and blending or loaded directly into the incline transfer cart. Baskets that are staged in the pool area would have an additional movement step from the storage pool to the incline transfer cart. Once loaded onto the incline transfer cart, assembly baskets would be transported out of the pool and into the assembly drying stations, where up to six baskets could be loaded into each of the two assembly dryers. The assembly transfer system pool and cell would both be located in confinement areas with high-efficiency particulate air filtration (BSC 2001c, Section 5.3.2.2). Spent nuclear fuel assembly baskets could be dropped or impacted during lifts in the pools to the basket staging racks, transport to the pool storage area, or transport up the inclined transfer canal as a result of mechanical failures, control system failures, or operator error. Event sequences that occur underwater involving spent nuclear fuel assembly baskets are identified in Table 2-6 by sequence numbers 1-05 through 1-09 (BSC 2001c, Section 5.3.2.2). The primary safety strategy is to confine radionuclide particulate releases to the assembly transfer system pool water by designing the pool system in accordance with ANSI/ANS-57.7-1988, American National Standard Design Criteria for an Independent Spent Fuel Storage Installation (Water Pool Type). The water treatment system will provide the capability to filter radioactive material, purify the water, and remove floating debris from the surface of pools. Workers will be able to use vacuums to remove particles from pool walls and floors (DOE 2002a, Section 2.2.4.2.9). This same system provides the capability for cleanup of any radionuclide particulate releases into the pool water. In addition, spent nuclear fuel assembly baskets can be dropped or impacted onto the floor or into one of the assembly dryers as a result of mechanical or control system failure of the dry assembly transfer machine or operator error. Spent nuclear fuel assembly basket event sequences that occur in the cell are identified in Table 2-6 by event sequence numbers 1-10 and 1-11 (BSC 2001c, Section 5.3.2.2). The Category 2 event sequences evaluated in Preliminary Preclosure Safety Assessment for Monitored Geologic Repository Site Recommendation (BSC 2001c, Section 5.3.3) would occur as a result of drops or collisions among handling equipment, unsealed disposal containers, or unsealed transportation casks. The bounding Category 2 internal event sequences that are expected to result in radiological releases are identified in Table 2-7. DOE spent nuclear fuel, high-level radioactive waste, and immobilized plutonium waste forms were also considered, but it was determined that event sequences involving commercial spent nuclear fuel bounded the potential consequences of handling other waste forms. Spent Nuclear Fuel Assembly Basket Collision During Transfer—A spent nuclear fuel assembly basket collides with a wall or heavy object in the assembly transfer system pool, causing a breach and subsequent release. This event could occur during transfer either from the assembly basket rack to the pool area or from the pool area to the incline transfer cart. The pool water serves as a barrier to radionuclide particulate release, so only the radioactive gases are released to the Waste Handling Building environment (BSC 2001c, Section 5.3.3.1). The primary safety strategy is to confine radionuclide particulate releases within the assembly transfer system pool by designing the pool system in accordance with ANSI/ANS-57.7-1988. Uncontrolled Descent of Incline Transfer Cart—A remotely operated incline transfer cart containing a spent nuclear fuel assembly basket loses control during ascent up the incline transfer canal, which results in an uncontrolled descent and impact with the assembly transfer system pool, which causes a breach of the spent nuclear fuel assembly and subsequent release. The pool water serves as a barrier to radionuclide particulate release, so only the radioactive gases are released to the Waste Handling Building environment (BSC 2001c, Section 5.3.3.2). The primary safety strategy is to confine particulate releases within the assembly transfer system pool by designing the pool system in accordance with ANSI/ANS-57.7-1988. Handling Equipment Drop onto Spent Nuclear Fuel Assembly Basket in Pool—A lifting yoke (or other heavy object) is dropped onto an uncanistered spent nuclear fuel assembly in the assembly transfer system pool, causing a breach and subsequent release. The pool water serves as a barrier to particulate release; so only the radioactive gases are released to the Waste Handling Building environment (BSC 2001c, Section 5.3.3.3). The primary safety strategy is to confine radionuclide particulate releases within the assembly transfer system pool by designing the pool system in accordance with ANSI/ANS-57.7-1988. Handling Equipment Drop onto Spent Nuclear Fuel Assembly Basket in Cell—A lifting yoke (or other heavy object) is dropped onto an uncanistered spent nuclear fuel assembly in the assembly transfer system cell, causing a breach and subsequent release (BSC 2001c, Section 5.3.3.4). The strategy is to confine radionuclide particulate releases within the Waste Handling Building by relying on the high-efficiency particulate air filters in the heating, ventilation, and air conditioning system (BSC 2001c, Section 5.3.3.4). Unsealed Disposal Container Collision—A loaded, unsealed disposal container collides with a wall, shield door, or other heavy object, resulting in the release of a fraction of its radioactive material contents (BSC 2001c, Section 5.3.3.5). The strategy is (1) to confine radionuclide particulate releases within the Waste Handling Building by using the heating, ventilation, and air conditioning system's high-efficiency particulate air filters and (2) to provide design features (e.g., limit switches, redundant controls, emergency switch) and safe load paths that would minimize the likelihood of a collision that could result in a radioactive material release (BSC 2001c, Section 5.3.3.5). Unsealed Disposal Container Drop and Slapdown—A loaded, unsealed disposal container is dropped by the disposal container bridge crane onto a welding fixture or staging fixture. After dropping, the unsealed disposal container is presumed to slap down onto the floor and release a fraction of its radionuclide contents. The drop height for this event is the normal handling height in the disposal container handling cell (BSC 2001c, Section 5.3.3.6). The strategy is (1) to confine radionuclide particulate releases within the Waste Handling Building by using the heating, ventilation, and air conditioning system's high-efficiency particulate air filters and (2) to provide design features (e.g., limit switches for lift height, interlocks, redundant controls, redundant cables, physical restraints) that would minimize unsealed disposal container drops and potential radioactive material releases (BSC 2001c, Section 5.3.3.6). Handling Equipment Drop onto Unsealed Disposal Container—A lifting yoke (or other heavy object) is dropped onto a loaded, unsealed disposal container, resulting in the release of a fraction of its radioactive material contents (BSC 2001c, Section 5.3.3.7). The strategy is (1) to confine radionuclide particulate releases within the Waste Handling Building by using the heating, ventilation, and air conditioning system's high-efficiency particulate air filters and (2) to provide design features that would minimize handling equipment drops onto spent nuclear fuel assemblies inside a disposal container (BSC 2001c, Section 5.3.3.7). Unsealed Transportation Cask Drop into Cask Preparation Pit—A transportation cask, without impact limiters and with its lid unbolted, is dropped from the normal lift height into the cask preparation pit in the assembly transfer system pool area (BSC 2001c, Section 5.3.3.8). The strategy is (1) to confine radionuclide particulate releases within the Waste Handling Building by using the heating, ventilation, and air conditioning system's high-efficiency particulate air filters and (2) to provide design features that prevent or minimize cask drops (e.g., limit switches, interlocks, redundant control circuitry or cable restraints) or reduce the impact of a drop (e.g., shock absorber at base of pit) (BSC 2001c, Section 5.3.3.8). Unsealed Transportation Cask Drop into Cask Unloading Pool—A transportation cask, without impact limiters and with its lid unbolted, is dropped by the cask bridge crane into the assembly transfer system cask unloading pool (BSC 2001c, Section 5.3.3.9). The strategy is to confine radionuclide particulate releases within the assembly transfer system pool by designing the pool system in accordance with ANSI/ANS-57.7-1988. In addition, radionuclide particulate mitigation in the assembly transfer system pool area is provided by the secondary heating, ventilation, and air conditioning confinement ventilation system (BSC 2001c, Section 5.3.3.9).2.3.3.2.2 Internal Event Sequences with No Radioactive Material Release
For these event sequences, features of the design either prevent the event sequence from occurring or prevent a radionuclide release if the event occurs. Design features to prevent the event sequence can either physically prevent the event from occurring (e.g., by eliminating, at certain steps, the lifting of transportation casks or canistered waste), or reduce the event sequence frequency below the cutoff frequency of one in 1 million per year (e.g., by using redundant control features in cranes and control systems). Design features that prevent a release are based on the premise that Category 1 and Category 2 event sequences will occur and that affected structures, systems, and components must be designed to prevent the waste form from releasing radionuclides during such an event sequence. Prime examples of this include the waste package event sequences, which establish design bases for the waste package to ensure that the waste package will not breach as a result of Category 1 or Category 2 event sequences. Section 3.5 of Yucca Mountain Science and Engineering Report (DOE 2002a) provides waste package event sequence analyses. Preliminary Preclosure Safety Assessment for Monitored Geologic Repository Site Recommendation (BSC 2001c, Section 5.3.4, Table 5-7) identifies these events.2.3.3.2.3 Beyond Category 1 and Category 2 Event Sequences
Beyond Category 1 and Category 2 event sequences are those that have less than one chance in 10,000 of occurring before permanent closure. This corresponds to an annual frequency of less than 10-6 per year, based on an assumed preclosure lifetime of 100 years. Event sequences satisfying this frequency criterion are not within the frequency range for Category 1 and Category 2 event sequences in the definition of event sequence in 10 CFR 963.2 (66 FR 57298) and are not analyzed further. However, structures, systems, and components reducing event sequences below 10-6 per year are considered in the design basis. Appendix A in Preliminary Preclosure Safety Assessment for Monitored Geologic Repository Site Recommendation (BSC 2001c) considers the impact of lower-temperature operating modes on the identification of beyond Category 1 and Category 2 event sequences. The frequency of two events were found to be influenced by the choice of thermal operating modes. These events are aircraft crash into the surface facility and rockfall onto a waste package in the subsurface facility. Aircraft hazards are impacted by increases in the surface facility's size, which would accompany an operating mode in which spent nuclear fuel is aged before being emplaced underground. However, Appendix A4.2 of Preliminary Preclosure Safety Assessment for Monitored Geologic Repository Site Recommendation (BSC 2001c) considered the influence of the thermal operating modes on the surface facility size and concluded that the aircraft hazards are likely to remain beyond Category 1 and Category 2 event sequences. Rockfall onto a waste package in the subsurface becomes more likely with increases in the preclosure period, which would accompany an operating mode with extended forced ventilation. However, Appendix A4.1 of Preliminary Preclosure Safety Assessment for Monitored Geologic Repository Site Recommendation (BSC 2001c) considered the possible increase in the preclosure period and changes in the thermal operating modes on the drift temperature and concluded rockfall is likely to remain beyond Category 1 and Category 2 event sequences with design optimization (e.g., optimized ground support features, waste package emplacement strategy). Table 5-12 of Preliminary Preclosure Safety Assessment for Monitored Geologic Repository Site Recommendation (BSC 2001c, Section 5.4) identifies these events.2.3.3.3 Category 1 Event Sequence Consequences
Design Basis Event Frequency and Dose Calculation for Site Recommendation (BSC 2001e) evaluated the consequences of Category 1 event sequences. For offsite radiation doses, Category 1 event sequences are based on the following (BSC 2001e, Section 6.1.1):2.3.3.4 Category 2 Event Sequence Consequences
Design Basis Event Frequency and Dose Calculation for Site Recommendation (BSC 2001e) evaluated the consequences of Category 2 event sequences. Offsite radiation doses (uncontrolled areas) for bounding Category 2 event sequences were based on the following (BSC 2001e, Section 6.1.2):2.3.4 Preliminary Description of the Structures, Systems, Components, Equipment, and Operator Actions Intended to Prevent or Mitigate Accidents
This section provides a description of operator actions to mitigate or prevent accidents. At this stage in the design process, operator actions have not been explicitly factored into the event analyses. However, operator actions carried out according to prepared, reviewed, approved, and tested operating procedures are intended to prevent accidents. The overall design and operation of the repository will ensure the protection of both the repository workers and the public. Personnel who operate repository systems that are important to safety would be trained commensurate to their work assignment. They will follow procedures (developed, reviewed, and approved in accordance with the requirements of the quality assurance program) for operating equipment and facilities covered by the NRC license for the repository. In the Waste Handling Building, waste would be transferred from transportation casks to disposal containers. Fuel-handling operators would be trained consistent with approved procedures. Fuel movement and handling would be performed in accordance with formal, detailed, and approved operating procedures (Curry 2001, Section 2.4). Transportation casks containing uncanistered spent nuclear fuel or spent nuclear fuel in nondisposable canisters would be moved into a cask unloading pool, where the cask lid would be removed. Nondisposable canisters would have the canister lid cut off. The spent nuclear fuel assemblies would be removed from the cask or nondisposable canister and placed in metal baskets for storage in pools or moved directly to the spent nuclear fuel dryer cell for drying (BSC 2001c, Section 4.1). The placement (i.e., selective combining) of individual assemblies from the fuel blending inventory pools into a particular disposal container would be predetermined by engineering calculations that take into account the thermal output, reactivity, fuel assembly type, size, and compatibility of waste forms. Operators would fill the disposal containers according to the calculations. By precisely identifying each fuel assembly, operators would ensure that the blending is performed according to engineering calculations (Curry 2001, Section 2.4). To reduce the possibility of radioactive contamination from uncanistered assemblies, operators would remotely place the assemblies in disposal containers and attach inner lids in a shielded loading room. Before placing the inner lid on a disposal container, operators would independently verify that the assemblies placed into that particular disposal container are the ones identified by the blending calculations. The disposal containers would then be transferred to another room in the Waste Handling Building, where operators would weld the closure lids onto the disposal container. Activities for handling uncanistered spent nuclear fuel or spent nuclear fuel in nondisposable canisters would be remotely controlled, and would take place under water or in a shielded room that protects workers, the public, and the environment (Curry 2001, Section 2.4). The processes for handling canistered fuel assemblies (i.e., fuel assemblies contained in sealed canisters) and DOE high-level radioactive waste would be similar. Transportation casks containing canistered waste would be moved into a shielded room. Using remotely controlled equipment, operators would remove the lids from the casks and remove the canistered spent nuclear fuel assemblies or high-level radioactive waste from the transportation casks and place this waste into disposal containers or nearby staging racks to await codisposal with other high-level radioactive waste or canistered fuel assemblies (Curry 2001, Section 2.4). Precise canister identification would ensure disposal containers are loaded according to preestablished and verified loading instructions (Curry 2001, Section 2.4). After initial loading, operators would transfer the disposal container to the welding area of the Waste Handling Building. Workers would then weld the disposal container's inner, middle, and outer lids into place. The inner lid would be welded and inspected before the middle lid is welded. The middle weld would be welded and inspected and stress relieved before the outer lid is welded into place. Then the outer lid would be inspected, and stress will be relieved. A loaded, closed, welded, inspected, and certified disposal container is called a waste package. Operators would perform these steps using remotely controlled equipment (Curry 2001, Section 2.4). The final Waste Handling Building operation would involve decontaminating each waste package (if needed) and placing it in the transporter for delivery to the underground emplacement area. When the transporter is brought into the Waste Handling Building, the waste package would be rotated onto its side, lifted, and carried by a crane into a separate room for decontamination. Following decontamination, the waste package would be placed on a pallet and put on a rolling bedplate just outside the transporter's shielded area. A remotely controlled transfer mechanism would pull the rolling bedplate into the shielded transporter. Then the transporter would close its doors, and the waste package pallet assembly would be ready to be moved underground for emplacement via a rail system that connects the Waste Handling Building to the geologic repository (Curry 2001, Section 2.4). Repository workers would perform loading activities according to established procedures. Before loading the transporter, operators would identify each waste package and confirm that the waste package they are about to load is the waste package specified in the day's loading instructions. Use of the waste package transporter would provide adequate shielding to reduce external radiation to safe levels for workers in the vicinity, particularly those involved in transporting the waste package underground (Curry 2001, Section 2.4). Using the rail-supported transporter, each waste package would be transported to the subsurface drifts. When the transporter reaches the assigned emplacement drift, the loading mechanism would push the rolling bedplate with the waste package out of the transporter. If a waste package must be retrieved, the same loading mechanism could be used to pull the rolling bedplate and waste package back into the transporter at the emplacement drift entrance (Curry 2001, Section 2.4). The doors of the transporter would be remotely controlled (Curry 2001, Section 2.4). A remotely controlled emplacement gantry would lift the waste package/pallet unit from the rolling bedplate, carry it into the emplacement drift, and emplace it on its designated support structure within the drift. This gantry would be powered by a third rail conducting a direct-current power supply (Curry 2001, Section 2.5). The subsurface ventilation system would consist of two separate and independent fan systems and flow networks, separated by air locks that can be moved from position to position. One system would provide air to the development operations area, while another system would ventilate the waste emplacement operations area. Development of new emplacement areas and emplacement of waste in previously prepared areas would take place simultaneously over a period of approximately 20 years. Air pressure in the areas under development would be maintained at a higher level than the pressure in the emplacement side. In the unlikely event that radioactive particulates are released into the subsurface air stream on the emplacement side, the pressure differential would prevent the spread of these particles across the air locks to the area where the new drifts are being excavated (Curry 2001, Section 2.7). Radiation monitoring would be conducted in the waste emplacement operations areas, and any incremental increases in ambient radiation levels would be investigated to ensure that no radioactive contamination is released from the waste packages. Such a contamination event is unlikely, but is accounted for in the repository design (Curry 2001, Section 2.7).2.4 EVALUATION OF THE PRECLOSURE SUITABILITY CRITERIA
10 CFR 963.12 (66 FR 57298) calls for the DOE to evaluate the preclosure suitability of the Yucca Mountain site using the preclosure suitability criteria outlined in 10 CFR 963.14. These criteria are (1) the ability to contain radioactive materials and limit releases (10 CFR 963.14(a)); (2) the ability to implement control and emergency systems to limit exposure to radiation (10 CFR 963.14(b)); (3) the ability to maintain a system and components that perform functions important to safety (10 CFR 963.14(c)); and (4) the ability to preserve the option to retrieve waste during the preclosure period (10 CFR 963.14(d)). Sections 2.4.1 through 2.4.4 [2.4.1, 2.4.2, 2.4.3, 2.4.4] present the approach used to evaluate the potential repository using each criterion, a regulatory evaluation of the likelihood of meeting each criterion, and a summary of the preliminary conclusions.2.4.1 Ability to Contain Radioactive Material and to Limit Releases of Radioactive Materials
This criterion was evaluated in a preliminary preclosure safety assessment of the relevant features of the repository design. Those aspects of the design that are geared toward or support the function of containing radioactive material and limiting releases of radioactive materials are found in a number of documents. The primary reference used in the evaluation was Preliminary Preclosure Safety Assessment for Monitored Geologic Repository Site Recommendation (BSC 2001c). Other key references included Design Basis Event Frequency and Dose Calculation for Site Recommendation (BSC 2001e) and Final Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada (DOE 2002c).2.4.1.1 Evaluation
For the evaluation described in 10 CFR 963.13(a) (66 FR 57298), the ability to contain radioactive material and to limit releases of radioactive material was evaluated giving consideration to the ability of the potential repository design to be likely to meet the Category 1 and Category 2 event sequence licensing-related dose limits in the NRC's 10 CFR 63.111(a) and (b) and 10 CFR 63.204 (66 FR 55732), as referenced in the DOE guidelines at 10 CFR 963.13 and 10 CFR 963.2.2.4.1.1.1 Category 1 Event Sequence Consequences
Design Basis Event Frequency and Dose Calculation for Site Recommendation (BSC 2001e) evaluated the consequences of bounding Category 1 event sequences. Offsite radiation doses for bounding Category 1 event sequences and normal effluents and emissions were based on the following (BSC 2001e, Section 6.1.1):