An issue was raised by the public concerning which companies would manufacture transportation casks.Response Per standard government practice, the procurement contract for the transportation casks and waste packages would be open to all qualified bidders. The contracts would be awarded utilizing government procurement procedures. At this time, the DOE does not know who would be awarded these contracts. 4.5.2 (9222) Summary Comment
An issue has been raised by the public concerning the lifetime of the drip shields and waste packages.Response The DOE forecasts the lifetime of the titanium drip shield by utilizing an integrated degradation code with conservative parameters. A key element in the prediction of the long-term performance of the drip shield, and hence its lifetime, is the waste package degradation code. The waste package degradation code integrates the individual material degradation models. This code is described in the S&ER Rev. 1, Section 4.2.4. Both conservative and realistic versions were developed. The most conservative case analysis was to evaluate the effects of the conservative model abstractions of several key corrosion model parameters. Those parameters are localized corrosion and general corrosion parameters, along with corrosion rate bias to account for silicate deposits. This case represents the worst case combination of those parameters from the perspective of the first drip shield failure time. As shown in the SSPA, Volume 1, Section 7.4, the results of this case indicate that the earliest possible failure time of a drip shield for the upper bound profile is about 20,000 years, much earlier than the realistic case (about 50,000 years). The results also show that the initial failure comes from general corrosion. As shown in Section 4.2.4.4 of the S&ER, the results of this case indicate that the earliest possible failure time of a waste package from corrosion for the upper bound profile is about 12,000 years, much earlier than the realistic case (about 50,000 years). A supplemental analysis was performed that included the potential for early waste package failure. The analysis contained in Section 3.2.8.2 of the SSE assumed that a few (3 or less) waste packages could fail prior to 10,000 years due to assumed, undetected, improper heat treatment of the final closure weld. [BSC (Bechtel SAIC Company) 2001. "Total System Performance Assessment—Analyses for Disposal of Commercial and DOE Waste Inventories at Yucca Mountain—Input to Final Environmental Impact Statement and Site Suitability Evaluation." SL986M3 REV 00 ICN 01. Las Vegas, Nevada: Bechtel SAIC Company. ACC: MOL.20011114.0246. Section 5.2.4.2.]. The remaining waste packages would have lifetimes of about 80,000 years to over one million years.
4.5.3 Waste Form and Waste Package—Other
No comments received or comments addressed elsewhere.4.6 preCLOSURE design
4.6.1 Hypothetical Operational Accidents
4.6.1 (11) Summary CommentMembers of the public have raised the issue of a criticality-caused explosion at a Yucca Mountain repository, where circumstances are thought to being analogous to the Mayak Chemical Complex near Kyshtym, Russia, where a chemical explosion involving a radioactive liquid waste storage tank occurred in 1957. Members of the public have raised the concern that degradation of the waste package in the period following closure of the repository may lead to reconfiguring the spent nuclear fuel and high-level radioactive waste into potentially critical geometries. This long-term degradation of the waste form and package raises the concern that postclosure criticality may occur without appropriate design controls. The concern members of the public have raised over the possibility of an explosive nuclear criticality originates from a report by two DOE scientists at the Los Alamos National Laboratory, Dr. C.D. Bowman and Dr. F. Venneri, which concluded that an explosive nuclear criticality was credible [Bowman, C.D. and Venneri, F. 1995. "Underground Autocatalytic Criticality from Plutonium and Other Fissile Materials." LA-UR-94-4022. Albuquerque, New Mexico: Los Alamos National Laboratory. ACC: HQO.19950314.0027.]. Members of the public have raised the concern that enhanced material migration resulting from natural (earthquake, volcanism) external events could increase the likelihood of a criticality event.Response As stated in Section 4.3.3.2 of the S&ER Rev. 1, there is no potential event sequence in the preclosure period of the repository that would result in a criticality at a repository at Yucca Mountain. A criticality event would require the configuration of fuel with sufficient fissionable material to sustain a chain reaction. For the 10,000-year postclosure period., nuclear criticality is screened out from the TSPA nominal case analysis based on its probability being below that required for inclusion in the TSPA analysis. Nevertheless, because of significant interest in this topic, the DOE investigated the potential consequences of a criticality event. The results of the investigation indicate that a nuclear criticality would not have a significant impact on repository performance. Limiting the potential for, and consequences of, criticality during the postclosure phase of the geologic repository relies on multiple barriers, natural and engineered. The natural barrier system consists of the rock formations of the repository and includes the geologic, mechanical, chemical, and hydrological properties of the site. As defined in the NRC's licensing regulation at 10 CFR Part 63 [66 FR 55732], the engineered barrier system comprises the waste packages and the underground facility in which they are emplaced. A waste package is the generic term for describing the waste form (radioactive waste and any encapsulating or stabilizing matrix) and any containers, shielding, packing, and other absorbent materials immediately surrounding an individual package. The underground facility consists of the underground structure and openings that penetrate the underground structure (e.g., ramps, shafts, and boreholes, including their seals). The engineered barrier system would minimize the potential for conditions that would be conducive to a criticality event after the repository has been permanently closed. It is recognized that defense-in-depth (a design strategy based on a system of multiple, independent, and redundant barriers, designed to ensure that failure in any one barrier does not result in failure of the entire system) is needed against criticality events even if, as currently expected, the forecasted consequences of such events for the repository's performance and for the health and safety of the public would be very small. Therefore, scenarios and conditions that contribute significantly to the overall postclosure criticality risk would be examined, with an intent to incorporate reasonable measures (add or strengthen diverse or redundant barriers to criticality) to reduce the risk. Risk-informed, performance-based analysis would be used to determine the effectiveness of the measures [YMP (Yucca Mountain Site Characterization Project) 2000. "Disposal Criticality Analysis Methodology Topical Report." YMP/TR-004Q, Rev. 01. Las Vegas, Nevada: Yucca Mountain Site Characterization Office. ACC: MOL.20001214.0001.]. An important aspect of defense-in-depth involves taking advantage of the many natural and engineered features of the site and repository to make the probability and consequences of postclosure criticality as low as feasible. The engineered barriers would collectively make the probability of a postclosure criticality low. For a criticality to occur, multiple changes in conditions (waste package breach, water intrusion and retention, removal of neutron absorbers) must occur. Should a criticality occur, however, the confining features of the underground natural barriers would protect against releases of radionuclides to the accessible environment. The features eventually implemented are expected to provide barriers to postclosure criticality that are both diverse (dissimilar methods to limit susceptibility to common-mode failures) and redundant (multiple barriers performing the same function that reduces the probability of criticality). Examples of diverse barriers are the waste package inner barrier, neutron-absorbing materials in the basket, and the steel (which displaces moderator) in the basket materials. Similarly, the use of two separate barriers (waste package and drip shield) to impede entry of water into the waste form is an example of the use of redundant barriers. The waste package itself impedes entry of water into the waste form, and the drip shield limits or prevents damage to the waste package from dripping water or rockfall [Ibid.]. Appendix H of the FEIS examines repository safety by evaluating a spectrum of credible radiological accidents and estimating their impacts. Section 5 and Appendix H, Section 2, deal specifically with the issue of criticality for repository operations and after repository closure, respectively. In both cases, the analysis concludes that criticality accidents would be extremely unlikely, and if they occurred, the impacts would not be significant. The report by Bowman and Venneri [Bowman, C.D. and Venneri, F. 1995. "Underground Autocatalytic Criticality from Plutonium and Other Fissile Materials." LA-UR-94-4022. Albuquerque, New Mexico: Los Alamos National Laboratory. ACC: HQO.19950314.0027.] dealt with autocatalytic criticality. As reported in Section 4.3.3.2.3 of the S&ER Rev. 1, this type of criticality has been found to be not credible [Paperiello, C.J.1995. "Review of Potential for Underground Autocatalytic Criticality." Letter from C.J. Paperiello (NRC) to L. Barrett (DOE/OCRWM), August 7, 1995, with enclosure. ACC: HQO.19950912.0002.]. Autocatalytic criticality is not possible for low-enriched waste forms like commercial spent nuclear fuel, nor is it possible for the waste form inside the waste package. Even for highly enriched waste forms or those containing nearly pure plutonium-239 (which excludes commercial spent nuclear fuel), achieving a critical mass outside a waste package would require the entire fissile content of the waste package to be spread uniformly in a nearly spherical shape. In addition, it would require the extremely unlikely commingling of large amounts of transported fissile material from at least two waste packages containing highly enriched waste forms [Canavan, G.H.; Colgate, S.A.; Judd, O.P.; Petschek, A.G.; and Stratton, T.F. 1995. "Comments on Nuclear Excursions and Criticality Issues." LA-UR: 95 0851. Los Alamos, New Mexico: Los Alamos National Laboratory. ACC: HQO.19950314.0028.] [CRWMS M&O 1996. "Probabilistic External Criticality Evaluation." BB0000000-01717-2200-00037 REV 00. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.19961029.0024.]. Because the igneous rock at Yucca Mountain is not likely to contain deposits that can efficiently accumulate fissile material, the probability of creating such a critical mass from a single or multiple waste packages containing highly enriched waste forms is so low as to be not credible [Kastenberg, W.E.; Peterson, P.F.; Ahn, J.; Burch, J.; Casher, G.; Chambre, P.L.; Greenspan, E.; Olander, D.R.; Vujic, J.L.; Bessinger, B.; Cook. N.G.W.; Doyle, F.M.; and Hilbert, L.B., Jr. 1996. "Considerations of Autocatalytic Criticality of Fissile Materials in Geologic Repositories." Nuclear Technology," volume 115, pages 298-308. Hinsdale, Illinois: American Nuclear Society. TIC: 247504.]. There have been a number of criticality accidents in Russia since 1953 resulting in a total of seven worker deaths (no deaths to members of the public). However, none of these involved handling or burial of radioactive waste. The event that occurred at the Mayak Chemical Complex near Kyshtym, Russia, in the Ural Mountains, on September 29, 1957, was a chemical explosion involving a radioactive liquid waste storage tank, not a criticality accident. There are no liquid radioactive wastes or explosive chemicals allowed for disposal at the proposed Yucca Mountain repository. Therefore, events similar to this one are not possible at Yucca Mountain and this event is not relevant to repository safety. The potential impact of disruptive natural events (e.g., seismic or igneous intrusion) on the risk of criticality in the repository has been studied. Seismic events can produce a rapid change in the configurations of waste forms and waste packages. If the resulting configuration has a neutron multiplication factor, k-effective, above the critical limit, the seismic event has provided a rapid reactivity insertion mechanism that could lead to a transient criticality. The identification of the initial, pre-insertion configurations has been incorporated into the repository criticality analysis methodology.
4.6.2 Accident Event Sequences
4.6.2 (8) Summary CommentAn issue has been raised about the reliability of the accident probabilities.Response The DOE has established criteria for the consideration of possible events associated with a repository at Yucca Mountain that are consistent with relevant regulations, and consequences of event sequences applicable to a repository at Yucca Mountain have been calculated. As discussed in Chapter 2 of the SSE, event sequences determined to be applicable to a repository at Yucca Mountain cover a full range of probable events, from normal operational events that might be reasonably anticipated to occur during the design life of the facility to very low-probability events that are not expected to occur. The probability of a preclosure operational event sequence is based on the frequency of occurrence. Event sequences are categorized as Category 1 or Category 2, as described in 10 CFR 63.2 as referenced in 10 CFR 963. Category 1 event sequences are expected to occur one or more times before permanent closure of the repository. Based on a preclosure operating period of 100 years, the Category 1 event sequence frequency is 0.01 events per year or higher. Other event sequences that have at least one chance in 10,000 of occurring before permanent closure are defined as Category 2. Based on a preclosure operating period of 100 years, the frequency range is between 0.01 and 0.000001 per year. Event sequences that are expected to occur at a frequency of less than once in 10,000 events, based on a 100-year preclosure operational period, are considered to be not credible. Category 2 event sequences are low probability events and the radiological consequences of these events have been calculated as described in Chapter 2 of the SSE. The Category 1 and Category 2 event sequence frequencies were calculated based on a 100-year preclosure operational period. If the closure of the repository extended to 325 years, the use of a 100-year preclosure period to screen event sequences would be unchanged since the surface fuel handling operations would be completed after approximately 24 years. There would be no waste forms in the surface facility after the waste package subsurface emplacement operations are completed. Chapter 2 of the SSE notes that extension of the preclosure operations to 325 years would impact the screening criteria for subsurface event sequences; but no new event sequences have been identified that would impact the selection of bounding event sequences that could result in a radioactive release. The DOE calculation method of event sequence probabilities indicates that a repository at Yucca Mountain would be likely to be below the NRC's radiation protection standards. 4.6.2 (18060) Summary Comment
An issue has been raised concerning the effects of earthquakes on surface dry storage containers, fuel storage pools, and underground tunnels.Response Spent fuel storage casks may be employed to accommodate the inventory of spent fuel required to meet operating requirements with regard to the thermal load of waste packages. These casks would be required to meet all applicable NRC criteria, including the ability to adequately cool the fuel in the event of an earthquake and subsequent tipover. The Pool Fuel Blending Inventory Building and the associated fuel blending inventory pools were described in Section 2.2.4.2.2 of the S&ER and Section 2.2.4.2.2 of the S&ER Rev. 1. The potential hazards from the fuel blending inventory pools and the surface aging facility (i.e., design options to support a range of thermal operating modes) were evaluated in Sections 5.3.2.1 and 5.3.2.3 of the S&ER Rev. 1, and were found to have no effect on the selection of bounding event sequences that result in radionuclide release. Accidents involving the spent fuel storage modules in a surface aging facility are evaluated in Appendix H of the FEIS. An earthquake event could cause the modules to tip over, but the storage canisters and welded seams would withstand such tipovers without damage. A tipover of a storage module would not block all cooling vents, and the spent nuclear fuel would not overheat. A surface aging facility would comply with all applicable NRC licensing requirements, which would include seismic design criteria specific to the repository. The DOE and others have determined the design basis vibratory ground motion frequency and fault displacement appropriate for design of a repository at Yucca Mountain. Structures, systems, and components important to radiological safety would be designed to withstand the effects of a design basis earthquake without posing a threat to public health and safety. Seismic design basis vibratory ground motion frequency and fault displacement in the vicinity of the Yucca Mountain site have been the focus of a great deal of study by the DOE and others. As stated in Chapter 2 of the SSE, the DOE seismic design methodology and the earthquake occurrence frequencies, fault displacement, and vibratory ground motion hazards in the Yucca Mountain vicinity to be used for design of structures, systems, and components determined to be important to radiological safety and waste isolation was established by "Preclosure Seismic Design Methodology for a Geologic Repository at Yucca Mountain" [Yucca Mountain Site Characterization Project 1997. "Preclosure Seismic Design Methodology for a Geologic Repository at Yucca Mountain." Topical Report YMP/TR-003-NP, Rev. 2. Las Vegas, Nevada: Yucca Mountain Site Characterization Office. ACC: MOL.19971009.0412.]. As described in Chapter 2 of the SSE, the structures, systems, and components important to radiological safety would be designed to withstand the effects of a design basis earthquake. For licensing the NRC requires in 10 CFR Part 63 that a repository be designed and constructed to withstand a design basis earthquake without posing a threat to public health and safety. The design and construction attributes necessary to ensure that structures, systems, and components would not be compromised during a seismic event are well understood and would be applied to the repository. In addition, in Chapter 2 of the SSE, it is noted that earthquake ground motion at the repository level would be significantly less than on the ground surface, and underground repository structures would be designed for appropriate ground motion levels. The repository would be designed and constructed to comply with applicable codes, standards, and NRC regulations for licensing to withstand the design basis earthquake. For example, Chapter 2 of the S&ER Rev. 1 states that the emplacement gantry would be designed to prevent the drop of a waste package as a result of a design basis earthquake; and that the cranes and hoists in the canister transfer system would be designed to remain on their rails during and following a design basis earthquake.
4.6.3 Terrorist Attack and Sabotage Impacts
4.6.3 (55) Summary CommentMembers of the public have raised the concern that there exists the potential for sabotage at a repository at Yucca Mountain. This concern is raised for both the preclosure and postclosure periods. Members of the public have expressed the concern that a repository would represent a possible target for terrorists given the inventory of radioactive material stored. They have expressed the concern that the DOE failed to analyze the potential for a terrorist attack utilizing nuclear weapons on Yucca Mountain. Members of the public have expressed the concern that a nuclear warhead could be placed in a waste package and made to detonate after the waste has been permanently stored. In addition, they have expressed the concern that high-level radioactive waste and spent nuclear fuel stored on the surface at the a repository at Yucca Mountain could become targets for air attacks.Response Spent nuclear fuel and high-level radioactive waste would be permanently entombed in a sealed geologic repository at Yucca Mountain, a remote location in an area of low population density. Postclosure access to the material by intruders would be extraordinarily difficult. In addition, for licensing, the NRC (10 CFR 63.21 and 10 CFR 73.51) requires any repository at Yucca Mountain to have preclosure physical protection. These regulations specify a performance objective, which provides "high assurance that activities involving spent nuclear fuel and high-level waste do not constitute an unreasonable risk to public health and safety." The regulation requires that spent nuclear fuel and high-level radioactive waste be stored in a protected area such that: (1) access to the material would require passage through or penetration of two physical barriers: the outer barrier would have isolation zones on each side to facilitate observation and threat assessment, would be continually monitored, and would be protected by an active alarm system; (2) adequate illumination would be provided for observation and threat assessment; (3) the area would be monitored by random patrol; and (4) access would be controlled by a lock system and personnel identification would be used to limit access to authorized persons. A trained, equipped, and qualified security force would be required to conduct surveillance, assessment, access control, and communications to ensure adequate response to any security threat. Liaison with a response force would be required to permit timely response to unauthorized entry or activities. In addition, 10 CFR Part 63 requires (by reference to 10 CFR Part 72) that comprehensive receipt, periodic inventory, and disposal records be kept for spent nuclear fuel and high-level waste in storage. The DOE would comply with any revisions to the licensing-related regulations concerning proposed protection and security. Although it is not possible to predict whether sabotage events would occur, and if they did, the nature of such events, the DOE examined various accident scenarios, which provide an approximation of the types of consequences that could occur (see FEIS, Appendix H).
4.6.4 External Hazard Impacts
4.6.4 (10) Summary CommentMembers of the public have raised a concern that possible resumption of underground nuclear weapons testing or other operations at the Nevada Test Site could adversely affect repository operations or long-term isolation of spent nuclear fuel and high-level radioactive waste. In addition, members of the public are concerned that other nearby military or industrial activity may also affect repository operations.Response NRC guidance typically requires applicants to include in their submittals a section addressing "nearby military and industrial activities." The purpose of this section is to identify such activities that might pose a risk to the safe operation of the facility for which a license is sought. Operations on the Nevada Test Site and Nellis Air Force Range, and operations at nearby facilities outside the Nevada Test Site were found to include no activities that would impact the preclosure or postclosure safety. No military or commercial industrial activities would be conducted within 8 kilometers (5 miles) of the Yucca Mountain site. The NRC guidance in NUREG-0800 [NRC 1987. "Design of Structures, Components, Equipment, and Systems." Chapter 3 of "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants." NUREG-0800. Washington, D.C.: U.S. Nuclear Regulatory Commission. TIC: 203894.] specifies that facilities and activities at distances greater than 8 kilometers (5 miles) should be considered for consequence analysis if these activities have the potential for affecting nuclear safety-related features. The area surrounding an 8-kilometer (5-mile) radius of a repository at Yucca Mountain would include the balance of the Nevada Test Site, land owned or controlled by the U.S. Air Force and the U.S. Bureau of Land Management, and the balance of any land withdrawal for a repository. Chapter 8 of the FEIS contains an evaluation of the military or industrial activities that are conducted at distances greater than 8 kilometers (5 miles) from the Yucca Mountain site to determine their potential impact on a repository at Yucca Mountain. The Nevada Test Site and Nellis Air Force Range and were found to have no activities that would impact preclosure or postclosure safety. The remote location of a repository at Yucca Mountain (more than 8 kilometers [5 miles] from Nevada Test Site facilities, more than 20 kilometers [13 miles] from nearby commercial industrial operations and U.S. Highway 95, and more than 32 kilometers [20 miles] from Nellis Air Force Range facilities) is the major reason there would be no impacts to a repository. Proposed activities (e.g., Kistler Aerospace Corporation Launch Operations) at the Nevada Test Site cannot be fully evaluated because of a lack of design information. When these operations are further developed and information becomes available, an evaluation of potential event sequences from these operations would be performed. The "Final Environmental Impact Statement for the Nevada Test Site and Off-Site Locations in the State of Nevada" [DOE 1996. "Final Environmental Impact Statement for the Nevada Test Site and Off-Site Locations in the State of Nevada." DOE/EIS 0243. Las Vegas, Nevada: U.S. Department of Energy, Nevada Operations Office. MOL.20010727.0190 through MOL.20010727.0191.] was used to identify current and future planned facilities and activities on the Nevada Test Site. Projects and activities covered in the environmental impact statement include the defense, waste management, environmental restoration, nondefense research and development, work for others programs, and Nevada Test Site support activities. Nuclear stockpile stewardship at the Nevada Test Site includes no nuclear yield (sub-critical) nuclear weapons testing and science-based weapons experimentation. The locations for these tests are within either the Nuclear Test Zones or the Nuclear and High Explosive Test Zones. The closest point of these zones to the repository is approximately 24 kilometers (15 miles). Weapons testing is conducted either in vertical drill holes or in underground tunnels located in the general area of past weapons tests. The environmental impact statement evaluation of potential impacts from underground testing at the Nevada Test Site [Ibid.] concluded that the only impact such testing could impose on a repository at Yucca Mountain would be ground motion associated with the energy released from the detonation of conventional high explosives. Even if nuclear weapon testing were to resume, the ground motion effects were determined not to exceed the seismic design criteria. In other words, the design basis earthquake for a repository was determined to provide the greatest ground motion effects. Therefore, since a repository would be designed to survive the design basis earthquake, it would also survive ground motion from any nuclear tests. Some military aircraft participating in activities at the Nellis Air Force Range and the Tonopah Test Range carry explosive ordnance. However, this ordnance is not armed until the aircraft arrives at the test range. Any ordnance that was armed but was not dropped cannot be carried over the Nevada Test Site. Consequently, an aircraft crash near the repository would not be expected to detonate any ordnance carried by the aircraft. No event sequences that could cause a radiological release as a result of any military sorties flying out of Nellis Air Force Range and the Tonopah Test Range were determined to be within a Category 1 or Category 2 event sequence. However, many of the parameters used to quantify the frequency of these events (e.g., fraction of flights that would pass near the Waste Handling Building and probability of objects or armament dropping from an aircraft) would be reevaluated during the preparation of any license application. The existing analyses of potential military or industrial hazards to a repository sited at Yucca Mountain have provided an understanding of the potential risks and any further analyses that may be needed to adequately support a license application. If an aircraft crash is subsequently determined to be within a Category 1 or Category 2 event sequence, or if there are any new industrial developments that warrant consideration, the potential consequences would be calculated considering all pertinent conditions (e.g., explosives or quantity of jet fuel carried on the aircraft), and applicable design measures would be evaluated during the preparation of any license application. Operations on the Nevada Test Site and Nellis Air Force Range, operations at nearby facilities outside the Nevada Test Site, and transportation routes were examined in "Industrial/Military Activity-Initiated Accident Screening Analysis" [CRWMS M&O 1999. "Industrial/Military Activity-Initiated Accident Screening Analysis." ANL-WHS-SE-000004 REV 00. Las Vegas, Nevada: CRWMS M&O. Sections 7 and 8. ACC: MOL.20000307.0381.] and found to have no events that would impact preclosure or postclosure safety. 4.6.4 (46) Summary Comment
Members of the public have expressed concerns about why the DOE would accept foreign research reactor spent nuclear fuel and other spent nuclear fuels from foreign commercial reactors that would be disposed at a repository at Yucca Mountain.Response Consistent with the Nuclear Non-Proliferation Act and as described in the "Final Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuels" [DOE 1996. "Final Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel." DOE/EIS-0218F. Washington, D.C.: U.S. Department of Energy. MOL.20010727.0195 and MOL.20010727.0196.], the United States has begun bringing back highly enriched uranium created in the United States and leased to other countries in the 1950s. The purpose of bringing the material back to the United States is to support the nuclear weapons nonproliferation policy. The only foreign research reactor spent nuclear fuel that may be emplaced at Yucca Mountain would be material for which the United States retains ownership. The Nuclear Non-Proliferation Act also addresses the need to increase the effectiveness of international safeguards and controls on peaceful nuclear activities to prevent proliferation. In addition, the United States can accept limited quantities of foreign spent nuclear fuel without Congressional approval if the President determines that an emergency situation requires acceptance and is in the national interest, and notifies Congress with a detailed explanation and justification. 4.6.4 (15711) Summary Comment
An issue has been raised concerning the Waste Handling Building Ventilation system and its quality safety classification.Response The Waste Handling Building confinement area ventilation and the emergency power source and distribution systems are classified as Quality Level-2. The quality classification process for a repository is applied to structures, systems, and components that are "important to safety" and "important to waste isolation." The Waste Handling Building confinement area ventilation and the emergency power source and distribution systems have been determined to be important to safety. Structures, systems, and components are assigned quality level classifications that represent the relative importance of a structure, system, or component to the health and safety of the public or to the radiological safety of workers. Because the NRC developed the licensing regulation for a repository at Yucca Mountain (i.e., 10 CFR Part 63) as a risk-informed, performance-based rule, the quality classification criteria are derived in a risk-informed framework. 10 CFR Part 63 allows the DOE to categorize (or assign different levels of quality assurance to) structures, systems, or components whose failure to function has different risk of dose implications. Part of the classification criteria for licensing is quantitative and is derived from the 10 CFR Part 63 preclosure criteria for offsite doses and worker doses. Limiting dose criteria are imposed on different categories of event frequencies such that they become risk-informed performance criteria. These risk-informed regulatory criteria are an essential element in the quality classification. Models and results of the preclosure safety analysis are used to assess the change in frequency and/or consequences that occur when a given structure, system, or component is assumed unavailable (e.g., failed). Proposed 10 CFR Part 63 used the term "design basis events." The S&ER and the SSE referred to Category 1 and Category 2 design basis events. The final version of 10 CFR Part 63 has replaced the term "design basis events" with "event sequences." There is no difference between Category 1 and 2 design basis events and Category 1 and 2 event sequences. Category 1 event sequences are expected to occur one or more times before permanent closure of the repository. Category 2 event sequences have at least 1 chance in 10,000 of occurring before permanent closure. The DOE has assumed a 100-year preclosure operating period, although the preclosure period may be up to 325 years. Important to safety, with reference to structures, systems, or components, means those engineered features of a geologic repository operations area whose function is (1) to provide reasonable assurance that high-level waste can be received, handled, packaged, stored, emplaced, and retrieved without exceeding the licensing requirements of 10 CFR 63.111(b)(1) for Category 1 event sequences (i.e., 15 millirem total effective dose equivalent); or (2) to prevent or mitigate Category 2 event sequences that could result in radiological doses exceeding the values specified at 10 CFR 63.111(b)(2) to any individual located on or beyond any point on the boundary of the site (i.e., 5 rem total effective dose equivalent). The potential dose for Category 1 event sequences is calculated as if the structure, system, or component being evaluated fails when called upon to mitigate consequences. The unmitigated dose resulting from the removal of the structure, system, or component is added to aggregate offsite doses for Category 1 event sequences. This approach provides a risk-informed basis for classifying each structure, system, or component. If the revised dose exceeds 15 millirem per year total effective dose equivalent but is less than 100 millirem per year total effective dose equivalent, the structure, system, or component is classified as Quality Level-2. If the revised dose exceeds 100 millirem per year total effective dose equivalent, the structure, system, or component is classified as Quality Level-1. The rationale for classifying the Waste Handling Building ventilation system as Quality Level-2 includes (1) the 15-millirem per year limit is a constraint on this potential source of radiation exposure, and (2) compliance would be monitored during any repository operations (i.e., practices that produce doses in excess of the 15-millirem limit may be subject to corrective actions). Compliance analyses for Category 2 event sequences would demonstrate, for each event sequence assessed individually, that the offsite dose is less than 5 rem total effective dose equivalent. The classification analyses reassess the dose after a structure, system, or component functional failure is performed. If the dose exceeds 5 rem, the structure, system, or component is classified Quality Level-1; if the dose is less than 5 rem but greater than 100 millirem the structure, system, and component is classified Quality Level-2; if the dose is less than 100 millirem but greater than 15 millirem, the structure, system, or component is classified Quality Level-3; otherwise the structure, system, or component is not subject to the requirements of the quality assurance program and is classified as conventional quality. No event sequence has been identified that would exceed the 100 millirem total effective dose equivalent classification guidance for the Waste Handling Building ventilation system; therefore, the ventilation system is not required to mitigate or prevent event sequences that could exceed 100 millirem without use of the building's ventilation system. Portions of the ventilation system are conservatively classified as Quality Level-2 based on preventing an offsite dose greater than 15 millirem per year. As noted in the previous paragraphs, the Waste Handling Building ventilation system is classified as Quality Level-2. Therefore, the emergency power source and distribution system is also correctly classified as Quality Level-2 because its function is to support the Waste Handling Building ventilation system and failure of the emergency power source and distribution system would not result in an offsite dose that exceeds 100 millirem total effective dose equivalent classification guidance. 4.6.4 (11810) Summary Comment
An issue has been raised concerning whether the DOE would notify the community in the event of a radioactive release.Response In accordance with NRC licensing regulations, an emergency response plan would be developed for a repository at Yucca Mountain that would include provisions for notification of the public of radiological releases to the environment. The design of a repository would include radiological monitoring systems to ensure the radiological safety of the public and the environment. Operational controls that would be implemented include a radiological environmental monitoring and surveillance program that would provide assurance that the facility is functioning as intended to limit releases to the environment and evidence that the public dose is as low as is reasonably achievable. The program would measure radiation and concentrations of radionuclides from facility operations including those that are most likely to result in exposure to members of the public.
4.7 postCLOSURE SAFETY and total system performance assessment
4.7.01 Waste Package and Drip Shield Degradation
4.7.01 (43) Summary CommentIssues have been raised by members of the public regarding the chemical environment around the waste package and drip shield. These issues are grouped into three categories and include: concern that the DOE does not understand the chemical environment around the waste packages which impacts waste package materials testing and the ability to predict long term waste package performance; the potential for microbial attack of the waste package from microorganisms in the water; and the potential long term release of chromium into the environment following waste package corrosion.Issue Issues were raised by members of the general public that the DOE does not fully understand the chemical environment in the repository that should be utilized for waste package materials testing, including the presence of salts that could come in contact with the waste packages. Response The chemistries of waters chosen for testing have a broad range that duplicate the most challenging environmental conditions expected within the repository. The environment of the waste packages changes with time. Upon emplacement, the waste packages would still be fairly warm and the air would be low in relative humidity, assisted through the use of active ventilation. Just prior to repository closure, the drip shields would be inserted. Thereafter, the waste packages would continue to cool and the relative humidity would rise to near 100 percent. Thus, moisture is likely to condense on the drip shields and the waste packages, although the drip shields would prevent water from dripping directly from the roofs of the emplacement drifts onto the packages for thousands of years. This dripping water could be concentrated in its contained salts due to evaporative condensation processes or by the presence of deliquescent salts brought in by the ventilation system. Two types of water were evaluated (S&ER Rev. 1, Section 4.2.4.2.4) which cover the major sources of water potentially contacting the waste packages. These include the J-13 well water (a bicarbonate water) and rock pore water (a chloride-sulfate water). Analytical modeling as well as evaporation tests were conducted to estimate the compositions of the waters and salts as dryness was approached. The J-13 water formed salts containing chlorides, nitrates, carbonates and silicates. Calcium, as well as magnesium, precipitates as carbonates early in the evaporation process, while chlorides and nitrates, being more soluble, precipitate when evaporation is nearly complete. These salts have a deliquescent (liquefying by water absorption from air) point of about 50 percent relative humidity at 120 degrees Celsius (248 degrees Fahrenheit) with a resulting concentrated solution pH of about 12. The rock pore water was also evaluated using this process. Here chloride and sulfate salts predominated with a lower deliquescent point and a concentrated solution pH of about 6, which is near neutral. Thus, the water chemistry resulting from the near evaporation of J-13 water was used as a basis for subsequent corrosion testing because this chemistry is considered to be more aggressive than that of the concentrated pore water. However, testing with pore-type waters is planned. The likely concentrations of heavy metals were also included in the testing program. Although it is recognized that at very long repository times, the waters contacting the waste package would be more dilute, the use of concentrated solutions throughout the entire period results in conservative estimates. Several topics were identified and addressed during discussions between the DOE and the NRC on the effects of corrosion processes on the lifetimes of the containers. The topics include the credible range of brine water chemistry, the effect of introduced materials, the concentration and effect of minor (or trace) elements, the characteristics of the evolution of the types of brine waters, and an evaluation of periodic water drip evaporation. The DOE has agreed to update documentation regarding these chemistry issues in a revision to the analysis model reports on environment on the surfaces of the drip shield and waste package outer barrier and on general and localized corrosion of the drip shield. Issue An issue has been raised by a member of the general public regarding the extent of de-alloying and corrosion enhancement due to the presence of microbes in Yucca Mountain. Response Microbial corrosion, also called microbially influenced corrosion, is possible if the right combination of microbes, nutrients, temperature and relative humidity exist. Microbially influenced corrosion has been studied by the DOE by utilizing combinations of microbes found at Yucca Mountain. Microcosm tests were conducted to determine which nutrients limited the growth of microbes. The low level of phosphate in J-13 well water, the water assumed representative of that contacting the waste packages, appears to be the principal nutrient-limiting factor to microbial growth. In corrosion tests, the nickel-base alloy, Alloy 22, has shown very good resistance to attack against a consortium of Yucca Mountain microbes. Thus, the microbes would not cause accelerated corrosion of the waste packages. However, some evidence for de-alloying, or selective dissolution, perhaps from welded areas, has been observed. Thus, for conservatism, a multiplying factor of two was added to general and localized corrosion rates. For further detail, see Section 4.2.4.3.3 of the S&ER Rev. 1. As a result of discussion between the DOE and the NRC on the effects of corrosion on the life times of waste packages, the DOE would provide documentation on microbial effects and the enhancement factor in a revision to the analysis model report on general and localized corrosion of waste package outer barrier. Issue Issues have been raised by members of the public regarding the analysis performed by the DOE on the potential release of chromium from the waste packages to the environment. Response The DOE has evaluated the potential for release of chromium from the dissolution of the chromium-containing Alloy 22 waste package and the stainless steel inner container materials. Corrosion testing has shown the dissolution rate of Alloy 22 and stainless steel under a wide range of conditions is very low, less than one micron per year. However, as a result of the dissolution process, a small amount of hexavalent chromium is put into solution that ultimately could reach the accessible environment. The concentrations of chromium released from waste packages that potentially could reach the accessible environment have been calculated utilizing conservative assumptions. These have yielded concentrations well below EPA's maximum contaminant level goal. 4.7.01 (44) Summary Comment
Issues have been raised by members of the public regarding the forecast of long term waste package performance. These issues are grouped into four categories and include: using short-term data for making long-term performance forecasts; calculations for forecasting long-term waste package performance; confidence in making long-term performance forecasts; and the differences in long-term waste package performance under hot or cold repository operating conditions. (For the purposes of describing testing programs, "long term" refers to tests that may be conducted over a period of tens of years through closure of a repository. For all other purposes, the term "long term" refers to durations of 5,000 to 10,000 years or greater.)Issue An issue has been raised by the public that long-term forecasts were inappropriately based on short-term data and questioning how the DOE would augment the existing database to provide some credibility for long-term lifetime forecasts. Response The DOE recognizes that forecasts are based on short-term data. Thus, the DOE would support the forecast of the long-term performance by continuing tests as performance confirmation tests. These efforts would be augmented by the consideration of natural and engineering analogues. For the purposes of describing testing programs, "long term" refers to tests that may be conducted over a period of tens of years through closure of a repository. For all other purposes, the term "long term" refers to durations of 5,000 to 10,000 years or greater.) The DOE recognizes that current forecasts of the lifetime of containers are based upon its short-term and ongoing long-term test data, as well as data in the literature. Very long-term data under Yucca Mountain conditions are not yet available. Thus, the DOE intends to support forecasts of lifetimes of container materials by a combination of long-term testing, in situ and other performance confirmation tests, and the use of analogues, consistent with the guidance provided in the American Society for Testing Materials C 1174-97 standard. [ASTM C 1174-97. 1998 "Standard Practice for Prediction of the Long-Term Behavior of Materials, Including Waste Forms, Used in Engineered Barrier Systems (EBS) for Geological Disposal of High-Level Radioactive Waste." West Conshohocken, Pennsylvania: American Society for Testing and Materials. TIC: 246015.]. Testing would continue during waste emplacement and preclosure to collect long-term test data under conditions prototypical of those expected at Yucca Mountain. The data generated would continue to go to the scientists and engineers who determine the long-term performance of the materials as part of the determination of total system performance. The DOE would actively monitor the waste packages and the conditions within the drifts following the start of emplacement operations to detect any significant changes from baseline conditions, to confirm that subsurface conditions are consistent with the assumptions used in performance analyses, and to confirm that barrier systems and components are operating as expected. Coupons of container materials, as well as waste packages, could be retrieved and examined, if necessary. Details of the performance confirmation program can be found in Section 4.6.1.1 of the S&ER Rev. 1. The DOE is exploring analogues of Alloy 22 to provide additional confidence in its performance. One such analogue is josephinite, a nickel-iron mineral found in some streambeds. While this material contains no chromium, an important element in Alloy 22, it does have a stable passive film. A general discussion of natural analogues can be found in the S&ER Rev. 1, Section 2.1.5.4. These efforts plus the further development of a long-term mechanistic model for long-term general and localized corrosion and passive film stability provide confidence in the DOE's ability to forecast the long-term performance of waste package materials and, hence, the long-term container lifetime. Issue Issues have been raised by members of the public regarding the expected lifetime of the waste packages and how that expected lifetime was calculated. Response A key element in the prediction of the long-term performance of the waste package outer barrier, and hence the lifetime of the waste package, is the waste package degradation code. The Waste Package Degradation code integrates the individual material degradation models and corrosion test data. The code is described in the S&ER Rev. 1, Section 4.2.4. Both conservative and realistic versions have been developed. The most conservative case analysis was to evaluate the effects of the conservative model abstractions of several key corrosion model parameters. Those parameters are stress corrosion cracking-related parameters and general corrosion parameters, along with corrosion rate bias to account for silicate deposits. This case represents the worst case combination of those parameters from the perspective of time to first waste package failure. As shown in the S&ER Rev. 1, Section 4.2.4.4, the results of this case indicate the earliest possible failure time of a waste package for the upper bound profile is about 12,000 years (based on the TSPA-SR model), much earlier than the more realistic median profile (about 50,000 years). The time to fail 10 percent of waste packages for the upper bound profile is about 22,000 years. Uncertainty and variability are accounted for in the Waste Package Degradation code. Sensitivity analyses were also performed considering a number of failures prior to 10,000 years, as shown in Section 4.5.4, of the S&ER Rev. 1. A more recent analysis, contained in Section 3.2.8.2 of the SSE, assumed that a few (3 or less) waste packages could fail prior to 10,000 years due to assumed, undetected, improper heat treatment of the final closure weld. [BSC (Bechtel SAIC Company) 2001. "Total System Performance Assessment—Analyses for Disposal of Commercial and DOE Waste Inventories at Yucca Mountain—Input to Final Environmental Impact Statement and Site Suitability Evaluation." SL986M3 REV 00 ICN 01. Las Vegas, Nevada: Bechtel SAIC Company. ACC: MOL.20011114.0246. Section 5.2.4.2.]. The remaining waste packages would have lifetimes of about 80,000 years to over one million years. Issue Issues have been raised by members of the public regarding the difficulty in estimating long-term performance and how the DOE would provide confidence in its forecasts. Response The combination of long-term and short-term tests, the use of analogues, information gathered during in situ monitoring, plus the further development of a long-term mechanistic model for passive film stability provide confidence in the DOE's ability to forecast the long-term performance of waste package materials. It is recognized that the forecasting of performance of an engineered system for thousands of years is unprecedented. However, the DOE has chosen to follow the guidance provided in the ASTM C 1174-97 standard. [ASTM C 1174-97. 1998. "Standard Practice for Prediction of the Long-Term Behavior of Materials, Including Waste Forms, Used in Engineered Barrier Systems (EBS) for Geological Disposal of High-Level Radioactive Waste." West Conshohocken, Pennsylvania: American Society for Testing and Materials. TIC: 246015.]. This practice utilizes a combination of experiments, including short-term and long-term tests under both service and accelerated conditions, and models along with analogue information. Testing at conditions expected at Yucca Mountain is referred to as service-condition testing. Testing outside of the range is considered to be aggressive and is thus categorized as accelerated testing. It should be noted that for accelerated testing to be meaningful, the corrosion mechanism must not be changed from that which is active under the service conditions. The DOE would also take advantage of information gathered during in situ monitoring. Issue Issues have been raised by members of the public about the difference in waste package performance with the hot versus cold scenarios and whether the higher repository temperatures would be a concern for rockfall. Response Although a range of thermal operating modes was investigated, waste package performance as evaluated by the expected maximum dose to the public did not vary greatly with repository temperatures. The performance of repository system under a lower-temperature operating mode was discussed in detail in Section 2.1.5 of the S&ER Rev. 1. At lower temperatures, the overall amount of rockfall is likely to be lower, but the localized amount of rockfall could be greater due to nonuniform temperatures along the drift. In addition, if the ventilation period is extended, the potential for preclosure rockfall would increase. Another result of lower temperatures is that corrosion susceptibility is lowered and uncertainty is reduced. However, aqueous processes are initiated sooner in contrast to the higher operating mode where the initiation of corrosion is delayed. Each of the degradation mechanisms utilized to forecast the performance of the waste package includes temperature as a variable. Thus, the response of the waste package to a range of thermal conditions is built into the models. See the S&ER Rev. 1, Section 4.4.5.1.2, for further detail.
4.7.02 Waste Form Degradation and Radionuclide Release
4.7.02 (38) Summary CommentIssues have been raised by members of the public regarding the types of nuclear materials that would be stored in the repository, the initial condition of these waste forms upon receipt, the assumptions and methods for forecasting degradation of these waste forms over time, the need for additional testing of waste form degradation, and geophysical and chemical processes related to the solubility of radionuclides.Response The high-level radioactive waste forms the DOE would place in the repository consist primarily of commercial light water reactor spent nuclear fuel, high-level waste glasses containing the radioactive residues from nuclear weapons production, and DOE spent nuclear fuel. The great majority of waste received consists of assemblies that contain uranium oxide fuel pellets encased in intact zirconium alloy tubes. All waste forms will be in a solid form. Materials that could ignite or react chemically at a level that would compromise containment or isolation would be emplaced in canisters which are designed such that they could not be breached during preclosure repository operation. Neither the waste forms nor the waste packages would contain free liquids that could compromise waste containment. Materials that are regulated as hazardous waste under the RCRA would not be disposed in a repository. Waste form degradation affects the repository performance analysis after the waste package has been breached or otherwise failed. Water contact with the waste form is the primary cause of degradation after the waste package no longer provides for isolation of the waste form. Waste form degradation occurs in the TSPA only after the waste package container breaches, exposing the waste form to the air or water or both environments of the Engineered Barrier System. Breach of the waste package container is conservatively assumed to result in direct exposure of the DOE spent nuclear fuel and high-level waste glass forms to the repository air and water environment. In the cases of the high-level waste glass and DOE spent fuel, no radionuclide containment credit is taken in TSPA for their canisters. In this way, the DOE conservatively accounts for waste form degradation as a source of radionuclide release. The application of the waste form degradation models involves the extrapolation over periods of time that are orders of magnitude greater than the experimental test periods used to generate the degradation models. The DOE model development conformed to an American Society for Testing and Materials standard [ASTM C 1174-97. 1998. "Standard Practice for Prediction of the Long-Term Behavior of Materials, Including Waste Forms, Used in Engineered Barrier Systems (EBS) for Geologic Disposal of High-Level Radioactive Waste." American Society for Testing and Materials." West Conshohocken, Pa. TIC: 246015. Sections 19 and 20.] in developing waste form degradation models. Since TSPA analyses (S&ER Rev. 1, Section 4.2.6.3.6) have shown that the overall performance of the repository is very insensitive to the degradation rate of the DOE spent fuel, the emphasis, whenever possible, is on the application of upper-limit or bounding-degradation models for the spent fuel degradation. Confirmation testing of materials behavior models (e.g., the commercial spent fuel degradation model) could be required for the repository, where many years might pass from the time a decision is made to proceed until the repository is closed. The DOE expects that additional investigations would continue throughout the entire project period and the repository design, or decisions regarding the repository, may be changed to reflect newly identified information (S&ER Rev. 1, Sections 4.6.1 and 4.6.1.1.1). Confirmation testing of key materials behavior models would be performed as necessary for the repository. This could include testing that would be required to establish the corrosion characteristics of the high-level waste pour canister. The DOE expects these investigations could continue throughout the entire project period and that the repository design or decisions regarding the repository would be changed to reflect newly identified information (S&ER Rev. 1, Section 4.6.1). The DOE has considered the geochemistry of the system by evaluating the solubilities of key radionuclides. The solubilities of a number of radionuclide-bearing solids were measured as a function of water composition and temperature (S&ER Rev. 1, Section 4.2.6.2.7). Uranyl minerals would precipitate under the oxidation conditions expected when waste package breach exposes the waste forms to incoming water. Laboratory tests and field observations on natural analogue materials suggest the most common secondary uranyl phases to form under repository conditions would be schoepite, soddyite, uranophane, and sodium boltwoodite. Additionally, because carbonate levels tend to be higher at high pH and lower at low pH, the formation of soluble complexes of uranium and plutonium carbonates tend to increase at high pH (S&ER Rev. 1, Section 4.2.6.3.2). Neptunium solubilities are similar at pH 7 and pH 8.5 and are observed to decrease with increasing temperature; neptunium solubilities at pH 6 are one to two orders of magnitude higher than at pH 7 to 8.5. In general plutonium solubility is about three orders of magnitude lower, and is less affected by pH than that of neptunium. Increasing temperature decreases the solubility of plutonium. Under conservative assumptions of oxidizing repository conditions, both laboratory measurements and thermodynamic analysis indicate that no insoluble salts of technetium, chlorine, or cesium form. Each form is relatively large monovalent ions, which are exceedingly soluble. Therefore, the solubility of each is set in TSPA to 1.0 moles per liter, which lets their inventory in the waste form determine their release rate. Carbon and strontium both form less soluble metal carbonate minerals. Rather than perform a complex prediction of carbon and strontium solubility, their solubility was conservatively set at 1.0 moles per liter. In this way the DOE does account, albeit conservatively, for the solubilities of key radionuclides in the TSPA.
4.7.03 Unsaturated Zone Transport
4.7.03 (66) Summary CommentComments involved the effects of climatic changes, rapid water movement, fast pathways, fractured rock, radionuclide releases, and general analyses of fluid flow and radionuclide transport in the unsaturated zone surrounding a repository at Yucca Mountain.Response The DOE believes that any radionuclide releases reaching the accessible environment through the unsaturated and saturated zones during the 10,000 year postclosure period would likely be below the NRC's radiation protection standards for licensing. Section 1.4 of the S&ER Rev. 1 describes the geologic settings of Yucca Mountain and the surrounding region in detail. The DOE believes that the information in the S&ER Rev. 1 and the FEIS on the amount and type of contaminants that would be released over time from a repository and from other sources in the region have been adequately described and analyzed. Percolating water at the waste-emplacement horizon is unlikely to contact the waste packages because the excavation of the waste-emplacement drifts creates a capillary barrier that tends to divert water around the drift opening. This phenomenon limits the amount of water that can contact the waste packages. Section 4.2.1.4 of the S&ER Rev. 1 discusses the breakthrough time required for a nonsorbing tracer to reach the water table from the repository horizon. The breakthrough time of a nonsorbing tracer is not equivalent to travel time of water in the unsaturated zone. Breakthrough times are related to concentration limits that constitute a "significant arrival" while travel times in the unsaturated zone are related to the arrival of "significant volume" of water. Because both properties are difficult to measure, they must be calculated. The breakthrough time of a radionuclide is usually less than the travel time of water. The arrival of the first radionuclides represent a very low probability event. Breakthrough times can vary from 400 years to 600,000 years depending on the upper and lower infiltration rates. Section 4.2.8.3.3 of the S&ER Rev. 1 provides related discussion about both field measurements and numerical predictions of the movement of radionuclides from a repository horizon to the water table. Importantly, scientists consider travel time of water in the unsaturated zone to be probabilistic, with a range of values reflecting uncertainty associated with flow and transport, in all rock units regardless of type or location. In addition to travel time of water in the unsaturated zone as a measure of performance, chemical processes would tend to retard transport of radionuclides. Section 4.2.8.3.3 of the S&ER Rev. 1 provides related discussion about both field measurements and numerical predictions of the movement of radionuclides from the repository horizon to the water table. Chemical transport characteristics measured in boreholes suggest that the vitric and zeolitic rock units between the repository and the water table would effectively inhibit transport of sorbing radionuclides. The S&ER Rev. 1, Sections 4.1, 4.2, and 4.4 included an evaluation of climate change and its effects on long-term performance. These effects included increased infiltration, increased flux at depth, increased radioactive material transport at depth after waste package failure, and a shortened path to the water table because of changes in water table elevation. The evaluations also considered three climate scenarios: present day, a monsoon climate, and a glacial-transition climate during and beyond the 10,000-year postclosure period. Section 3.3.2.1 of the SSE included an evaluation of post 10,000-year climate change including glacial climates and its effects on unsaturated zone flow characteristics and long-term performance. Extreme precipitation events discussed in Section 4.2.1 of the S&ER Rev. 1 do not greatly influence the infiltration rates. This is because the subsurface tends to "damp" the extreme events (particularly in the Paintbrush nonwelded stratigraphic unit) to produce a nearly uniform infiltration rate with time at depth. Extreme precipitation events are more closely associated with surface runoff events. Locality-based infiltration rates were used (not whole-mountain averages) to derive infiltration rates for repository zones modeled in the performance analysis. The measured corrosion rate for the waste package outer barrier material, Alloy 22, is very small. The waste packages would eventually corrode and release radionuclides into the groundwater. However, dose estimates are below applicable NRC radiation protection standards for licensing in 10 CFR Part 63 (S&ER Rev. 1, Sections 4.4 and SSE, Sections 4.1 and 4.2). In summary, the DOE recognizes that some radionuclides would eventually enter the accessible environment outside a repository. Dose estimates are below applicable NRC radiation protection standards for licensing in 10 CFR Part 63 (SSE, Sections 3.1.2 and 4.2). 4.7.03 (2018) Summary Comment
An issue was raised by the public regarding heater testing. The second heater test has not been completed, and there is to date no serious evaluation of how the combined stresses of excavation and intense heating will affect the hydraulic properties of the surrounding rock. However, seepage into the drift is sensitive to the size of fracture openings, which is likely to be sensitive to excavation and heating stresses. Simply put, thermally stressed rock may show more seepage than unstressed rock. Given this uncertainty concerning a critical aspect of site performance, it seems premature to proceed with a site recommendation.Response The DOE believes that it sufficiently understands the relationship between thermally induced stresses and changes in permeability. The DOE has evaluated the effects of thermal loading as it pertains to fracture deformation impact to permeability. A recent study considered the effect of fracture deformation on permeability in the rock mass surrounding an emplacement drift in a repository [BSC (Bechtel SAIC Company) 2001. "Coupled Thermal-Hydrologic-Mechanical Effects on Permeability Analysis and Models Report." ANL-NBS-HS-000037 REV 00. Las Vegas, Nevada: Bechtel SAIC Company. ACC: MOL.20010822.0092.]. As discussed in Section 7.1 of this report, the analysis indicates that the overall permeability would be significantly reduced in vertical fractures, while the permeability of horizontal fractures would remain relatively unchanged. These findings are pertinent because many fractures are orientated vertically. Note that the second heater test (Drift Scale Test) was used to assist with the validation of the model used in the report cited above. 4.7.03 (2020) Summary Comment
An issue was raised by the public regarding the treatment of radionuclide diffusion within the rock matrix.Response Matrix diffusion can be a major element of radionuclide retardation in unwelded tuffs or in transitional welded/nonwelded tuffs. However, for geologic units at Yucca Mountain comprised of either densely welded tuffs or tuffs with significant secondary mineralization, the pore connections are unusually "tight" which severely restricts movement of water from the fractures into the matrix. This observation is based on the difficulty of extracting pore water for analysis and the lack of equilibration between matrix water and fracture water in the perched water zones. This "tight" condition has led to the modeling of those zones as having reduced matrix diffusion. Therefore, matrix diffusion in welded tuffs is believed to be sufficiently understood and properly implemented into the performance assessment of the repository (S&ER Rev. 1, Sections 4.2.1 and 4.2.8, and SSE, Section 3.3.7.1.3). 4.7.03 (10282) Summary Comment
An issue was raised by the public regarding the treatment of fracture porosity and fracture sealing.Response The DOE has analyzed the effects of fracture porosity and fracture sealing. The DOE has determined that the magnitude of the porosity changes is related to the initial fracture porosity and the presence of heterogeneities in fracture properties. Changes in porosity of less than 10 percent were obtained with an earlier model that did not account for as large a precipitation rate during final dryout. Using a heterogeneous fracture permeability distribution, changes in porosity of several percent can occur locally, however because there is already a range of at least 4 orders of magnitude in fracture permeability and 2-3 orders of magnitude variation in aperture, changes due to mineral precipitation are generally within this initial variability. The model results are, however, based on average properties and therefore local effects may be greater or less than the model simulations forecast (SSE, Section 3.3, and S&ER Rev. 1, Section 4.2.8). As per discussion in Section 3.3 of the SSE, chloride concentrations calculated for the high-temperature case, were checked as part of the review process. The reported values are those obtained by the model simulations. Note that the plotted chloride concentrations are those from distinct times which may not in all cases correspond to those concentrations used in the abstractions (i.e., the concentration during the dryout period is higher than that during the highest period on the graph of chloride concentration versus time). 4.7.03 (10895) Summary Comment
An issue was raised by the public regarding the influence of heat on the movement of water in the rock.Response An in-depth understanding of thermally driven coupled processes, including hydrological, mechanical, and chemical behavior, has been developed from the thermal testing program at Yucca Mountain. As discussed in Section 4.2.2.2.3 of the S&ER Rev. 1, the DOE has completed several in situ thermal tests, and the largest of these continues underground at Yucca Mountain. In addition, other comparable thermal tests have been conducted internationally to investigate geologic disposal of heat-producing waste. All of the thermal tests conducted at Yucca Mountain have included observation and measurement of thermal-hydrological-mechanical-chemical behavior. These measurements are compared to numerical simulations of the respective tests to validate thermal-hydrological, thermal-hydrological-mechanical, and thermal-hydrological-chemical models. Consequently, much work has been done at Yucca Mountain and elsewhere to better understand coupled processes. With regard to the inherent uncertainty associated with coupled processes including thermal, hydrological, mechanical, and chemical analyses, as they pertain to performance assessment, Sections 4.2.1, 4.2.2, 4.2.8, and 4.4.1 of the S&ER Rev. 1 describe how the DOE addressed these issues. The DOE recognizes that the acquisition of additional data could further reduce these uncertainties. Studies are planned to gather additional information. Section 4.6 of the S&ER Rev. 1 describes the types of tests, experiments, and analyses that the DOE would conduct during the performance confirmation and monitoring phases under the umbrella of an integrated test and evaluation program (S&ER Rev. 1, Section 5.4). The performance confirmation program could continue for as long as 300 years after waste emplacement operations have been completed. The purpose of the performance confirmation program is to evaluate the adequacy of the information used to demonstrate compliance with the licensing performance objectives in Subpart E of 10 CFR Part 63 through various tests, experiments, and analyses.
4.7.04 Saturated Zone Transport
4.7.04 (68) Summary CommentIssues have been raised by members of the public regarding fluid flow and radionuclide transport in the saturated zone underlying a repository at Yucca Mountain including general analyses of fluid flow, radionuclide transport, the potential existence of fast pathways, and the impact of seismic activity.Response The flow of groundwater from Yucca Mountain would be limited to the Death Valley flow system in southern California and would not intersect surrounding areas, including the Las Vegas Valley or other areas of southern California. Any releases of radionuclides during the 10,000-year postclosure period. The objective of the saturated zone flow and transport process model and the corresponding components of a repository's performance assessment (see S&ER Rev. 1, Section 4.4) is to evaluate the migration of radionuclides from their introduction at the water table below Yucca Mountain to the location at which releases and doses must be assessed. Consistent with NRC licensing regulations, the RMEI is assumed to live at a point above the highest concentration of radionuclides in the simulated plume of contamination where the plume crosses the southernmost boundary of the controlled area (at a latitude of 36 degrees 40 minutes 13.6661 seconds North). This distance is approximately 18 kilometers (11 miles) from within the repository footprint. The main output of the saturated zone flow and transport process models used directly by the TSPA is an assessment of the concentration of radionuclides in groundwater and the time it takes for various radionuclides to be transported from areas beneath a repository to the accessible environment. As described in Section 4.2.9 of the S&ER Rev. 1 and Section 3.3.8 of the SSE, the DOE has conducted an extensive program to characterize the direction and nature of groundwater movement and radionuclide transport from a repository at Yucca Mountain. This characterization has determined that the very small amount of water that percolates through Yucca Mountain to the water table would then travel southward. The groundwater beneath Yucca Mountain merges and mixes with groundwater beneath Fortymile Wash. This groundwater then flows toward, and mixes with, the large groundwater reservoir beneath the Amargosa Desert. The natural discharge point of this groundwater occurs farther south in Franklin Lake Playa, an area of extensive evapotranspiration. A minor volume may flow south toward Tecopa into the southern Death Valley area. A fraction of the groundwater may flow through fractures in the relatively impermeable Precambrian rocks at the southeastern end of the Funeral Mountains toward springs in the Furnace Creek area of Death Valley. Potentiometric data indicate that a divide could exist in the Funeral Mountains between the Amargosa Desert and Death Valley. This divide would limit discharge from the shallow flow system but would not necessarily affect the flow from the deeper carbonate aquifer that may contribute discharge to springs in the Furnace Creek area. Geochemical, isotopic, and temperature data indicate that water discharging from springs in the Furnace Creek area is a mixture of water from basin-fill aquifers in the northwestern Amargosa Desert and from deeper flow in the regional carbonate aquifer. Groundwater in the northwestern Amargosa Desert originates in Oasis Valley and from the eastern slope of the Funeral Mountains, both of which are west of the flow paths that extend southward from Yucca Mountain. Even if part of the flow from Yucca Mountain mixes with the carbonate pathway that supplies the springs in Furnace Creek, it would be too little to noticeably affect the water quality of these springs. Considering the small amount of water that would infiltrate through a repository compared to the total amount of water moving through the basin (approximately 0.2 percent or less), and the large distances involved (more than 60 kilometers [37 miles] from the source), any component of flow from Yucca Mountain that traveled along this long and complicated flow path would be significantly diluted. Section 1.4 of the S&ER Rev. 1 describes the geologic setting of Yucca Mountain and the surrounding region. Section 4.4 of the S&ER Rev. 1 and Sections 8.2 and 8.3 of the FEIS consider the cumulative impacts to groundwater from a repository, the Nevada Test Site, and other activities in the area that could contribute to long-term groundwater pollution. The amount and type of contaminants released over time from a repository and from other sources in the region have been adequately described and analyzed in the S&ER Rev. 1 and the FEIS. Estimated releases to the accessible environment after 10,000 years would be limited geographically to the groundwater movement system described in Section 4.2.9 of the S&ER Rev. 1; contaminants from a repository could not reach the Las Vegas Valley, the Colorado River, or any other parts of Nevada and California outside of the Death Valley groundwater system. Based on discussion in Section 4.3.2.2 of the S&ER Rev. 1 and Section 3.3.10.2 of the SSE, seismic/earthquake activity would not significantly alter the performance assessment of a repository. In addition, monitoring of contemporary seismicity indicates that earthquake activity continues in the Yucca Mountain vicinity. Consequently, seismic activity is treated in the TSPA as an expected event within the nominal (rather than disruptive) scenario, but with an uncertain magnitude and frequency. Because an earthquake could have consequences on the repository's underground facilities and hydrologic system, the DOE has performed an in-depth analysis to provide ground motion and fault displacement hazard results for both preclosure design determination and for postclosure performance assessment at both the surface and the repository level. During an earthquake, the underground portion of the repository, located at a depth of about 300 meters (1,000 feet), would experience significantly less vibratory ground motion than the surface facilities. Section 4.3.3.1 of the S&ER Rev. 1 discusses several views regarding fluctuations in the elevation of the water table. According to one assertion, the water table at Yucca Mountain has risen in the past to elevations that are higher than the waste emplacement horizon beneath Yucca Mountain. Based on the results of analyses reported in Section 4.3.3.1 of the S&ER Rev. 1, the DOE does not believe that any credible combination of future climate change, earthquakes, and volcanic eruptions could raise the water table sufficiently high enough to inundate the waste emplacement horizon. The National Research Council established a panel that reviewed the pertinent literature and data available up to 1992. This panel consulted with scientists involved in related field and laboratory studies. The panel concluded that none of the evidence offered as proof of groundwater upwelling in and around Yucca Mountain could reasonably be attributed to that process. In summary, the DOE recognizes that some radionuclides could eventually enter the environment outside a repository. However, TSPA dose estimates are below applicable NRC radiation protection standards for licensing in 10 CFR Part 63, (S&ER Rev. 1, Section 4.4, and SSE, Sections 4.1 and 4.2).
4.7.05 Biosphere Performance—Nominal Performance
4.7.05 (28) Summary CommentIssues were raised by members of the public regarding potential leakage of radioactive material from a repository at Yucca Mountain into the groundwater and the associated consequences. The issues are grouped in four general categories: the likelihood and need to prevent releases; the potential for and extent of effects on people and the environment, including farm products; potential radiation doses and health consequences, and whether the radiation protection standards will be met; and protection of groundwater as a resource.Issue Issues have been raised by members of the public regarding the possibility of a Yucca Mountain repository contaminating the groundwater with radioactive material. Various concerns have been expressed regarding the likelihood, and the need for prevention of release and transport of this material to the groundwater in the vicinity. Response The DOE is aware of the consequences of potential releases of radioactive material from a Yucca Mountain repository entering the environment through the groundwater pathway. The DOE's design and evaluation of the repository reflects an effort to minimize these potential consequences. Thus, the DOE's conceptual design incorporates a system of multiple engineered and natural barriers working together to keep water away from the waste and to protect public health and the environment for thousands of years. The DOE's goal for geologic disposal at Yucca Mountain would be to isolate radioactive wastes in a relatively small area for a very long time. The repository safety strategy is built upon four key attributes that would minimize the risk of radioactive contamination of groundwater: limited water entering emplacement drifts, long-lived waste package and drip shield, limited release of radioactive material from engineered barriers, and delayed and diluted radioactive material concentrations provided by natural barriers. This approach to design—employing multiple barriers that are developed using careful evaluation of the repository system's retention capabilities—is how the DOE is dealing with the potential for radioactive contamination of groundwater at a repository. Issue Issues have been raised by members of the public regarding the potential for people and the environment to be affected by the transport of groundwater contaminated with radioactive material in the regional aquifer. Concerns have been expressed about the extent of the area potentially affected, and consequences if local farm products are exported. Response The DOE includes groundwater contamination with radioactive material in its analyses of potential public health risks associated with a repository at Yucca Mountain. The largest potential risk to groundwater users is to the people of Amargosa Valley because groundwater in the saturated zone beneath Yucca Mountain flows in a generally southerly direction toward this community (FEIS, Section 3.1.4). As indicated in Chapter 5 of the FEIS, overall human health impacts to area residents would be small, and would not occur until the far future. As explained in the S&ER Rev. 1 and the SSE, the hypothetical person studied to calculate dose would live year-round above the highest concentration of radionuclides in the groundwater contamination plume approximately 18 kilometers (11 miles) from within the repository footprint. The RMEI would annually consume or use 3,000 acre-feet of groundwater taken from potentially contaminated sources. RMEI is defined in the NRC licensing regulations 10 CFR 63.312 (66 FR 55814). The DOE has conducted an extensive program to characterize the direction and nature of groundwater movement from the Yucca Mountain site. The results are described in the S&ER Rev. 1, Section 4.2.9. The general path of water that percolates through Yucca Mountain is southward toward Amargosa Valley, then beneath the area around Death Valley Junction in the southern Amargosa Desert. The groundwater beneath Yucca Mountain merges and mixes with groundwater beneath Fortymile Wash. This groundwater then flows toward, and mixes with, the large groundwater reservoir in the Amargosa Desert. The major natural discharge point of this groundwater is farther south in Franklin Lake Playa, an area of extensive evapotranspiration. Issue Issues have been raised by members of the public regarding whether, despite efforts to the contrary, radioactive material would be released from a repository at Yucca Mountain over time and contaminate groundwater in the vicinity. Issues also address the amount of potential radiation doses that could be incurred as a result of people using this groundwater, the health consequences of those doses, and whether health protection standards would be met. Response Although engineered and natural systems eventually degrade or change, the DOE has shown that a repository can be designed, constructed, operated, monitored, and eventually closed so that radiation protection standards established by the NRC for licensing would not likely be exceeded. These standards (10 CFR Part 63 [66 FR 55732]) prescribe radiation dose limits that a repository must meet during the 10,000-year postclosure period. Section 3.1.2 of the SSE discusses the DOE's recent (September 2001) estimate of annual doses from groundwater potentially contaminated with radioactive material released from the repository [BSC (Bechtel SAIC Company) 2001. "Total System Performance Assessment—Analyses for Disposal of Commercial and DOE Waste Inventories at Yucca Mountain—Input to the Final Environmental Impact Statement and Site Suitability Evaluation." REV 00 ICN 01. Las Vegas, Nevada: Bechtel SAIC Company. ACC: MOL.20011114.0246.]. These doses were calculated for the RMEI located approximately 18 kilometers (11 miles) from within a repository footprint. In summarizing the analyses results, the SSE states that the peak mean dose calculated over the 10,000-year postclosure period is 1.7 x 10E-5 millirem per year for the higher-temperature repository operating mode and 1.1 x 10E-5 millirem per year for the lower-temperature operating mode. These estimated doses are extremely low (SSE, Section 3.1.2). As described in the FEIS, the DOE has determined that there would be essentially no radiation-related health impacts to the population around Yucca Mountain caused by contaminated groundwater. The results of analyses presented in the SSE, Section 3.1.2, and the FEIS, Section 5.4, show that the annual radiation dose for the nominal scenario is extremely low for more than 10,000 years after closure of the repository. Issue Issues have been raised by members of the public regarding protection of groundwater from potential leakage of radioactive material from a repository at Yucca Mountain. Response Section 801 of the Energy Policy Act of 1992 requires the EPA (not DOE) to set standards for the protection of public health and safety from releases of radioactive materials stored or disposed of at Yucca Mountain. For licensing, the NRC groundwater protection standards follow the EPA's groundwater protection standards in 40 CFR Part 197 (66 FR 32074), which are compatible with relevant EPA drinking water standards for the entire United States. For licensing, the NRC groundwater protection standard is 4 millirem per year. This is low compared to the average radiation exposure from natural sources of radiation of 300 millirem per year. To address the NRC groundwater protection standards for licensing, the revised supplemental TSPA model evaluated groundwater concentrations of radionuclides released from the disposal system into the accessible environment consistent with 10 CFR 63.331 and 10 CFR 63.332 (66 FR 55814). Section 3.1.2.5 of the SSE discusses the results of these analyses. A description of the revised supplemental model can be found in "Total System Performance Assessment—Analyses for Disposal of Commercial and DOE Waste Inventories at Yucca Mountain—Input to Final Environmental Impact Statement and Site Suitability Evaluation" [Williams, N.H. 2001. "Contract No. DE-AC08-01RW12101—Total System Performance Assessment—Analyses for Disposal of Commercial and DOE Waste Inventories at Yucca Mountain—Input to Final Environmental Impact Statement and Site Suitability Evaluation REV 00 ICN 02." Letter from N.H. Williams (BSC) to J.R. Summerson (DOE/YMSCO), December 11, 2001, RWA:cs-1204010670, with enclosure. ACC: MOL.20011213.0056.]. The calculated peak mean activity concentration for alpha-emitting radionuclides for the 10,000-year postclosure period is 1.8 x 10E-8 picocuries per liter, from a repository at Yucca Mountain that operates at higher temperatures and 3.3 x 10E-8 picocuries per liter for lower-temperature operation. These gross alpha concentrations do not include 1.1 picocuries per liter from natural background. The NRC groundwater protection standard for licensing, in 10 CFR Part 63, limits gross alpha concentration to 15 picocuries per liter, which includes natural background. The calculated peak mean activity concentration for combined radium-226 and radium-228 activity for the 10,000-year postclosure period is less than 1 x 10E-10 picocuries per liter for both the higher- and the lower-temperature operating mode. These radium concentrations do not include 1.04 picocuries per liter from natural background. The NRC groundwater protection standards for licensing, in 10 CFR Part 63, limit total radium concentration to 5 picocuries per liter, which includes natural background. The revised supplemental model also calculated the mean annual dose from the combined beta- and photon-emitting radionuclides based on consuming 2 liters of groundwater per day. For beta- and photon-emitting radionuclides, the model forecasts 2.3 x 10E-5 millirem per year from a repository at Yucca Mountain operating at higher temperatures and 1.3 x 10E-5 millirem per year for operating at lower temperatures. Both calculations consider the whole body and any organ. The NRC groundwater protection standard for licensing, in 10 CFR Part 63, limits exposures from consuming two liters of water per day from the representative volume of groundwater to 4 millirem per year to the whole body or any organ. 4.7.05 (29) Summary Comment
This summary comment addresses issues raised by members of the public regarding the radiation dose and associated health impacts due to potential leakage of radioactive material from a repository at Yucca Mountain into the groundwater. The issues are grouped in these five general categories: effects of potential releases of radioactive material to the groundwater, including radiation doses and health consequences; potential effects on plants and animals; potential harm to Native Americans; population groups potentially more exposed than the RMEI, and changes over time that could produce worst case scenarios; and lastly, concerns that the EPA's groundwater protection standards might be relaxed for a Yucca Mountain repository.Issue Issues have been raised by members of the public regarding the effects of potential releases of radioactive material from a repository at Yucca Mountain, including the potential for contaminating groundwater in the vicinity, the radiation doses that could be incurred as people use this groundwater, and the health consequences of those doses. Response As described in the FEIS, the DOE has determined that there would be essentially no radiation-related health impacts to the population around Yucca Mountain caused by releases from the disposal system into groundwater. The results of analyses presented in the SSE, Section 3.1.2, and the FEIS, Section 5.4, forecast that for 10,000 years, the radionuclide concentrations in groundwater and radiological exposures from using and consuming this groundwater would fall well below the groundwater and individual protection standards issued by the EPA and the NRC. Section 801 of the Energy Policy Act of 1992 requires the EPA (not the DOE) to set standards for the protection of public health and safety from releases of radioactive materials stored or disposed of at Yucca Mountain. For licensing, the NRC's groundwater and individual protection licensing standards, in 10 CFR Part 63, implement the corresponding EPA standards in 40 CFR Part 197 (66 FR 32074). The groundwater protection standards are compatible with relevant EPA drinking water standards for the entire United States, and the individual protection standard is 15 millirem per year. The dose standards are low compared to the average radiation exposure from natural sources of radiation of 300 millirem per year. To address the NRC individual protection standard for licensing, the revised supplemental model calculated the radiological exposure that the RMEI would receive from using groundwater (including irrigation and consumption). The RMEI would be located approximately 18 kilometers (11 miles) from within the repository footprint, above the highest concentration of radionuclides in the predominant direction of groundwater flow. The model forecasts a peak mean dose, calculated over the 10,000-year postclosure period, for a nominal scenario and a probability-weighted peak mean dose for the disruptive scenario. The nominal scenario includes both seismic activity and the assumed premature failure of up to three waste packages. The disruptive scenario includes igneous activity. For the nominal scenario during the 10,000-year postclosure period, the revised supplemental model forecasts a peak mean dose of 1.7 x 10E-5 millirem per year for the higher-temperature operating mode and 1.1 x 10E-5 millirem per year for the lower-temperature operating mode. The doses are attributed to the assumed failure of a few (3 or less) waste packages, due to assumed, undetected, improper heat treatment of the final closure weld. Figure 3-4 of the SSE shows mean annual dose results from the TSPA-SR model, supplemental model, and revised supplemental model that were developed to forecast nominal performance. These doses would be approximately one-third lower using an annual water demand of 3,000 acre-feet as specified in the NRC licensing regulations. Additionally, Figures 3-7 and 3-8 of the SSE show that groundwater contamination does not occur before 1,000 to 2,000 years, assuming these early waste package failures. For the disruptive scenario during the 10,000-year postclosure period, the revised supplemental model forecasts a probability-weighted peak mean dose of 0.1 millirem per year for both the higher- and lower-temperature operating modes. The disruptive scenario includes igneous eruption and intrusive events. In an eruptive event, it is assumed that magma would destroy some waste packages and bring this waste to the surface. In an intrusive event, it is assumed that the magma destroys some waste packages, but increases the potential for the waste to contaminate groundwater rather than bringing the waste to the surface. The mean annual probability of an igneous event is approximately 1 in 60 million per year. Although highly improbable, this event was not excluded from TSPA because its probability is greater than the screening threshold of 1 in 100 million per year, consistent with the NRC's licensing regulations at 10 CFR 63.342. The DOE combined the forecasted doses for the nominal and disruptive scenarios consistent with the NRC's guidance [NRC (U.S. Nuclear Regulatory Commission) 2000. "Issue Resolution Status Report, Key Technical Issue: Total System Performance Assessment and Integration." Rev. 3. Washington, D.C.: U.S. Nuclear Regulatory Commission. TIC: 249045.]. This produced a forecasted probability-weighted peak mean annual dose of 0.1 millirem per year. The NRC individual protection standard for licensing, in 10 CFR Part 63, is 15 millirem per year to the RMEI. Regarding concern that the estimated doses might affect future generations, the FEIS presents the results of analyses pertinent to this matter in Section 5.4.2. Table 5-7 of the FEIS shows the added risk of latent cancer fatality if an individual received the dose calculated for the RMEI modeled during the maximum dose year in the 10,000-year postclosure period. The incremental lifetime risk from exposure, during the maximum year, would be 6 x 10E-10 (i.e., one chance in approximately 1.7 billion). For perspective, cancer from all other sources is fatal to about one in four persons. Issue Issues have been raised by members of the public regarding whether radioactively contaminated groundwater would contaminate desert soil and result in harm to plants and animals, including endangered species. Response The DOE has determined that it is unlikely that there would be any adverse radiation-related impacts to the environment around Yucca Mountain caused by contaminated groundwater. Section 5.4 of the FEIS indicates that forecasted long-term levels of radioactive material concentration in groundwater and the resulting dose levels at the forecasted discharge area in Amargosa Valley would be low. More specifically, the SSE, Section 3.1.2, Figure 3-3, shows a very small dose (1.7 x 10E-5 millirem per year for the higher-temperature repository operating mode, and 1.1 x 10E-5 millirem per year for the lower-temperature operating mode) to the RMEI during the 10,000-year postclosure period for the nominal groundwater scenario. The DOE concludes that the dose rates to plants and animals in Amargosa Valley would be unlikely to cause measurable detrimental effects in populations of any species. This result is based on dose response information developed by the International Atomic Energy Agency for terrestrial ecosystems including the most radiosensitive species [IAEA (International Atomic Energy Agency) 1992. "Effects of Ionizing Radiation on Plants and Animals at Levels Implied by Current Radiation Protection Standards." Technical Reports Series No. 332. Vienna, Austria: International Atomic Energy Agency. TIC: 243768. Page 53.]. Issue Issues have been raised by members of the public regarding concern that Native Americans would be harmed due to potential exposure to groundwater contaminated with radioactive material leaking from a repository at Yucca Mountain. Response The analyses of potential groundwater-related exposures, performed by modeling a RMEI in accordance with EPA's standards in 40 CFR Part 197 and NRC's licensing regulations in 10 CFR Part 63, provide conservative dose estimates. The EPA has specifically considered the Paiute and Shoshone Tribes' traditional and customary uses of the area around Yucca Mountain to determine whether the RMEI's location used in the EPA's standard (40 CFR Part 197) prescribes a higher exposure than these Native Americans are likely to receive. The EPA states "...we conclude, after considering their description of tribal uses of the area, that the rural-residential RMEI (the EPA's regulatory receptor) is fully protective of tribal resources" (66 FR 32090). The EPA discusses the basis for this conclusion. First, tribal use of natural springs involves water that would have lower contamination levels as a result of repository releases than would wells at the location specified in 40 CFR Part 197, which tap aquifers closer and more directly affected. Second, "...tribal use of wildlife and non-irrigated vegetation should not contribute significantly to total individual dose estimates. Gaseous releases from the repository are not a significant contributor to individual doses...through inhalation or rainfall, and should contribute less to contamination of wildlife and non-irrigated vegetation than the use [by the RMEI] of contaminated well water for raising crops and animals for food consumption" (66 FR 32090). The EPA provides another reason to conclude the receptor prescribed in 40 CFR Part 197 is more conservative than, and thus protective of, Native Americans. The EPA states that the RMEI "...is assumed to be a full-time resident continually exposed to radiation coming from the disposal system. It appears that the tribal uses are intermittent and involve resources which are less likely to be contaminated, resulting in lower doses...." (66 FR 32091). Issue Issues have been raised by members of the public indicating that there are better choices among the potentially exposed population (for example, subsistence farmers) to use as the model receptor because they could receive much higher radiation doses in worst case scenarios. Comment input recommends that the DOE employ these alternatives in order to perform more conservative assessments. Response The RMEI used in the DOE's assessments is consistent with the EPA's standards in 40 CFR Part 197 and NRC's licensing regulations in 10 CFR Part 63. Thus, the use of alternative model receptors would be inappropriate. The DOE has developed its biosphere model using the RMEI prescribed by the EPA and the NRC (40 CFR Part 197 and 10 CFR Part 63 [66 FR 55732], respectively). The RMEI lives in the accessible environment above the highest concentration of radionuclides in the plume of contamination; has a diet and living style representative of the people who now reside in the town of Amargosa Valley; uses well water with concentrations of radionuclides based on an annual water demand of 3,000 acre-feet; drinks two liters of water per day from a groundwater well at that location; and is an adult. The regulations detail methods to be followed in arriving at the diet and living styles for the RMEI. The RMEI is specified to behave in a way that is expected to incur high exposure from groundwater potentially contaminated with radioactivity released from the repository. Because of location and lifestyle, the RMEI would be expected to receive higher dose from a repository than other members of the public who might be exposed, for example, by a portion of their diet being food or dairy products produced in Amargosa Valley. Comment input suggested that there are people who might be more exposed than the RMEI, and the DOE should use their characteristics in the dose assessments. An alternative provided is subsistence farmers, which is discussed below. This discussion may also apply to other possible alternatives. The EPA states, "...we could not find nor did any other party demonstrate that there is the subsistence-farmer lifestyle at, or downgradient from, Yucca Mountain" (66 FR 32089). The EPA also discusses subsistence farming in its support document [EPA (U.S. Environmental Protection Agency) 2001. 40 CFR Part 197. "Technical Support Document: Characterization and Comparison of Alternative Dose Receptors for Individual Radiation Protection for a Repository at Yucca Mountain." Washington, D.C.: U.S. Environmental Protection Agency. TIC: 250266.]. It states, "Past attempts to achieve subsistence farming in the Yucca Mountain region, even with incentives and subsidies... have failed." In the absence of current residents with this lifestyle in the vicinity of Yucca Mountain, including subsistence farming, the EPA states, "Any future projection involves speculation" (66 FR 32091). Such projection is not pursued because the EPA is "...following NAS's [National Academy of Sciences] recommendation to use current technology and living patterns because speculation upon future society and lifestyle variations can be endless and not scientifically supportable..." [Ibid.]. The EPA sums up these issues by stating, "...we believe that the RMEI [reasonably maximally exposed individual] approach is sufficiently conservative and that it is fully protective of the general population (including women and children, the very young, the elderly, and the infirm)" (66 FR 32089). The NRC has adopted the RMEI approach in its licensing regulations, 10 CFR Part 63 (66 FR 55732). Issue Issues have been raised by members of the public regarding the methods and rationale used to select values for the parameters that describe the characteristics and behaviors of the model receptor. The commenter advised the DOE that these assigned values are significant to the correctness of exposure analyses and dose calculations. Response The RMEI is consistent with the EPA's and the NRC's prescriptions of diet and living style characteristics for assessments to assure the public is adequately protected. The RMEI that the DOE uses in the TSPA dose assessment model is prescribed by NRC licensing regulations. The EPA's standard (40 CFR Part 197) and the NRC's licensing regulation (10 CFR Part 63 [66 FR 55732]) prescribe the diet and living style for the RMEI. These attributes are based on the diet and living style of current residents of Amargosa Valley. These regulations also prescribe additional conservatism. The diet and living style attributes of the RMEI that are most important to radiation exposure from the contaminated groundwater pathway are consuming locally grown food irrigated with potentially contaminated groundwater and drinking that water. The values used in the model for food ingestion are based on data obtained in a 1997 survey of the Amargosa Valley area conducted by the DOE [DOE (U.S. Department of Energy) 1997. "The 1997 Biosphere Food Consumption Survey Summary Findings and Technical Documentation." Las Vegas, NV: U.S. Department of Energy. Office of Civilian Radioactive Waste Management. ACC: MOL.19981021.0301.]. This detailed information for the RMEI is used in concert with diet patterns obtained from currently available census information in model receptor development. Issue Issues have been raised by members of the public regarding concern that some worst case scenarios are not used in analyzing potential radiation doses due to the groundwater pathway. These commenters indicate that such things as climate change, extreme cultural swings and economic scenarios should be included in analyzing potential exposure from groundwater over the long time periods after closure of a repository at Yucca Mountain. Response In developing its assessments for the groundwater pathway, the DOE has complied with the EPA standards and NRC licensing regulations for incorporating changes over time. The DOE has considered features, events, and processes (FEPs) that could affect the repository's performance over the 10,000-year postclosure period. The FEPs include climate change, volcanic activity, and seismic events among others. The DOE has performed analyses of potential exposures related to groundwater by modeling the RMEI in accordance with the EPA's standards in 40 CFR Part 197 and NRC's licensing regulations in 10 CFR Part 63. In explaining its regulatory decisions, the EPA states, "Extremes of behavior are not used as the basis for protection...Instead, cautious, but reasonable, assumptions would be used to establish the individual(s) most highly exposed" [EPA (U.S. Environmental Protection Agency) 2001. 40 CFR Part 197. "Technical Support Document: Characterization and Comparison of Alternative Dose Receptors for Individual Radiation Protection for a Repository at Yucca Mountain." Washington, D.C.: U.S. Environmental Protection Agency. TIC: 250266.]. The EPA also states, "The objective is to project doses that are within reason rather than extreme, but well above the average for the exposed population. This approach will estimate a level of exposure that is protective of the vast majority of exposed persons but is still within a reasonable range and not highly speculative" [Ibid., page 11.]. Regarding societal changes over long time periods, the National Academy of Sciences [National Research Council 1995. "Technical Bases for Yucca Mountain Standards." Washington, D.C.: National Academy Press. TIC: 217588. Page 122.], the NRC (66 FR 55757), and the EPA (66 FR 32091) have determined that speculation about future states of society can be almost unlimited. These organizations conclude that assumptions should be used for regulatory compliance calculations, and these assumptions should reflect current living patterns. The EPA's standard (40 CFR 197.15) states, "The DOE