[Federal Register Volume 59, Number 199 (Monday, October 17, 1994)] [Unknown Section] [Page 0] From the Federal Register Online via the Government Publishing Office [www.gpo.gov] [FR Doc No: 94-25585] [[Page Unknown]] [Federal Register: October 17, 1994] VOL. 59, NO. 199 Monday, October 17, 1994 NUCLEAR REGULATORY COMMISSION 10 CFR Parts 50, 52 and 100 RIN 3150-AD93 Reactor Site Criteria Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants and Proposed Denial of Petition From Free Environment, Inc. et al. AGENCY: Nuclear Regulatory Commission. ACTION: Proposed rule and proposed denial of petition from Free Environment, Inc. et al. ----------------------------------------------------------------------- SUMMARY: The Nuclear Regulatory Commission (NRC) is proposing to amend its regulations to update the criteria used in decisions regarding power reactor siting, including geologic, seismic, and earthquake engineering considerations for future nuclear power plants. The proposed rule would allow NRC to benefit from experience gained in the application of the procedures and methods set forth in the current regulation and to incorporate the rapid advancements in the earth sciences and earthquake engineering. In addition, this proposed rule benefits from the public comments received on the first proposed revision of the regulations. This proposed rule primarily consists of two separate changes, namely, the source term and dose considerations, and the seismic and earthquake engineering considerations of reactor siting. The Commission is also proposing to deny the remaining issue in petition (PRM-50-20) filed by Free Environment, Inc. et al. DATES: Comment period expires February 14,1995. Comments received after this date will be considered if it is practical to do so, but the Commission is able to assure consideration only for comments received on or before this date. ADDRESSES: Mail written comments to: Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Attention: Docketing and Service Branch. Deliver comments to 11555 Rockville Pike, Rockville, Maryland, between 7:45 am and 4:15 pm, Federal workdays. Copies of the regulatory analysis, the environmental assessment and finding of no significant impact, and comments received may be examined at the NRC Public Document Room at 2120 L Street NW. (Lower Level), Washington, DC. FOR FURTHER INFORMATION CONTACT: Dr. Andrew J. Murphy, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 415-6010, concerning the seismic and earthquake engineering aspects and Mr. Leonard Soffer, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 415-6574, concerning other siting aspects. SUPPLEMENTARY INFORMATION: I. Background. II. Objectives. III. Genesis. IV. Alternatives. V. Major Changes. A. Reactor Siting Criteria (Nonseismic). B. Seismic and Earthquake Engineering Criteria. VI. Related Regulatory Guides and Standard Review Plan Section. VII. Future Regulatory Action. VIII. Referenced Documents. IX. Electronic Format. X. Questions. XI. Finding of No Significant Environmental Impact: Availability. XII. Paperwork Reduction Act Statement. XIII. Regulatory Analysis. XIV. Regulatory Flexibility Certification. XV. Backfit Analysis. I. Background The present regulation regarding reactor site criteria (10 CFR part 100) was promulgated April 12, 1962 (27 FR 3509). NRC staff guidance on exclusion area and low population zone sizes as well as population density was issued in Regulatory Guide 4.7, ``General Site Suitability Criteria for Nuclear Power Stations,'' published for comment in September 1974. Revision 1 to this guide was issued in November 1975. On June 1, 1976, the Public Interest Research Group (PIRG) filed a petition for rulemaking (PRM-100-2) requesting that the NRC incorporate minimum exclusion area and low population zone distances and population density limits into the regulations. On April 28, 1977, Free Environment, Inc. et al., filed a petition for rulemaking (PRM-50-20). The remaining issue of this petition requests that the central Iowa nuclear project and other reactors be sited at least 40 miles from major population centers. In August 1978, the Commission directed the NRC staff to develop a general policy statement on nuclear power reactor siting. The ``Report of the Siting Policy Task Force'' (NUREG- 0625) was issued in August 1979 and provided recommendations regarding siting of future nuclear power reactors. In the 1980 Authorization Act for the NRC, the Congress directed the NRC to decouple siting from design and to specify demographic criteria for siting. On July 29, 1980 (45 FR 50350), the NRC issued an Advance Notice of Proposed Rulemaking (ANPRM) regarding revision of the reactor site criteria, which discussed the recommendations of the Siting Policy Task Force and sought public comments. The proposed rulemaking was deferred by the Commission in December 1981 to await development of a Safety Goal and improved research on accident source terms. On August 4, 1986 (51 FR 23044), the NRC issued its Policy Statement on Safety Goals that stated quantitative health objectives with regard to both prompt and latent cancer fatality risks. On December 14, 1988 (53 FR 50232), the NRC denied PRM-100-2 on the basis that it would unnecessarily restrict NRC's regulatory siting policies and would not result in a substantial increase in the overall protection of the public health and safety. Because of possible renewed interest in power reactor siting, the NRC is proceeding with a rulemaking in this area. The Commission proposes to address the remaining issue in PRM-50-20 as part of this rulemaking action. Appendix A to 10 CFR part 100, ``Seismic and Geologic Siting Criteria for Nuclear Power Plants,'' was originally issued as a proposed regulation on November 25, 1971 (36 FR 22601), published as a final regulation on November 13, 1973 (38 FR 31279), and became effective on December 13, 1973. There have been two amendments to 10 CFR part 100, appendix A. The first amendment, issued November 27, 1973 (38 FR 32575), corrected the final regulation by adding the legend under the diagram. The second amendment resulted from a petition for rulemaking (PRM-100-1) requesting that an opinion be issued that would interpret and clarify Appendix A with respect to the determination of the Safe Shutdown Earthquake. A notice of filing of the petition was published on May 14, 1975 (40 FR 20983). The substance of the petitioner's proposal was accepted and published as an immediately effective final regulation on January 10, 1977 (42 FR 2052). The first proposed revision to these regulations was published for public comment on October 20, 1992 (57 FR 47802). The availability of the five draft regulatory guides and the standard review plan section that were developed to provide guidance on meeting the proposed regulations was published on November 25, 1992 (57 FR 55601). The comment period for the proposed regulations was extended two times. First, the NRC staff initiated an extension (58 FR 271) from February 17, 1993 to March 24, 1993, to be consistent with the comment period on the draft regulatory guides and standard review plan section. Second, in response to a request from the public, the comment period was extended to June 1, 1993 (58 FR 16377). The proposed regulations published on October 20, 1992 (57 FR 47802) and draft guidance documents cited in the availability notice published on November 25, 1992 (57 FR 55601) are withdrawn because of the substantive nature of the changes to be made in response to public comments and are replaced with the second proposed revision of the regulations presented in this document. II. Objectives The objectives of this proposed regulatory action are to-- 1. State basic site criteria for future sites that, based upon experience and importance to risk, have been shown as key to protecting public health and safety; 2. Provide a stable regulatory basis for seismic and geologic siting and applicable earthquake engineering design of future nuclear power plants that will update and clarify regulatory requirements and provide a flexible structure to permit consideration of new technical understandings; and 3. Relocate source term and dose requirements that apply primarily to plant design into 10 CFR part 50. III. Genesis The proposed regulatory action reflects changes that are intended to (1) benefit from the experience gained in applying the existing regulation and from research; (2) resolve interpretive questions; (3) provide needed regulatory flexibility to incorporate state-of-the-art improvements in the geosciences and earthquake engineering; and (4) simplify the language to a more ``plain English'' text. In addition, the proposed regulatory action will benefit from public comments received on the first proposed revision of the regulations and guidance documents. The proposed regulatory action would apply to applicants who apply for a construction permit, operating license, preliminary design approval, final design approval, manufacturing license, early site permit, design certification, or combined license on or after the effective date of the final regulations. Criteria not associated with the selection of the site or establishment of the Safe Shutdown Earthquake Ground Motion (SSE) have been placed into 10 CFR part 50. This action is consistent with the location of other design requirements in 10 CFR part 50. Because the revised criteria presented in the proposed regulation would not be applied to existing plants, the licensing bases for existing nuclear power plants must remain part of the regulations. Therefore, the non-seismic and seismic reactor site criteria for current plants would be retained as subpart A and appendix A to 10 CFR part 100, respectively. The proposed revised reactor site criteria would be added as subpart B in 10 CFR part 100 and would apply to site applications received on or after the effective date of the final regulations. Non-seismic site criteria would be added as a new Sec. 100.21 to subpart B in 10 CFR part 100. The criteria on seismic and geologic siting would be added as a new Sec. 100.23 to subpart B in 10 CFR part 100. The dose calculations and the earthquake engineering criteria would be located in 10 CFR part 50 (Sec. 50.34(a) and Appendix S, respectively). Because Appendix S is not self executing, applicable sections of part 50 (Sec. 50.34 and Sec. 50.54) are revised to reference appendix S. The proposed regulation would also make conforming amendments to 10 CFR part 52. Section 52.17(a)(1) would be amended to reflect changes in 50.34(a)(1) and 10 CFR Part 100. IV. Alternatives The first alternative considered by the Commission was to continue using current regulations for site suitability determinations. This is not considered an acceptable alternative. Accident source terms and dose calculations currently primarily influence plant design requirements rather than siting. It is desirable to state basic site criteria which, through importance to risk, have been shown to be key to assuring public health and safety. Further, significant advances in understanding severe accident behavior, including fission product release and transport, as well as in the earth sciences and in earthquake engineering have taken place since the promulgation of the present regulation and deserve to be reflected in the regulations. The second alternative considered was replacement of the existing regulation with an entirely new regulation. This is not an acceptable alternative because the provisions of the existing regulations form part of the licensing bases for many of the operating nuclear power plants and others that are in various stages of obtaining operating licenses. Therefore, these provisions should remain in force and effect. The approach of establishing the revised requirements in new sections to 10 CFR part 100 and relocating plant design requirements to 10 CFR part 50 while retaining the existing regulation was chosen as the best alternative. The public will benefit from a clearer, more uniform, and more consistent licensing process that incorporates updated information and is subject to fewer interpretations. The NRC staff will benefit from improved regulatory implementation (both technical and legal), fewer interpretive debates, and increased regulatory flexibility. Applicants will derive the same benefits in addition to avoiding licensing delays caused by unclear regulatory requirements. V. Major Changes A. Reactor Siting Criteria (Nonseismic) Since promulgation of the reactor site criteria in 1962, the Commission has approved more than 75 sites for nuclear power reactors and has had an opportunity to review a number of others. In addition, light-water commercial power reactors have accumulated about 1800 reactor-years of operating experience in the United States. As a result of these site reviews and operational experience, a great deal of insight has been gained regarding the design and operation of nuclear power plants as well as the site factors that influence risk. In addition, an extensive research effort has been conducted to understand accident phenomena, including fission product release and transport. This extensive operational experience together with the insights gained from recent severe accident research as well as numerous risk studies on radioactive material releases to the environment under severe accident conditions have all confirmed that present commercial power reactor design, construction, operation and siting is expected to effectively limit risk to the public to very low levels. These risk studies include the early ``Reactor Safety Study'' (WASH-1400), published in 1975, many Probabilistic Risk Assessment (PRA) studies conducted on individual plants as well as several specialized studies, and the recent ``Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants,'' (NUREG-1150), issued in 1990. Advanced reactor designs currently under review are expected to result in even lower risk and improved safety compared to existing plants. Hence, the substantial base of knowledge regarding power reactor siting, design, construction and operation reflects that the primary factors that determine public health and safety are the reactor design, construction and operation. Siting factors and criteria, however, are important in assuring that radiological doses from normal operation and postulated accidents will be acceptably low, that natural phenomena and potential man-made hazards will be appropriately accounted for in the design of the plant, and that site characteristics are amenable to the development of adequate emergency plans to protect the public and adequate security measures to protect the plant. The Commission has also had a long standing policy of siting reactors away from densely populated centers, and is continuing this policy in the proposed rule. The Commission is proposing to incorporate basic reactor site criteria in the proposed rule to accomplish the above purposes. The Commission proposes to retain source term and dose calculations to verify the adequacy of a site for a specific plant, but source term and dose calculations will be relocated to part 50, since experience has shown that these calculations have tended to influence plant design aspects such as containment leak rate or filter performance rather than siting. No specific source term would be referenced in part 50. Rather, the source term would be required to be one that is ``* * * assumed to result in substantial meltdown of the core with subsequent release into the containment of appreciable quantities of fission products.'' Hence, this guidance could be utilized with the source term currently used for light-water reactors, or used in conjunction with revised accident source terms, currently under development within the NRC staff as well as in the industry. The proposed relocation of source term and dose calculations to part 50 represent a partial decoupling of siting from accident source term and dose calculations. The siting criteria are envisioned to be utilized together with standardized plant designs whose features will be certified in a separate design certification rulemaking procedure. Each of the standardized designs would specify an atmospheric dilution factor that would be required to be met, in order to meet the dose criteria at the exclusion area boundary. For a given standardized design, a site having relatively poor dispersion characteristics would require a larger exclusion area distance than one having good dispersion characteristics. Additional design features would be discouraged in a standardized design to compensate for otherwise poor site conditions. Although individual plant tradeoffs would be discouraged for a given standardized design, a different standardized design could require a different atmospheric dilution factor. For custom plants that do not involve a standardized design, the source term and dose criteria will continue to provide assurance that the site is acceptable for the proposed design. Rationale for Individual Criteria A. Exclusion Area. An exclusion area surrounding the immediate vicinity of the plant has been a requirement for siting power reactors from the very beginning. This area provides a high degree of protection to the public from a variety of potential plant accidents and also affords protection to the plant from potential man-related hazards. The Commission considers an exclusion area to be an essential feature of a reactor site and is proposing to retain this requirement for future reactors. The proposed rule issued for comment in October 1992 proposed a minimum distance to the exclusion area boundary of 0.4 miles (640 meters), based upon the suggested value given in Regulatory Guide 4.7, without utilizing source term and dose calculations. This was based upon a conservative evaluation of the performance of fission product cleanup systems such as containment sprays or filter systems. Numerous comments were received stating that source term and dose calculations should be retained, and that the exclusion area distance should also be based upon a more realistic evaluation of actual fission product cleanup systems. In response to these comments, the Commission is proposing, in the present rule, to retain the use of source term and dose calculations, in part 50, to verify that an applicant's proposed exclusion area distance is adequate to assure that the radiological dose to an individual will be acceptably low in the event of a postulated accident. However, as noted above, if source term and dose calculations are used in conjunction with standardized designs, unlimited plant tradeoffs to compensate for poor site conditions would not be permitted. For plants that do not involve standardized designs, the source term and dose calculations would continue to provide assurance that the site is acceptable for the proposed design. The present regulation requires that the exclusion area be of such size that an individual located at any point on its boundary for two hours immediately following onset of the postulated fission product release would not receive a total radiation dose in excess of 25 rem to the whole body or 300 rem to the thyroid gland. A footnote in the present regulation notes that a whole body dose of 25 rem has been stated to correspond numerically to the once in a lifetime accidental or emergency dose to radiation workers which could be disregarded in the determination of their radiation exposure status (NBS Handbook 69 dated June 5, 1959). However, the same footnote also clearly states that the Commission's use of this value does not imply that it considers it to be an acceptable limit for an emergency dose to the public under accident conditions, but only that it represents a reference value to be used for evaluating plant features and site characteristics intended to mitigate the radiological consequences of accidents in order to provide assurance of low risk to the public under postulated accidents. The Commission, based upon extensive experience in applying this criterion, and in recognition of the conservatism of the assumptions in its application (a large fission product release within containment associated with major core damage, maximum allowable containment leak rate, a postulated single failure of any of the fission product cleanup systems, such as the containment sprays, adverse site meteorological dispersion characteristics, an individual presumed to be located at the boundary of the exclusion area at the centerline of the plume for two hours without protective actions), believes that this criterion has clearly resulted in an adequate level of protection. As an illustration of the conservatism of this assessment, the maximum whole body dose received by an actual individual during the Three Mile Island accident in March 1979, which involved major core damage, was estimated to be about 0.1 rem. In the proposed rule, the Commission is proposing two changes in this area. First, the Commission is proposing that the use of different doses for the whole body and thyroid gland be replaced by a single value of 25 rem, total effective dose equivalent (TEDE). The total effective dose equivalent concept is consistent with part 20 of the Commission's regulations, and is defined as the deep dose equivalent (for external exposures) plus the committed effective dose equivalent (for internal exposures). The deep dose equivalent is the same as the present whole body dose, while the committed effective dose equivalent is the sum of the products of doses to selected body organs times weighting factors for each organ that are representative of the radiation risk associated with that organ. The proposed use of the total effective dose equivalent, or TEDE, is based upon two considerations. First, since it utilizes a risk consistent methodology to assess the radiological impact of all relevant nuclides upon all body organs, use of TEDE promotes a uniformity and consistency in assessing radiation risk that may not exist with the separate whole body and thyroid organ dose values in the present regulation. Second, use of TEDE lends itself readily to the application of updated accident source terms, which can vary not only with plant design, but in which additional nuclides besides the noble gases and iodine are predicted to be released into containment. The Commission has examined the current dose criteria of 25 rem whole body and 300 rem thyroid with the intent of selecting a TEDE numerical value equivalent to the risk implied by the current dose criteria. These risks consist of the risk of developing cancer some time after the exposure (latent cancer incidence), as well as a delayed risk of cancer fatality (latent cancer fatality). For a dose of 25 rem whole body, the individual risk of latent cancer fatality is estimated to be about 2.5 x 10-2; the risk of latent cancer incidence is about twice that (using risk coefficients expressed by ICRP Publication 60 and in NUREG/CR-4214). For a dose of 300 rem thyroid, the risk of latent cancer fatality is about 2 x 10-3; the risk of latent cancer incidence is about a factor of ten higher. If the risk of latent cancer fatality is selected as the appropriate risk measure to be used, the current dose criteria represent a risk of about 2.7 x 10-2. Using a risk coefficient of about 10-3 per rem, the risk of latent cancer fatality implied by the current dose criteria is equivalent to 27 rem TEDE. (BEIR V estimates a latent cancer fatality risk coefficient of about 5 x 10-4 per rem, if the dose is received over a period of days or more; however, if the exposure period is shorter, such as 2 hours, the risk coefficient is approximately double.) If latent cancer incidence rather than fatality were used, the current dose criteria would correspond to a value of about 35 rem TEDE. The Commission is proposing to use the risk of latent cancer fatality as the appropriate risk measure since quantitative health objectives (QHOs) for it have been established in the Commission's Safety Goal policy. Although the current dose criteria are equivalent in risk to 27 rem TEDE, as noted above, the Commission is proposing to use 25 rem TEDE as the dose criterion for plant evaluation purposes, since this value is essentially the same level of risk as the current criteria. Nevertheless, the Commission is specifically requesting comments on the use of TEDE. Comments are requested on whether the current dose criteria should be modified to utilize the total effective dose equivalent, or TEDE, concept. The Commission is also requesting comments on whether a TEDE value of 25 rem (consistent with latent cancer fatality), or 34 rem (consistent with latent cancer incidence), or some other value should be used. Finally, because the thyroid weighting factor is equal to a value of 0.03, there exists a theoretical possibility that an accidental release composed only of iodine could result in a TEDE less than 25 rem, yet result in a thyroid dose of over 800 rem. Although the Commission believes that the likelihood that an actual accident would release only iodine is highly unlikely, comments are also requested as to whether the dose criterion should also include a ``capping'' limitation, that is, an additional requirement that the dose to any individual organ not be in excess of some fraction of the total. The second change being proposed in this area is in regard to the time period that a hypothetical individual is assumed to be at the exclusion area boundary. While the duration of the time period remains at a value of two hours, the Commission is proposing that this time period not be fixed in regard to the appearance of fission products within containment, but that various two-hour periods be examined with the objective that the dose to an individual not be in excess of 25 rem TEDE for any two-hour period after the appearance of fission products within containment. The Commission is proposing this change to reflect improved understanding of fission product release into the containment under severe accident conditions. For an assumed instantaneous release of fission products, as contemplated by the present rule, the two hour period that commences with the onset of the fission product release clearly results in the highest dose to a hypothetical individual offsite. Improved understanding of severe accidents shows that fission product releases to the containment do not occur instantaneously, and that the bulk of the releases may not take place for about an hour or more. Hence, the two-hour period commencing with the onset of fission product release may not represent the highest dose that an individual could be exposed to over any two-hour period. As a result, the Commission is proposing that various two-hour periods be examined to assure that the dose to a hypothetical individual at the exclusion area boundary will not be in excess of 25 rem TEDE over any two-hour period after the onset of fission product release. B. Site Dispersion Factors. Site dispersion factors have been utilized to provide an assessment of dose to an individual as a result of a postulated accident. Since the Commission intends to require that a verification be made that the exclusion area distance is adequate to assure that the guideline dose to a hypothetical individual will not be exceeded under postulated accident conditions, as well as to assure that radiological limits are met under normal operating conditions, the Commission is proposing that the atmospheric dispersion characteristics of the site will be required to be evaluated, and that site dispersion factors based upon this evaluation be determined and used in assessing radiological consequences of normal operations as well as accidents. C. Low Population Zone. The present regulation requires that a low population zone (LPZ) be defined immediately beyond the exclusion area. Residents are permitted in this area, but the number and density must be such that there is a reasonable probability that appropriate protective measures could be taken in their behalf in the event of a serious accident. In addition, the nearest densely populated center containing more than about 25,000 residents must be located no closer than one and one-third times the outer boundary of the LPZ. Finally, the dose to a hypothetical individual located at the outer boundary of the LPZ over the entire course of the accident must not be in excess of the dose values given in the regulation. Before 1980, the LPZ generally defined the distance over which public protective actions were contemplated in the event of a serious accident. The regulations in 10 CFR 50.47 now requires plume exposure Emergency Planning Zones (EPZ) of about 10 miles for each plant. While the Commission considers that the siting functions intended for the LPZ, namely, a low density of residents and the feasibility of taking protective actions, have been accomplished by other regulations or can be accomplished by other guidance, the Commission continues to believe that a requirement that limits the radiological consequences over the course of the accident provides a useful evaluation of the plant's long-term capability to mitigate postulated accidents. For this reason, the Commission is proposing to retain the requirement that the dose consequences be evaluated at the outer boundary of the LPZ over the course of the postulated accident and that these not be in excess of 25 rem TEDE. D. Physical Characteristics of the Site. It has been required that physical characteristics of the site, such as the geology, seismology, hydrology, meteorology characteristics be considered in the design and construction of any plant proposed to be located there. The proposed rule would require that these characteristics be evaluated and that site parameters, such as design basis flood conditions or tornado wind loadings be established for use in evaluating any plant to be located on that site in order to ensure that the occurrence of such physical phenomena would pose no undue hazard. E. Nearby Transportation Routes, Industrial and Military Facilities. As for natural phenomena, it has been a long-standing NRC staff practice to review man-related activities in the site vicinity to provide assurance that potential hazards associated with such facilities or transportation routes will pose no undue risk to any plant proposed to be located at the site. The proposed rule would codify this practice. F. Adequacy of Security Plans. The proposed rule would require that the characteristics of the site be such that adequate security plans and measures for the plant could be developed. The Commission envisions that this would entail a small secure area considerably smaller than that envisioned for the exclusion area. G. Adequacy of Emergency Plans. The proposed rule would also require that the site characteristics be such that adequate plans to carry out protective measures for members of the public in the event of emergency could be developed. H. Siting Away From Densely Populated Centers. Population density considerations beyond the exclusion area have been required since issuance of part 100 in 1962. The current rule requires a ``low population zone'' (LPZ) beyond the immediate exclusion area. The LPZ boundary must be of such a size that an individual located at its outer boundary must not receive a dose in excess of the values given in part 100 over the course of the accident. While numerical values of population or population density are not specified for this region, the regulation also requires that the nearest boundary of a densely populated center of about 25,000 or more persons be located no closer than one and one-third times the LPZ outer boundary. Part 100 has no population criteria other than the size of the LPZ and the proximity of the nearest population center, but notes that ``where very large cities are involved, a greater distance may be necessary.'' Whereas the exclusion area size is based upon limitation of individual risk, population density requirements serve to set societal risk limitations and reflect consideration of accidents beyond the design basis, or severe accidents. Such accidents were clearly a consideration in the original issuance of part 100, since the Statement of Considerations (27 FR 3509; April 12, 1962) noted that: Further, since accidents of greater potential hazard than those commonly postulated as representing an upper limit are conceivable, although highly improbable, it was considered desirable to provide for protection against excessive exposure doses to people in large centers, where effective protective measures might not be feasible * * *. Hence, the population center distance was added as a site requirement. Limitation of population density beyond the exclusion area has the following benefits: (a) It facilitates emergency preparedness and planning; and (b) It reduces potential doses to large numbers of people and reduces property damage in the event of severe accidents. Although the Commission's Safety Goal policy provides guidance on individual risk limitations, in the form of the Quantitative Health Objectives (QHO), it provides no guidance with regard to societal risk limitations and therefore cannot be used to ascertain whether a particular population density would meet the Safety Goal. However, results of severe accident risk studies, particularly those obtained from NUREG-1150, can provide useful insights for considering potential criteria for population density. Severe accidents having the highest consequences are those where core-melt together with early bypass of or containment failure occurs. Such an event would likely lead to a ``large release'' (without defining this precisely). Based upon NUREG-1150, the probability of a core-melt accident together with early containment failure or bypass for some current generation LWRs is estimated to be between 10-5 and 10-6 per reactor year. For future plants, this value is expected to be less than 10-6 per reactor year. If a reactor was located nearer to a large city than current NRC practice permitted, the likelihood of exposing a large number of people to significant releases of radioactive material would be about the same as the probability of a core-melt and early containment failure, that is, less than 10-6 per reactor year for future reactor designs. It is worth noting that events having the very low likelihood of about 10-6 per reactor year or lower have been regarded in past licensing actions to be ``incredible'', and as such, have not been required to be incorporated into the design basis of the plant. Hence, based solely upon accident likelihood, it might be argued that siting a reactor nearer to a large city than current NRC practice would pose no undue risk. If, however, a reactor were sited away from large cities, the likelihood of the city being affected would be reduced because of two factors. First, because the wind is expected to blow in all directions with roughly the same frequency, the likelihood that radioactive material would actually be carried towards the city is reduced significantly because it is likely that the wind will blow in a direction away from the city. Second, the radiological dose consequences would also be reduced with distance because the radioactive material becomes increasingly diluted by the atmosphere and the inventory becomes depleted due to the natural processes of fallout and rainout before reaching the city. Analyses indicate that if a reactor were located at distances ranging from 10 to about 20 miles away from a city, depending upon its size, the likelihood of exposure of large numbers of people within the city would be reduced by factors of ten to one hundred or more compared with locating a reactor very close to a city. In summary, next-generation reactors are expected to have risk characteristics sufficiently low that the safety of the public is reasonably assured by the reactor and plant design and operation itself, resulting in a very low likelihood of occurrence of a severe accident. Such a plant can satisfy the QHOs of the Safety Goal with a very small exclusion area distance (as low as 0.1 miles). The consequences of design basis accidents, analyzed using revised source terms and with a realistic evaluation of engineered safety features, are likely to be found acceptable at distances of 0.25 miles or less. With regard to population density beyond the exclusion area, siting a reactor closer to a densely populated city than is current NRC practice would pose a very low risk to the populace. Nevertheless, the Commission considers that defense-in-depth considerations and the additional enhancement in safety to be gained by siting reactors away from densely populated centers should be maintained. The Commission is proposing a two-tier approach with regard to population density and reactor sites. The proposed rule states that reactor sites should be located away from very densely populated centers, and that areas of low population density are, generally, preferred. The Commission believes that a site not falling within these two categories, although not preferred, could be found acceptable under certain conditions. The Commission is not establishing specific numerical criteria for evaluation of population density in siting future reactor facilities because the acceptability of a specific site from the standpoint of population density must be considered in the overall context of safety and environmental considerations. The Commission's intent is to assure that a site that has significant safety, environmental or economic advantages is not rejected solely because it has a higher population density than other available sites. Population density is but one factor that must be balanced against the other advantages and disadvantages of a particular site in determining the site's acceptability. Thus, it must be recognized that sites with higher population density, so long as they are located away from very densely populated centers, can be approved by the Commission if they present advantages in terms of other considerations applicable to the evaluation of proposed sites. On April 28, 1977, Free Environment, Inc. et al., filed a petition for rulemaking (PRM-50-20) requesting, among other things, that ``the central Iowa nuclear project and other reactors be sited at least 40 miles from major population centers.'' The petitioner also stated that ``locating reactors in sparsely-populated areas * * * has been endorsed in non-binding NRC guidelines for reactor siting.'' The petitioner did not specify what constituted a major population center. The only NRC guidelines concerning population density in regard to reactor siting are in Regulatory Guide 4.7, issued in 1974, and revised in 1975, prior to the date of the petition. This guide states population density values of 500 persons per square mile out to a distance of 30 miles from the reactor, not 40 miles. Regulatory Guide 4.7 does provide effective separation from population centers of various sizes. Under this guide, a population center of about 25,000 or more residents should be no closer than 4 miles (6.4 km) from a reactor because a density of 500 persons per square mile within this distance would yield a total population of about 25,000 persons. Similarly, a city of 100,000 or more residents should be no closer than about 10 miles (16 km); a city of 500,000 or more persons should be no closer than about 20 miles (32 km), and a city of 1,000,000 or more persons should be no closer than about 30 miles (50 km) from the reactor. The Commission has examined these guidelines with regard to the Safety Goal. The Safety Goal quantitative health objective in regard to latent cancer fatality states that, within a distance of ten miles (16 km) from the reactor, the risk to the population of latent cancer fatality from nuclear power plant operation, including accidents, should not exceed one''tenth of one percent of the likelihood of latent cancer fatalities from all other causes. In addition to the risks of latent cancer fatalities, the Commission has also investigated the likelihood and extent of land contamination arising from the release of long lived radioactive species, such as cesium-137, in the event of a severe reactor accident. The results of these analyses indicate that the latent cancer fatality quantitative health objective noted above is met for current plant designs. From analysis done in support of this proposed change in regulation, the likelihood of permanent relocation of people located more than about 20 miles (50 km) from the reactor as a result of land contamination from a severe accident is very low. Therefore, the Commission concludes that the current NRC staff guidance in Regulatory Guide 4.7 provide a means of locating reactors away from population centers, including ``major'' population centers, depending upon their size, that would limit societal consequences significantly, in the event of a severe accident. The Commission finds that granting of the petitioner's request to specify population criteria out to 40 miles would not substantially reduce the risks to the public. As noted, the Commission also believes that a higher population density site could be found to be acceptable, compared to a lower population density site, provided there were safety, environmental or economic advantages to the higher population site. Granting of the petitioner's request would neglect this possibility and would make population density the sole criterion of site acceptability. For these reasons, the Commission has decided not to adopt the proposal by Free Environment, Incorporated. The Commission also notes that future population growth around a nuclear power plant site, as in other areas of the region, is expected but cannot be predicted with great accuracy, particularly in the long- term. Since higher population density sites are not unacceptable, per se, the Commission does not intend to consider license conditions or restrictions upon an operating reactor solely upon the basis that the population density around it may reach or exceed levels that were not expected at the time of site approval. Finally, the Commission wishes to emphasize that population considerations as well as other siting requirements apply only for the initial siting for new plants and will not be used in evaluating applications for the renewal of existing nuclear power plant licenses. Change to 10 CFR Part 50 The proposed change to 10 CFR part 50 would relocate from 10 CFR Part 100 the dose requirements for each applicant at specified distances. Because these requirements affect reactor design rather than siting, they are more appropriately located in 10 CFR part 50. These requirements would apply to future applicants for a construction permit, design certification, or an operating license. The Commission will consider after further experience in the review of certified designs whether more specific requirements need to be developed regarding revised accident source terms and severe accident insights. B. Seismic and Earthquake Engineering Criteria The following major changes in the proposed revision to Appendix A, ``Seismic and Geologic Siting Criteria for Nuclear Power Plants,'' to part 100, are associated with the proposed seismic and earthquake engineering criteria rule making. These changes reflect new information and research results, and incorporate the intentions of this regulatory action as defined in Section III of this proposed rule including comments from the public on the first proposed revision of the regulations. A specific document explaining the NRC staff's disposition of pertinent comments will be prepared coincident with the final rulemaking. 1. Separate Siting From Design Criteria not associated with site suitability or establishment of the Safe Shutdown Earthquake Ground Motion (SSE) have been placed into 10 CFR part 50. This action is consistent with the location of other design requirements in 10 CFR part 50. Because the revised criteria presented in the proposed regulation will not be applied to existing plants, the licensing basis for existing nuclear power plants must remain part of the regulations. The criteria on seismic and geologic siting would be designated as a new Sec. 100.23 to subpart B in 10 CFR part 100. Criteria on earthquake engineering would be designated as a new Appendix S, ``Earthquake Engineering Criteria for Nuclear Power Plants,'' to 10 CFR part 50. 2. Remove Detailed Guidance From the Regulation The current regulation contains both requirements and guidance on how to satisfy the requirements. For example, Section IV, ``Required Investigations,'' of Appendix A, states that investigations are required for vibratory ground motion, surface faulting, and seismically induced floods and water waves. Appendix A then provides detailed guidance on what constitutes an acceptable investigation. A similar situation exists in Section V, ``Seismic and Geologic Design Bases,'' of Appendix A. Geoscience assessments require considerable latitude in judgment. This latitude in judgment is needed because of limitations in data and the state-of-the-art of geologic and seismic analyses and because of the rapid evolution taking place in the geosciences in terms of accumulating knowledge and in modifying concepts. This need appears to have been recognized when the existing regulation was developed. The existing regulation states that it is based on limited geophysical and geological information and will be revised as necessary when more complete information becomes available. However, having geoscience assessments detailed and cast in a regulation has created difficulty for applicants and the staff in terms of inhibiting the use of needed latitude in judgment. Also, it has inhibited flexibility in applying basic principles to new situations and the use of evolving methods of analyses (for instance, probabilistic) in the licensing process. The proposed regulation would be streamlined, becoming a new section in Subpart B to 10 CFR part 100 rather than a new appendix to part 100. Also, the level of detail presented in the proposed regulation would be reduced considerably. This approach reflects the philosophy of the first proposed revision that the regulation only contains the basic requirements and that the detailed guidance, which is contained in the current regulation, Appendix A to 10 CFR part 100, be removed to guidance documents. Thus, the proposed regulation contains: (a) Required definitions, (b) A requirement to determine the geological, seismological, and engineering characteristics of the proposed site, and (c) A requirement to determine the Safe Shutdown Earthquake Ground Motion (SSE) and its uncertainty, to determine the potential for surface deformation, and to determine the design bases for seismically induced floods and water waves. The guidance documents describe how to carry out these required determinations. The key elements of the balanced approach to determine the SSE are presented in the following section. The elements are the guidance that will be fully described in the guidance documents. The proposed regulation is a new section in part 100 rather than an appendix to Part 100. The proposed regulation would identify and establish basic requirements. Detailed guidance, that is, the procedures acceptable to the NRC for meeting the requirements, would be contained in a draft regulatory guide to be issued for public comment as Draft Regulatory Guide, DG-1032, ``Identification and Characterization of Seismic Sources and Determination of Safe Shutdown Earthquake Ground Motions.'' 3. Uncertainties and Probabilistic Methods The existing approach for determining a Safe Shutdown Earthquake Ground Motion (SSE) for a nuclear reactor site, embodied in appendix A to 10 CFR part 100, relies on a ``deterministic'' approach. Using this deterministic approach, an applicant develops a single set of earthquake sources, develops for each source a postulated earthquake to be used as the source of ground motion that can affect the site, locates the postulated earthquake according to prescribed rules, and then calculates ground motions at the site. Although this approach has worked reasonably well for the past two decades, in the sense that SSEs for plants sited with this approach are judged to be suitably conservative, the approach has not explicitly recognized uncertainties in geosciences parameters. Because so little is known about earthquake phenomena (especially in the eastern United States), there have often been differences of opinion and differing interpretations among experts as to the largest earthquakes to be considered and ground-motion models to be used, thus often making the licensing process relatively unstable. Over the past decade, analysis methods for incorporating these different interpretations have been developed and used. These ``probabilistic'' methods have been designed to allow explicit incorporation of different models for zonation, earthquake size, ground motion, and other parameters. The advantage of using these probabilistic methods is their ability to not only incorporate different models and different data sets, but also to weight them using judgments as to the validity of the different models and data sets, and thereby providing an explicit expression for the uncertainty in the ground motion estimates and a means of assessing sensitivity to various input parameters. Another advantage of the probabilistic method is the target exceedance probability is set by examining the design bases of more recently licensed nuclear power plants. The proposed revision to the regulation now explicitly recognizes that there are inherent uncertainties in establishing the seismic and geologic design parameters and allows for the option of using a probabilistic seismic hazard methodology capable of propagating uncertainties as a means to address these uncertainties. The rule further recognizes that the nature of uncertainty and the appropriate approach to account for it depend greatly on the tectonic regime and parameters, such as, the knowledge of seismic sources, the existence of historical and recorded data, and the understanding of tectonics. Therefore, methods other than the probabilistic methods, such as sensitivity analyses, may be adequate for some sites to account for uncertainties. The NRC staff has achieved an appropriate balance between deterministic and probabilistic seismic hazard evaluations to be used in the revision of the seismic and geologic siting criteria for nuclear power plants. The key elements of this balanced approach are: --Conduct site-specific and regional geoscience investigations, --Target exceedance probability is set by examining the design bases of more recently licensed nuclear power plants, --Conduct probabilistic seismic hazard analysis and determine ground motion level corresponding to the target exceedance probability, --Determine if information from geoscience investigations change probabilistic results, --Determine site-specific spectral shape and scale this shape to the ground motion level determined above, --NRC staff review using all available data including insights and information from previous licensing experience, and --Update the data base and reassess probabilistic methods at least every ten years. Thus, the proposed approach requires thorough regional and site- specific geoscience investigations. The proposed approach reflects some of the comments of the U.S. utility industry. The U.S. Geological Survey provided a series of comments and recommendations that led to and can be met by the above integrated approach. Results of the regional and site-specific investigations must be considered in application of the probabilistic method. The current probabilistic methods, the NRC sponsored study conducted by Lawrence Livermore National Laboratory (LLNL) or the Electric Power Research Institute (EPRI) seismic hazard study, are essentially regional studies without detailed information on any specific location. The regional and site-specific investigations provide detailed information to update the database of the hazard methodology to make the probabilistic analysis site-specific. It is also necessary to incorporate local site geological factors such as stratigraphy and topography and to account for site-specific geotechnical properties in establishing the design basis ground motion. In order to incorporate local site factors and advances in ground motion attenuation models, ground motion estimates are determined using the procedures outlined in the Draft Standard Review Plan Section 2.5.2, Second Proposed Revision 3, ``Vibratory Ground Motion.'' Methods acceptable to the NRC staff for implementing the proposed regulation are described in Draft Regulatory Guide DG-1032, ``Identification and Characterization of Seismic Sources and Determination of Safe Shutdown Earthquake Ground Motions.'' The NRC staff's review approach to evaluate an application is described in Draft SRP Section 2.5.2. This review takes into account the information base developed in licensing more than 100 plants. This staff review is consistent with the intent of a USGS recommendation. Although the basic premise in establishing the target exceedance probability is that the current design levels are adequate, a staff review further assures that there is consistency with previous licensing decisions and that the scientific basis for decisions are clearly understood. This review approach will also assist in assessing the fairly complex regional probabilistic modeling which incorporates multiple hypotheses and a multitude of parameters. Furthermore, this process should provide a clear basis for the staff's decisions and facilitate communication with nonexperts. 4. Safe Shutdown Earthquake The existing regulation (10 CFR part 100, appendix A, section V(a)(1)(iv)) states ``The maximum vibratory accelerations of the Safe Shutdown Earthquake at each of the various foundation locations of the nuclear power plant structures at a given site shall be determined * * *''. The location of the seismic input motion control point as stated in the existing regulation has led to confrontations with many applicants that believe this stipulation is inconsistent with good engineering fundamentals. The proposed regulation would move the location of the seismic input motion control point from the foundation-level to the free-field at the free ground surface. The 1975 version of the Standard Review Plan placed the control motion in the free-field. The proposed regulation is also consistent with the resolution of Unresolved Safety Issue (USI) A-40, ``Seismic Design Criteria'' (August 1989), that resulted in the revision of Standard Review Plan Sections 2.5.2, 3.7.1, 3.7.2, and 3.7.3. However, the proposed regulation requires that the horizontal component of the Safe Shutdown Earthquake Ground Motion in the free-field at the foundation level of the structures must be an appropriate response spectrum considering the site geotechnical properties, with a peak ground acceleration of at least 0.1g. 5. Value of the Operating Basis Earthquake Ground Motion (OBE) and Required OBE Analyses The existing regulation (10 CFR, appendix A, section V(a)(2)) states that the maximum vibratory ground motion of the OBE is at least one half the maximum vibratory ground motion of the Safe Shutdown Earthquake ground motion. Also, the existing regulation (10 CFR, appendix A, section VI(a)(2)) states that the engineering method used to insure that structures, systems, and components are capable of withstanding the effects of the OBE shall involve the use of either a suitable dynamic analysis or a suitable qualification test. In some cases, for instance piping, these multi-facets of the OBE in the existing regulation made it possible for the OBE to have more design significance than the SSE. A decoupling of the OBE and SSE has been suggested in several documents. For instance, the NRC staff, SECY-79- 300, suggested that design for a single limiting event and inspection and evaluation for earthquakes in excess of some specified limit may be the most sound regulatory approach. NUREG-1061, ``Report of the U.S. Nuclear Regulatory Commission Piping Review Committee,'' Vol. 5, April 1985, (Table 10.1) ranked a decoupling of the OBE and SSE as third out of six high priority changes. In SECY-90-016, ``Evolutionary Light Water Reactor (LWR) Certification Issues and Their Relationship to Current Regulatory Requirements,'' the NRC staff states that it agrees that the OBE should not control the design of safety systems. Activities equivalent to OBE-SSE decoupling are also being done in foreign countries. For instance, in Germany their new design standard requires only one design basis earthquake (equivalent to the SSE). They require an inspection-level earthquake (for shutdown) of 0.4 SSE. This level was set so that the vibratory ground motion should not induce stresses exceeding the allowable stress limits originally required for the OBE design. The proposed regulation would allow the value of the OBE to be set at (i) one-third or less of the SSE, where OBE requirements are satisfied without an explicit response or design analyses being performed, or (ii) a value greater than one-third of the SSE, where analysis and design are required. There are two issues the applicant should consider in selecting the value of the OBE: first, plant shutdown is required if vibratory ground motion exceeding that of the OBE occurs (discussed below in Item 6, Required Plant Shutdown), and second, the amount of analyses associated with the OBE. An applicant may determine that at one-third of the SSE level, the probability of exceeding the OBE vibratory ground motion is too high, and the cost associated with plant shutdown for inspections and testing of equipment and structures prior to restarting the plant is unacceptable. Therefore, the applicant may voluntarily select an OBE value at some higher fraction of the SSE to avoid plant shutdowns. However, if an applicant selects an OBE value at a fraction of the SSE higher than one-third, a suitable analysis shall be performed to demonstrate that the requirements associated with the OBE are satisfied. The design shall take into account soil-structure interaction effects and the expected duration of the vibratory ground motion. The requirement associated with the OBE is that all structures, systems, and components of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the public shall remain functional and within applicable stress, strain and deformation limits when subjected to the effects of the OBE in combination with normal operating loads. As stated above, it is determined that if an OBE of one-third of the SSE is used, the requirements of the OBE can be satisfied without the applicant performing any explicit response analyses. In this case, the OBE serves the function of an inspection and shutdown earthquake. Some minimal design checks and the applicability of this position to seismic base isolation of buildings are discussed below. There is high confidence that, at this ground-motion level with other postulated concurrent loads, most critical structures, systems, and components will not exceed currently used design limits. This is ensured, in part, because PRA insights will be used to support a margins-type assessment of seismic events. A PRA-based seismic margins analysis will consider sequence-level High Confidence, Low Probability of Failures (HCLPFs) and fragilities for all sequences leading to core damage or containment failures up to approximately one and two-thirds the ground motion acceleration of the design basis SSE (Reference: Item II.N, Site- Specific Probabilistic Risk Assessment and Analysis of External Events, memorandum from Samuel J. Chilk to James M. Taylor, Subject: SECY-93- 087--Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advance Light-Water Reactor (ALWR) Designs, dated July 21, 1993. There are situations associated with current analyses where only OBE is associated with the design requirements, for example, the ultimate heat sink (see Regulatory Guide 1.27, ``Ultimate Heat Sink for Nuclear Power Plants''). In these situations, a value expressed as a fraction of the SSE response would be used in the analyses. Section VIII of this proposed rule identifies existing guides that would be revised technically to maintain the existing design philosophy. In SECY-93-087, ``Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advance Light-Water Reactor (ALWR) Designs,'' the NRC staff requested Commission approval on 42 technical and policy issues pertaining to either evolutionary LWRs, passive LWRs, or both. The issue pertaining to the elimination of the OBE is designated I.M. The NRC staff identified actions necessary for the design of structures, systems, and components when the OBE design requirement is eliminated. The staff clarified that guidelines should be maintained to ensure the functionality of components, equipment, and their supports. In addition, the staff clarified how certain design requirements are to be considered for buildings and structures that are currently designed for the OBE, but not the SSE. Also, the NRC staff has evaluated the effect on safety of eliminating the OBE from the design load combinations for selected structures, systems, and components and has developed proposed criteria for an analysis using only the SSE. Commission approval is documented in the Chilk to Taylor memorandum dated July 21, 1993, cited above. More than one earthquake response analysis for a seismic base isolated nuclear power plant design may be necessary to ensure adequate performance at all earthquake levels. Decisions pertaining to the response analyses associated with base isolated facilities will be handled on a case by case basis. 6. Required Plant Shutdow The current regulation (Section V(a)(2)) states that if vibratory ground motion exceeding that of the OBE occurs, shutdown of the nuclear power plant is required. The supplementary information to the final regulation (published November 13, 1973; 38 FR 31279, Item 6e) includes the following statement: ``A footnote has been added to Sec. 50.36(c)(2) of 10 CFR part 50 to assure that each power plant is aware of the limiting condition of operation which is imposed under section V(2) of appendix A to 10 CFR part 100. This limitation requires that if vibratory ground motion exceeding that of the OBE occurs, shutdown of the nuclear power plant will be required. Prior to resuming operations, the licensee will be required to demonstrate to the Commission that no functional damage has occurred to those features necessary for continued operation without undue risk to the health and safety of the public.'' At that time, it was the intention of the Commission to treat the Operating Basis Earthquake as a limiting condition of operation. From the statement in the Supplementary Information, the Commission directed applicants to specifically review 10 CFR Part 100 to be aware of this intention in complying with the requirements of 10 CFR 50.36. Thus, the requirement to shut down if an OBE occurs was expected to be implemented by being included among the technical specifications submitted by applicants after the adoption of Appendix A. In fact, applicants did not include OBE shutdown requirements in their technical specifications. The proposed regulation would treat plant shutdown associated with vibratory ground motion exceeding the OBE or significant plant damage as a condition in every operating license. A new Sec. 50.54(ff) would be added to the regulations to require a process leading to plant shutdown for licensees of nuclear power plants that comply with the earthquake engineering criteria in Paragraph IV(a)(3) of Proposed Appendix S, ``Earthquake Engineering Criteria for Nuclear Power Plants,'' to 10 CFR part 50. Immediate shutdown could be required until it is determined that structures, systems, and components needed for safe shutdown are still functional. Draft Regulatory Guide DG-1034, ``Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Post-Earthquake Actions,'' is being developed to provide guidance acceptable to the NRC staff for determining whether or not vibratory ground motion exceeding the OBE ground motion or significant plant damage had occurred and the timing of nuclear power plant shutdown. The guidance is based on criteria developed by the Electric Power Research Institute (EPRI). The decision to shut down the plant should be made within eight hours after the earthquake. The data from the seismic instrumentation, coupled with information obtained from a plant walk down, are used to make the determination of when the plant should be shut down, if it has not already been shut down by operational perturbations resulting from the seismic event. The guidance being developed in Draft Regulatory Guide DG-1034 is based on two assumptions, first, that the nuclear power plant has operable seismic instrumentation, including the equipment and software required to process the data within four hours after an earthquake, and second, that the operator walk down inspections can be performed in approximately four to eight hours depending on the number of personnel conducting the inspection. The regulation also includes a provision that requires the licensee to consult with the Commission and to propose a plan for the timely, safe shutdown of the nuclear power plant if systems, structures, or components necessary for a safe shutdown or to maintain a safe shutdown are not available. (This unavailability may be due to earthquake related damage.) Draft Regulatory Guide DG-1035, ``Restart of a Nuclear Power Plant Shut Down by a Seismic Event,'' is being developed to provide guidelines that are acceptable to the NRC staff for performing inspections and tests of nuclear power plant equipment and structures prior to plant restart. This guidance is also based on EPRI reports. Prior to resuming operations, the licensee must demonstrate to the Commission that no functional damage has occurred to those features necessary for continued operation without undue risk to the health and safety of the public. The results of post-shutdown inspections, operability checks, and surveillance tests must be documented in written reports and submitted to the Director, Office of Nuclear Reactor Regulation. The licensee shall not resume operation until authorized to do so by the Director, Office of Nuclear Reactor Regulation. 7. Clarify Interpretations In Sec. 100.23 to 10 CFR part 100, changes have been made to resolve questions of interpretation. As an example, definitions and required investigations stated in the proposed regulation would be significantly changed to eliminate or modify phrases that were more applicable to only the western part of the United States. The institutional definition for ``safety-related structures, systems, and components'' is drawn from appendix A to part 100 under III(c) and VI(a). With the proposed relocation of the earthquake engineering criteria to appendix S to part 50 and the proposed relocation and modification to dose guidelines in Sec. 50.34(a)(1), the definition of safety-related structures, systems, and components is included in part 50 definitions with reference to both the part 100 and part 50 dose guidelines. VI. Related Regulatory Guides and Standard Review Plan Section The NRC is developing the following draft regulatory guides and standard review plan sections to provide prospective licensees with the necessary guidance for implementing the proposed regulation. The notice of availability for these materials will be published in a later issue of the Federal Register. 1. DG-1032, ``Identification and Characterization of Seismic Sources and Determination of Shutdown Earthquake Ground Motions.'' The draft guide provides general guidance and recommendations, describes acceptable procedures and provides a list of references that present acceptable methodologies to identify and characterize capable tectonic sources and seismogenic sources. Section V.B.3 of this Proposed rule describes the key elements. 2. DG-1033, Third Proposed Revision 2 to Regulatory Guide 1.12, ``Nuclear Power Plant Instrumentation for Earthquakes.'' The draft guide describes seismic instrumentation type and location, operability, characteristics, installation, actuation, and maintenance that are acceptable to the NRC staff. 3. DG-1034, ``Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Post-Earthquake Actions.'' The draft guide provides guidelines that are acceptable to the NRC staff for a timely evaluation of the recorded seismic instrumentation data and to determine whether or not plant shutdown is required. 4. DG-1035, ``Restart of a Nuclear Power Plant Shut Down by a Seismic Event.'' The draft guide provides guidelines that are acceptable to the NRC staff for performing inspections and tests of nuclear power plant equipment and structures prior to restart of a plant that has been shut down because of a seismic event. 5. Draft Standard Review Plan Section 2.5.1, Proposed Revision 3, ``Basic Geologic and Seismic Information.'' The draft describes procedures to assess the adequacy of the geologic and seismic information cited in support of the applicant's conclusions concerning the suitability of the plant site. 6. Draft Standard Review Plan Section 2.5.2, Second Proposed Revision 3 ``Vibratory Ground Motion.'' The draft describes procedures to assess the ground motion potential of seismic sources at the site and to assess the adequacy of the SSE. 7. Draft Standard Review Plan Section 2.5.3, Proposed Revision 3, ``Surface Faulting.'' The draft describes procedures to assess the adequacy of the applicant's submittal related to the existence of a potential for surface faulting affecting the site. 8. DG-4003, Second Proposed Revision 2 to Regulatory Guide 4.7, ``General Site Suitability Criteria for Nuclear Power Plants.'' This guide discusses the major site characteristics related to public health and safety and environmental issues that the NRC staff considers in determining the suitability of sites. VII. Future Regulatory Action Several existing regulatory guides will be revised to incorporate editorial changes or maintain the existing design or analysis philosophy. These guides will be issued subsequent to the publication of the final regulations that would implement this proposed action. The following regulatory guides will be revised to incorporate editorial changes, for example to reference new sections to part 100 or appendix S to part 50. No technical changes will be made in these regulatory guides. 1. 1.57, ``Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components.'' 2. 1.59, ``Design Basis Floods for Nuclear Power Plants.'' 3. 1.60, ``Design Response Spectra for Seismic Design of Nuclear Power Plants.'' 4. 1.83, ``Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes.'' 5. 1.92, ``Combining Modal Responses and Spatial Components in Seismic Response Analysis.'' 6. 1.102, ``Flood Protection for Nuclear Power Plants.'' 7. 1.121, ``Bases for Plugging Degraded PWR Steam Generator Tubes.'' 8. 1.122, ``Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components.'' The following regulatory guides will be revised to update the design or analysis philosophy, for example, to change OBE to a fraction of the SSE: 1. 1.27, ``Ultimate Heat Sink for Nuclear Power Plants.'' 2. 1.100, ``Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power Plants.'' 3. 1.124, ``Service Limits and Loading Combinations for Class 1 Linear-Type Component Supports.'' 4. 1.130, ``Service Limits and Loading Combinations for Class 1 Plate-and-Shell-Type Component Supports.'' 5. 1.132, ``Site Investigations for Foundations of Nuclear Power Plants.'' 6. 1.138, ``Laboratory Investigations of Soils for Engineering Analysis and Design of Nuclear Power Plants.'' 7. 1.142, ``Safety-Related Concrete Structures for Nuclear Power Plants (Other than Reactor Vessels and Containments).'' 8. 1.143, ``Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants.'' Minor and conforming changes to other Regulatory Guides and standard review plan sections as a result of proposed changes in the nonseismic criteria are also planned. If substantive changes are made during the revisions, the applicable guides will be issued for public comment as draft guides. VIII. Referenced Documents An interested person may examine or obtain copies of the documents referenced in this proposed rule as set out below. Copies of NUREG-0625, NUREG-1150, and NUREG/CR-2239 may be purchased from the Superintendent of Documents, U.S. Government Printing Office, Mail Stop SSOP, Washington, DC 20402-9328. Copies are also available from the National Technical Information Service, 5285 Port Royal Road, Springfield, VA 22161. A copy is also available for inspection and copying for a fee in the NRC Public Document Room, 2120 L Street, NW. (Lower Level), Washington, DC. Copies of issued regulatory guides may be purchased from the Government Printing Office (GPO) at the current GPO price. Information on current GPO prices may be obtained by contacting the Superintendent of Documents, U.S. Government Printing Office, Mail Stop SSOP, Washington, DC 20402-9328. Issued guides may also be purchased from the National Technical Information Service on a standing order basis. Details on this service may be obtained by writing NTIS, 5826 Port Royal Road, Springfield, VA 22161. SECY 79-300, SECY 90-016, SECY 93-087, and WASH-1400 are available for inspection and copying for a fee at the Commission's Public Document Room, 2120 L Street, NW. (Lower Level), Washington, DC. IX. Submission of Comments in Electronic Format The comment process will be improved if each comment is identified with the document title, section heading, and paragraph number addressed. Commenters are encouraged to submit, in addition to the original paper copy, a copy of the letter in electronic format on 5.25 or 3.5 inch computer diskette; IBM PC/DOS or MS/DOS format. Data files should be provided in one of the following formats: WordPerfect, IBM Document Content Architecture/Revisable-Form-Text (DCA/RFT), or unformatted ASCII code. The format and version should be identified on the diskette's external label. X. Questions In addition to soliciting comments on all aspects of this rulemaking, the Commission specifically requests comments on the following questions. A. Nonseismic Criteria 1. Should the dose acceptance criteria be modified from 25 rem whole body and 300 rem to the thyroid to utilize the concept of total effective dose equivalent (TEDE), and if so, what TEDE value should be adopted? 2. Assuming that a dose acceptance criterion of 25 rem total effective dose equivalent (TEDE) is adopted, should an organ limitation or ``capping'' dose be included, and if so, what should such a limit be? XI. Finding of No Significant Environmental Impact: Availability The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in Subpart A of 10 CFR Part 51, that this proposed regulation, if adopted, would not be a major Federal action significantly affecting the quality of the human environment and therefore an environmental impact statement is not required. The revisions associated with the reactor siting criteria in 10 CFR part 100 and the relocation of the plant design requirements from 10 CFR part 100 to 10 CFR Part 50 have been evaluated against the current requirements. The Commission has concluded that relocating the requirement for a dose calculation to Part 50 and adding more specific site criteria to part 100 does not decrease the protection of the public health and safety over the current regulations. The proposed amendments do not affect nonradiological plant effluents and have no other environmental impact. The addition of Sec. 100.23 to 10 CFR part 100, and the addition of appendix S to 10 CFR part 50, will not change the radiological environmental impact offsite. Onsite occupational radiation exposure associated with inspection and maintenance will not change. These activities are principally associated with base line inspections of structures, equipment, and piping, and with maintenance of seismic instrumentation. Base line inspections are needed to differentiate between pre-existing conditions at the nuclear power plant and earthquake related damage. The structures, equipment and piping selected for these inspections are those routinely examined by plant operators during normal plant walkdowns and inspections. Routine maintenance of seismic instrumentation ensures its operability during earthquakes. The location of the seismic instrumentation is similar to that in the existing nuclear power plants. The proposed amendments do not affect nonradiological plant effluents and have no other environmental impact. The environmental assessment and finding of no significant impact on which this determination is based are available for inspection at the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC. Single copies of the environmental assessment and finding of no significant impact are available from Mr. Leonard Soffer, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 415-6574, or Dr. Andrew Murphy, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 415-6010. XII. Paperwork Reduction Act Statement This proposed regulation amends information collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). This proposed regulation has been submitted to the Office of Management and Budget for review and approval of the paperwork requirements. There is no public reporting burden related to the nonseismic siting criteria. Public reporting burden for the collection of information related to the seismic and earthquake engineering criteria is estimated to average 800,000 hours per response, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information. Send comments regarding this burden estimate or any other aspect of this collection of information, including suggestions for reducing this burden, to the Information and Records Management Branch (T-6 F33), U.S. Nuclear Regulatory Commission, Washington, DC 20555; and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0011 and 3150-0093), Office of Management and Budget, Washington, DC 20503. XIII. Regulatory Analysis The Commission has prepared a draft regulatory analysis on this proposed regulation. The analysis examines the costs and benefits of the alternatives considered by the Commission. The draft analysis is available for inspection in the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC. Single copies of the analysis are available from Mr. Leonard Soffer, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 415-6574, or Dr. Andrew J. Murphy, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 415-6010. The Commission requests public comment on the draft regulatory analysis. Comments on the draft analysis may be submitted to the NRC as indicated under the ADDRESSES heading. XIV. Regulatory Flexibility Certification In accordance with the Regulatory Flexibility Act of 1980 (5 U.S.C. 605(b)), the Commission certifies that this proposed regulation will not, if promulgated, have a significant economic impact on a substantial number of small entities. This proposed regulation affects only the licensing and operation of nuclear power plants. Nuclear power plant site applicants do not fall within the definition of small businesses as defined in Section 3 of the Small Business Act (15 U.S.C. 632), the Small Business Size Standards of the Small Business Administrator (13 CFR part 121), or the Commission's Size Standards (56 FR 56671; November 6, 1991). XV. Backfit Analysis The NRC has determined that the backfit rule, 10 CFR 50.109, does not apply to this proposed regulation, and therefore, a backfit analysis is not required for this proposed regulation because these amendments do not involve any provisions that would impose backfits as defined in 10 CFR 50.109(a)(1). The proposed regulation would apply only to applicants for future nuclear power plant construction permits, preliminary design approval, final design approval, manufacturing licenses, early site reviews, operating licenses, and combined operating licenses. List of Subjects 10 CFR Part 50 Antitrust, Classified information, Criminal penalty, Fire protection, Intergovernmental relations, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements. 10 CFR Part 52 Administrative practice and procedure, Antitrust, Backfitting, Combined license, Early site permit, Emergency planning, Fees, Inspection, Limited work authorization, Nuclear power plants and reactors, Probabilistic risk assessment, Prototype, Reactor siting criteria, Redress of site, Reporting and recordkeeping requirements, Standard design, Standard design certification. 10 CFR Part 100 Nuclear power plants and reactors, Reactor siting criteria. For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 553, the NRC is proposing to adopt the following amendments to 10 CFR parts 50, 52 and 100. PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES 1. The authority citation for part 50 continues to read as follows: Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68 Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 1244, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201, 2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88 Stat. 1242, as amended, 1244, 1246, (42 U.S.C. 5841, 5842, 5846). Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat. 2951 as amended by Pub. L. 102-486, sec. 2902, 106 Stat. 3123, (42 U.S.C. 5851). Section 50.10 also issued under secs. 101, 185, 68 Stat. 936, 955 as amended (42 U.S.C. 2131, 2235), sec. 102, Pub. L. 91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 50.54(dd) and 50.103 also issued under sec. 108, 68 Stat. 939, as amended (42 U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56 also issued under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a, 50.55a and Appendix Q also issued under sec. 102, Pub. L. 91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 50.58, 50.91 and 50.92 also issued under Pub. L. 97-415, 96 Stat. 2073 (42 U.S.C. 2239). Section 50.78 also issued under sec. 122, 68 Stat. 939 (42 U.S.C. 2152). Sections 50.80-50.81 also issued under sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234). Appendix F also issued under sec. 187, 68 Stat. 955 (42 U.S.C. 2237). 2. Section 50.2 is amended by adding in alphabetical order the definitions for Committed dose equivalent, Committed effective dose equivalent, Deep-dose equivalent, Exclusion area, Low population zone, Safety-related structures, systems, and components and Total effective dose equivalent to read as follows: Sec. 50.2 Definitions. * * * * * Committed dose equivalent means the dose equivalent to organs or tissues of reference that will be received from an intake of radioactive material by an individual during the 50-year period following the intake. Committed effective dose equivalent is the sum of the products of the weighting factors applicable to each of the body organs or tissues that are irradiated and the committed dose equivalent to these organs or tissues. * * * * * Deep-dose equivalent, which applies to external whole-body exposure, is the dose equivalent at a tissue depth of 1 cm (1000 mg/ cm2). * * * * * Exclusion area means that area surrounding the reactor, in which the reactor licensee has the authority to determine all activities including exclusion or removal of personnel and property from the area. This area may be traversed by a highway, railroad, or waterway, provided these are not so close to the facility as to interfere with normal operations of the facility and provided appropriate and effective arrangements are made to control traffic on the highway, railroad, or waterway, in case of emergency, to protect the public health and safety. Residence within the exclusion area shall normally be prohibited. In any event, residents shall be subject to ready removal in case of necessity. Activities unrelated to operation of the reactor may be permitted in an exclusion area under appropriate limitations, provided that no significant hazards to the public health and safety will result. * * * * * Low population zone means the area immediately surrounding the exclusion area which contains residents, the total number and density of which are such that there is a reasonable probability that appropriate protective measures could be taken in their behalf in the event of a serious accident. These guides do not specify a permissible population density or total population within this zone because the situation may vary from case to case. Whether a specific number of people can, for example, be evacuated from a specific area, or instructed to take shelter, on a timely basis will depend on many factors such as location, number and size of highways, scope and extent of advance planning, and actual distribution of residents within the area. * * * * * Safety-related structures, systems, and components means those structures, systems, and components that are relied on to remain functional during and following design basis (postulated) events to assure: (1) The integrity of the reactor coolant pressure boundary, (2) The capability to shutdown the reactor and maintain it in a safe shutdown condition, and (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in Sec. 50.34(a)(1) or Sec. 100.11 of this chapter. * * * * * Total effective dose equivalent (TEDE) means the sum of the deep- dose equivalent (for external exposures) and the committed effective dose equivalent (for internal exposures). * * * * * 3. In Sec. 50.8, paragraph (b) is revised to read as follows: Sec. 50.8 Information collection requirements: OMB approval. * * * * * (b) The approved information collection requirements contained in this part appear in Secs. 50.30, 50.33, 50.33a, 50.34, 50.34a, 50.35, 50.36, 50.36a, 50.48, 50.49, 50.54, 50.55, 50.55a, 50.59, 50.60, 50.61, 50.63, 50.64, 50.65, 50.71, 50.72, 50.80, 50.82, 50.90, 50.91, and Appendices A, B, E, G, H, I, J, K, M, N, O, Q, R, and S. * * * * * 4. In Sec. 50.34, footnotes 6, 7, and 8 are redesignated as footnotes 8, 9 and 10 and paragraph (a)(1) is revised and paragraphs (a)(12), (b)(10), and (b)(11) are added to read as follows: Sec. 50.34 Contents of applications; technical information. (a) * * * (1) Stationary power reactor applicants for a construction permit pursuant to this part, or a design certification or combined license pursuant to Part 52 of this chapter who apply on or after [EFFECTIVE DATE OF THE FINAL RULE], shall comply with paragraph (a)(1)(ii) of this section. All other applicants for a construction permit pursuant to this part or a design certification or combined license pursuant to part 52 of this chapter, shall comply with paragraph (a)(1)(i) of this section. (i) A description and safety assessment of the site on which the facility is to be located, with appropriate attention to features affecting facility design. Special attention should be directed to the site evaluation factors identified in part 100 of this chapter. The assessment must contain an analysis and evaluation of the major structures, systems and components of the facility which bear significantly on the acceptability of the site under the site evaluation factors identified in part 100 of this chapter, assuming that the facility will be operated at the ultimate power level which is contemplated by the applicant. With respect to operation at the projected initial power level, the applicant is required to submit information prescribed in paragraphs (a)(2) through (a)(8) of this section, as well as the information required by this paragraph, in support of the application for a construction permit, or a design approval. (ii) A description and safety assessment of the site and a safety assessment of the facility. It is expected that reactors will reflect through their design , construction and operation an extremely low probability for accidents that could result in the release of significant quantities of radioactive fission products. The following power reactor design characteristics and proposed operation will be taken into consideration by the Commission: (A) Intended use of the reactor including the proposed maximum power level and the nature and inventory of contained radioactive materials; (B) The extent to which generally accepted engineering standards are applied to the design of the reactor; (C) The extent to which the reactor incorporates unique, unusual or enhanced safety features having a significant bearing on the probability or consequences of accidental release of radioactive materials; (D) The safety features that are to be engineered into the facility and those barriers that must be breached as a result of an accident before a release of radioactive material to the environment can occur. Special attention must be directed to plant design features intended to mitigate the radiological consequences of accidents. In performing this assessment, an applicant shall assume a fission product release\6\ from the core into the containment assuming that the facility is operated at the ultimate power level contemplated. The applicant shall perform an evaluation and analysis of the postulated fission product release, using the expected demonstrable containment leak rate and any fission product cleanup systems intended to mitigate the consequences of the accidents, together with applicable site characteristics, including site meteorology, to evaluate the offsite radiological consequences. Site characteristics must comply with part 100 of this chapter. The evaluation must determine that: --------------------------------------------------------------------------- \6\The fission product release assumed for this evaluation should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events. Such accidents have generally been assumed to result in substantial meltdown of the core with subsequent release into the containment of appreciable quantities of fission products. --------------------------------------------------------------------------- (1) An individual located at any point on the boundary of the exclusion area for any 2 hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 25\7\ rem total effective dose equivalent (TEDE). --------------------------------------------------------------------------- \7\A whole body does of 25 rem has been stated to correspond numerically to the once in a lifetime accidental or emergency dose for radiation workers which, according to NCRP recommendations at the time could be disregarded in the determination of their radiation expousre status (see NBS Handbook 69 dated June 5, 1959). However, its use is not intended to imply that this number constitutes an acceptable limit for an emergency does to the public under accident conditions. Rather, this does value has been set forth in this section as a reference value, which can be used in the evaluation of plant design features with respect to postulated reactor accidents, in order to assure that such designs provide assurance of low risk of public exposure to radiation, in the event of such accidents. --------------------------------------------------------------------------- (2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a radiation dose in excess of 25 rem total effective dose equivalent (TEDE). (E) With respect to operation at the projected initial power level, the applicant is required to submit information prescribed in paragraphs (a)(2) through (a)(8) of this section, as well as the information required by this paragraph, in support of the application for a construction permit, or a design approval. * * * * * (12) On or after [EFFECTIVE DATE OF THE FINAL RULE], stationary power reactor applicants who apply for a construction permit pursuant to this part, or a design certification or combined license pursuant to part 52 of this chapter, as partial conformance to General Design Criterion 2 of appendix A to this part, shall comply with the earthquake engineering criteria in appendix S of this part. (b) * * * (10) On or after [EFFECTIVE DATE OF THE FINAL RULE], stationary power reactor applicants who apply for an operating license pursuant to this part, or a design certification or combined license pursuant to part 52 of this chapter, as partial conformance to General Design Criterion 2 of appendix A to this part, shall comply with the earthquake engineering criteria of appendix S to this part. However, if the construction permit was issued prior to [EFFECTIVE DATE OF THE FINAL RULE], the stationary power reactor applicant shall comply with the earthquake engineering criteria in Section VI of appendix A to part 100 of this chapter. (11) On or after [EFFECTIVE DATE OF THE FINAL RULE], stationary power reactor applicants who apply for an operating license pursuant to this Part, or a combined license pursuant to part 52 of this chapter, shall provide a description and safety assessment of the site and of the facility as in Sec. 50.34(a)(1)(ii) of this part. * * * * * 5. In Sec. 50.54, paragraph (ff) is added to read as follows: Sec. 50.54 Conditions of licenses. * * * * * (ff) For licensees of nuclear power plants that have implemented the earthquake engineering criteria in appendix S of this part, plant shutdown is required as provided in paragraph IV(a)(3) of appendix S. Prior to resuming operations, the licensee shall demonstrate to the Commission that no functional damage has occurred to those features necessary for continued operation without undue risk to the health and safety of the public and the licensing basis is maintained. 6. Appendix S to Part 50 is added to read as follows: Appendix S to Part 50--Earthquake Engineering Criteria for Nuclear Power Plants General Information This appendix applies to applicants for a design certification or combined license pursuant to part 52 of this chapter or a construction permit or operating license pursuant to Part 50 of this chapter on or after [EFFECTIVE DATE OF THE FINAL RULE]. However, if the construction permit was issued prior to [EFFECTIVE DATE OF THE FINAL RULE], the operating license applicant shall comply with the earthquake engineering criteria in Section VI of appendix A to 10 CFR part 100. I. Introduction Each applicant for a construction permit, operating license, design certification, or combined license is required by Sec. 50.34(a)(12), (b)(10), and General Design Criterion 2 of appendix A to this part to design nuclear power plant structures, systems, and components important to safety to withstand the effects of natural phenomena, such as earthquakes, without loss of capability to perform their safety functions. Also, as specified in Sec. 50.54(ff), nuclear power plants that have implemented the earthquake engineering criteria described herein must shut down if the criteria in paragraph IV(a)(3) of this appendix are exceeded. These criteria implement General Design Criterion 2 insofar as it requires structures, systems, and components important to safety to withstand the effects of earthquakes. II. Scope The evaluations described in this appendix are within the scope of investigations permitted by Sec. 50.10(c)(1). III. Definitions As used in these criteria: Combined license means a combined construction permit and operating license with conditions for a nuclear power facility issued pursuant to subpart C of part 52 of this chapter. Design Certification means a Commission approval, issued pursuant to subpart B of part 52 of this chapter, of a standard design for a nuclear power facility. A design so approved may be referred to as a ``certified standard design.'' The Operating Basis Earthquake Ground Motion (OBE) is the vibratory ground motion for which those features of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the public will remain functional. The Operating Basis Earthquake Ground Motion is only associated with plant shutdown and inspection unless specifically selected by the applicant as a design input. A response spectrum is a plot of the maximum responses (acceleration, velocity, or displacement) of idealized single- degree-of-freedom oscillators as a function of the natural frequencies of the oscillators for a given damping value. The response spectrum is calculated for a specified vibratory motion input at the oscillators' supports. The Safe Shutdown Earthquake Ground Motion (SSE) is the vibratory ground motion for which certain structures, systems, and components must be designed to remain functional. The structures, systems, and components required to withstand the effects of the Safe Shutdown Earthquake Ground Motion or surface deformation are those necessary to assure: (1) The integrity of the reactor coolant pressure boundary, (2) The capability to shut down the reactor and maintain it in a safe shutdown condition, or (3) The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guideline exposures of Sec. 50.34(a)(1)(ii). Surface deformation is distortion of geologic strata at or near the ground surface by the processes of folding or faulting as a result of various earth forces. Tectonic surface deformation is associated with earthquake processes. IV. Application to Engineering Design The following are pursuant to the seismic and geologic design basis requirements of Sec. 100.23 of this chapter: (a) Vibratory Ground Motion. (1) Safe Shutdown Earthquake Ground Motion. The Safe Shutdown Earthquake Ground Motion must be characterized by free-field ground motion response spectra at the free ground surface. In view of the limited data available on vibratory ground motions of strong earthquakes, it usually will be appropriate that the design response spectra be smoothed spectra. The horizontal component of the Safe Shutdown Earthquake Ground Motion in the free-field at the foundation level of the structures must be an appropriate response spectrum with a peak ground acceleration of at least 0.1g. The nuclear power plant must be designed so that, if the Safe Shutdown Earthquake Ground Motion occurs, certain structures, systems, and components will remain functional and within applicable stress, strain, and deformation limits. In addition to seismic loads, applicable concurrent normal operating, functional, and accident-induced loads must be taken into account in the design of these safety-related structures, systems, and components. The design of the nuclear power plant must also take into account the possible effects of the Safe Shutdown Earthquake Ground Motion on the facility foundations by ground disruption, such as fissuring, lateral spreads, differential settlement, liquefaction, and landsliding, as required in Sec. 100.23 to part 100 of this chapter. The required safety functions of structures, systems, and components must be assured during and after the vibratory ground motion associated with the Safe Shutdown Earthquake Ground Motion through design, testing, or qualification methods. The evaluation must take into account soil-structure interaction effects and the expected duration of vibratory motion. It is permissible to design for strain limits in excess of yield strain in some of these safety-related structures, systems, and components during the Safe Shutdown Earthquake Ground Motion and under the postulated concurrent loads, provided the necessary safety functions are maintained. (2) Operating Basis Earthquake Ground Motion. (i) The Operating Basis Earthquake Ground Motion must be characterized by response spectra. The value of the Operating Basis Earthquake Ground Motion must be set to one of the following choices: (A) One-third or less of the Safe Shutdown Earthquake Ground Motion design response spectra. The requirements associated with this Operating Basis Earthquake Ground Motion in paragraph (a)(2)(i)(B)(I) can be satisfied without the applicant performing explicit response or design analyses, or (B) A value greater than one-third of the Safe Shutdown Earthquake Ground Motion design response spectra. Analysis and design must be performed to demonstrate that the requirements associated with this Operating Basis Earthquake Ground Motion in paragraph (a)(2)(i)(B)(I) are satisfied. The design must take into account soil-structure interaction effects and the duration of vibratory ground motion. (I) When subjected to the effects of the Operating Basis Earthquake Ground Motion in combination with normal operating loads, all structures, systems, and components of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the public must remain functional and within applicable stress, strain, and deformation limits. (3) Required Plant Shutdown. If vibratory ground motion exceeding that of the Operating Basis Earthquake Ground Motion or if significant plant damage occurs, the licensee must shut down the nuclear power plant. If systems, structures, or components necessary for the safe shutdown of the nuclear power plant are not available after the occurrence of the OBE, the licensee must consult with the Commission and must propose a plan for the timely, safe shutdown of the nuclear power plant. Prior to resuming operations, the licensee must demonstrate to the Commission that no functional damage has occurred to those features necessary for continued operation without undue risk to the health and safety of the public. (4) Required Seismic Instrumentation. Suitable instrumentation must be provided so that the seismic response of nuclear power plant features important to safety can be evaluated promptly after an earthquake. (b) Surface Deformation. The potential for surface deformation must be taken into account in the design of the nuclear power plant by providing reasonable assurance that in the event of deformation, certain structures, systems, and components will remain functional. In addition to surface deformation induced loads, the design of safety features must take into account seismic loads, including aftershocks, and applicable concurrent functional and accident- induced loads. The design provisions for surface deformation must be based on its postulated occurrence in any direction and azimuth and under any part of the nuclear power plant, unless evidence indicates this assumption is not appropriate, and must take into account the estimated rate at which the surface deformation may occur. (c) Seismically Induced Floods and Water Waves and Other Design Conditions. Seismically induced floods and water waves from either locally or distantly generated seismic activity and other design conditions determined pursuant to Sec. 100.23 of this chapter must be taken into account in the design of the nuclear power plant so as to prevent undue risk to the health and safety of the public. PART 52--EARLY SITE PERMITS; STANDARD DESIGN CERTIFICATIONS; AND COMBINED LICENSES FOR NUCLEAR POWER PLANTS 7. The authority citation for part 52 continues to read as follows: Authority: Secs. 103, 104, 161, 182, 183, 186, 189, 68 Stat. 936, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 1244, as amended (42 U.S.C. 2133, 2201, 2232, 2233, 2236, 2239, 2282); secs. 201, 202, 206, 88 Stat. 1242, 1244, 1246, as amended (42 U.S.C. 5841, 5842, 5846). 8. In Sec. 52.17, the introductory text of paragraph (a)(1) and paragraph (a)(1)(vi) are revised to read as follows: Sec. 52.17 Contents of applications. (a)(1) The application must contain the information required by Sec. 50.33 (a) through (d), the information required by Sec. 50.34 (a)(12) and (b)(10), and to the extent approval of emergency plans is sought under paragraph (b)(2)(ii) of this section, the information required by Sec. 50.33 (g) and (j), and Sec. 50.34 (b)(6)(v). The application must also contain a description and safety assessment of the site on which the facility is to be located. The assessment must contain an analysis and evaluation of the major structures, systems, and components of the facility that bear significantly on the acceptability of the site under the radiological consequence evaluation factors identified in Sec. 50.34(a)(1) of this chapter. Site characteristics must comply with part 100 of this chapter. In addition, the application should describe the following: * * * * * (vi) The seismic, meteorological, hydrologic, and geologic characteristics of the proposed site; * * * * * PART 100--REACTOR SITE CRITERIA 9. and 10. The authority citation for Part 100 continues to read as follows: Authority: Secs. 103, 104, 161, 182, 68 Stat. 936, 937, 948, 953, as amended (42 U.S.C. 2133, 2134, 2201, 2232); sec. 201, as amended, 202, 88 Stat. 1242, as amended, 1244 (42 U.S.C. 5841, 5842). 11. Section 100.1 is revised to read as follows: Sec. 100.1 Purpose. (a) The purpose of this part is to establish approval requirements for proposed sites for stationary power and testing reactors subject to part 50 or part 52 of this chapter. (b) There exists a substantial base of knowledge regarding power reactor siting, design, construction and operation. This base reflects that the primary factors that determine public health and safety are the reactor design, construction and operation. (c) Siting factors and criteria are important in assuring that radiological doses from normal operation and postulated accidents will be acceptably low, that natural phenomena and potential man-made hazards will be appropriately accounted for in the design of the plant, and that the site characteristics are amenable to the development of adequate emergency plans to protect the public and adequate security measures to protect the plant. (d) This approach incorporates the appropriate standards and criteria for approval of stationary power and testing reactor sites. The Commission intends to carry out a traditional defense-in-depth approach with regard to reactor siting to ensure public safety. Siting away from densely populated centers has been and will continue to be an important factor in evaluating applications for site approval. 12. Section 100.2 is revised to read as follows: Sec. 100.2 Scope. The siting requirements contained in this part apply to applications for site approval for the purpose of constructing and operating stationary power and testing reactors pursuant to the provisions of parts 50 or 52 of this chapter. 13. Section 100.3 is revised to read as follows: Sec. 100.3 Definitions. As used in this part: Combined license means a combined construction permit and operating license with conditions for a nuclear power facility issued pursuant to subpart C of part 52 of this chapter. Early site permit means a Commission approval, issued pursuant to subpart A of part 52 of this chapter, for a site or sites for one or more nuclear power facilities. Exclusion area means that area surrounding the reactor, in which the reactor licensee has the authority to determine all activities including exclusion or removal of personnel and property from the area. This area may be traversed by a highway, railroad, or waterway, provided these are not so close to the facility as to interfere with normal operations of the facility and provided appropriate and effective arrangements are made to control traffic on the highway, railroad, or waterway, in case of emergency, to protect the public health and safety. Residence within the exclusion area shall normally be prohibited. In any event, residents shall be subject to ready removal in case of necessity. Activities unrelated to operation of the reactor may be permitted in an exclusion area under appropriate limitations, provided that no significant hazards to the public health and safety will result. Low population zone means the area immediately surrounding the exclusion area which contains residents, the total number and density of which are such that there is a reasonable probability that appropriate protective measures could be taken in their behalf in the event of a serious accident. These guides do not specify a permissible population density or total population within this zone because the situation may vary from case to case. Whether a specific number of people can, for example, be evacuated from a specific area, or instructed to take shelter, on a timely basis will depend on many factors such as location, number and size of highways, scope and extent of advance planning, and actual distribution of residents within the area. Population center distance means the distance from the reactor to the nearest boundary of a densely populated center containing more than about 25,000 residents. Power reactor means a nuclear reactor of a type described in Secs. 50.21(b) or 50.22 of this chapter designed to produce electrical or heat energy. A Response spectrum is a plot of the maximum responses (acceleration, velocity, or displacement) of idealized single-degree- of-freedom oscillators as a function of the natural frequencies of the oscillators for a given damping value. The response spectrum is calculated for a specified vibratory motion input at the oscillators' supports. The Safe Shutdown Earthquake Ground Motion is the vibratory ground motion for which certain structures, systems, and components must be designed pursuant to Appendix S to part 50 of this chapter to remain functional. Surface deformation is distortion of geologic strata at or near the ground surface by the processes of folding or faulting as a result of various earth forces. Tectonic surface deformation is associated with earthquake processes. Testing reactor means a testing facility as defined in Sec. 50.2 of this chapter. 14. Section 100.4 is added to read as follows: Sec. 100.4 Communications. Except where otherwise specified in this part, all correspondence, reports, applications, and other written communications submitted pursuant to 10 CFR part 100 should be addressed to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555, and copies sent to the appropriate Regional Office and Resident Inspector. Communications and reports may be delivered in person at the Commission's offices at 2120 L Street, NW., Washington, DC, or at 11555 Rockville Pike, Rockville, Maryland. 15. Section 100.8 is revised to read as follows: Sec. 100.8 Information collection requirements: OMB approval. (a) The Nuclear Regulatory Commission has submitted the information collection requirements contained in this part to the Office of Management and Budget (OMB) for approval as required by the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). OMB has approved the information collection requirements contained in this part under control number 3150-0093. (b) The approved information collection requirements contained in this part appear in Sec. 100.23 and Appendix A. 16. A heading for Subpart A (consisting of Secs. 100.10 and 100.11) is added directly before Sec. 100.10 and Secs. 100.10 and 100.11 are revised to read as follows: Subpart A--Evaluation Factors for Stationary Power Reactor Site Applications Before [EFFECTIVE DATE OF THE FINAL RULE] and for Testing Reactors Sec. 100.10 Factors to be considered when evaluating sites. 100.11 Determination of exclusion area, low population zone, and population center distance. Sec. 100.10 Factors to be considered when evaluating sites. Factors considered in the evaluation of sites include those relating both to the proposed reactor design and the characteristics peculiar to the site. It is expected that reactors will reflect through their design, construction and operation an extremely low probability for accidents that could result in release of significant quantities of radioactive fission products. In addition, the site location and the engineered features included as safeguards against the hazardous consequences of an accident, should one occur, should insure a low risk of public exposure. In particular, the Commission will take the following factors into consideration in determining the acceptability of a site for a power or testing reactor: (a) Characteristics of reactor design and proposed operation including-- (1) Intended use of the reactor including the proposed maximum power level and the nature and inventory of contained radioactive materials; (2) The extent to which generally accepted engineering standards are applied to the design of the reactor; (3) The extent to which the reactor incorporates unique or unusual features having a significant bearing on the probability or consequences of accidental release of radioactive materials; (4) The safety features that are to be engineered into the facility and those barriers that must be breached as a result of an accident before a release of radioactive material to the environment can occur. (b) Population density and use characteristics of the site environs, including the exclusion area, low population zone, and the population center distance. (c) Physical characteristics of the site, including seismology, meteorology, geology, and hydrology. (1) Appendix A to Part 100, ``Seismic and Geologic Siting Criteria for Nuclear Power Plants'' describes the nature of investigations required to obtain the geologic and seismic data necessary to determine site suitability and to provide reasonable assurance that a nuclear power plant can be constructed and operated at a proposed site without undue risk to the health and safety of the public. It describes procedures for determining the quantitative vibratory ground motion design basis at a site due to earthquakes and describes information needed to determine whether and to what extent a nuclear power plant need be designed to withstand the effects of surface faulting. (2) Meteorological conditions at the site and in the surrounding area should be considered. (3) Geological and hydrological characteristics of the proposed site may have a bearing on the consequences of an escape of radioactive material from the facility. Special precautions should be planned if a reactor is to be located at a site where a significant quantity of radioactive effluent might accidentally flow into nearby streams or rivers or might find ready access to underground water tables. (d) Where unfavorable physical characteristics of the site exist, the proposed site may nevertheless be found to be acceptable if the design of the facility includes appropriate and adequate compensating engineering safeguards. Sec. 100.11 Determination of exclusion area, low population zone, and population center distance. (a) As an aid in evaluating a proposed site, an applicant should assume a fission product release\1\ from the core, the expected demonstrable leak rate from the containment and the meteorological conditions pertinent to his site to derive an exclusion area, a low population zone and population center distance. For the purpose of this analysis, which shall set forth the basis for the numerical values used, the applicant should determine the following: --------------------------------------------------------------------------- \1\The fission product release assumed for these calculations should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidential events, that would result in potential hazards not exceeded by those from any accident considered credible. Such accidents have generally been assumed to result in substantial meltdown of the core with subsequent release of appreciable quantities of fission products. --------------------------------------------------------------------------- (1) An exclusion area of such size that an individual located at any point on its boundary for two hours immediately following onset of the postulated fission product release would not receive a total radiation dose to the whole body in excess of 25 rem\2\ or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure. --------------------------------------------------------------------------- \2\The whole body dose of 25 rem referred to above corresponds numerically to the once in a lifetime accidental or emergency dose for radiation workers which, according to NCRP recommendations may be disregarded in the determination of their radiation exposure status (see NBS Handbook 69 dated June 5, 1959). However, neither its use nor that of the 300 rem value for thyroid exposure as set forth in these site criteria guides are intended to imply that these numbers constitute acceptable limits for emergency doses to the public under accident conditions. Rather, this 25 rem whole body value and the 300 rem thyroid value have been set forth in these guides as reference values, which can be used in the evaluation of reactor sites with respect to potential reactor accidents of exceedingly low probability of occurrence, and low risk of public exposure to radiation. --------------------------------------------------------------------------- (2) A low population zone of such size that an individual located at any point on its outer boundary who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure. (3) A population center distance of at least one and one''third times the distance from the reactor to the outer boundary of the low population zone. In applying this guide, the boundary of the population center shall be determined upon consideration of population distribution. Political boundaries are not controlling in the application of this guide. Where very large cities are involved, a greater distance may be necessary because of total integrated population dose consideration. (b) For sites for multiple reactor facilities consideration should be given to the following: (1) If the reactors are independent to the extent that an accident in one reactor would not initiate an accident in another, the size of the exclusion area, low population zone and population center distance shall be fulfilled with respect to each reactor individually. The envelopes of the plan overlay of the areas so calculated shall then be taken as their respective boundaries. (2) If the reactors are interconnected to the extent that an accident in one reactor could affect the safety of operation of any other, the size of the exclusion area, low population zone and population center distance shall be based upon the assumption that all interconnected reactors emit their postulated fission product releases simultaneously. This requirement may be reduced in relation to the degree of coupling between reactors, the probability of concomitant accidents and the probability that an individual would not be exposed to the radiation effects from simultaneous releases. The applicant would be expected to justify to the satisfaction of the Commission the basis for such a reduction in the source term. (3) The applicant is expected to show that the simultaneous operation of multiple reactors at a site will not result in total radioactive effluent releases beyond the allowable limits of applicable regulations. Note: For further guidance in developing the exclusion area, the low population zone, and the population center distance, reference is made to Technical Information Document 14844, dated March 23, 1962, which contains a procedural method and a sample calculation that result in distances roughly reflecting current siting practices of the Commission. The calculations described in Technical Information Document 14844 may be used as a point of departure for consideration of particular site requirements which may result from evaluation of the characteristics of a particular reactor, its purpose and method of operation. Copies of Technical Information Document 14844 may be obtained from the Commission's Public Document Room, 2120 L Street NW.(Lower Level), Washington, DC, or by writing the Director of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555. 17. through 19. Subpart B (Secs. 100.20-100.23) is added to read as follows: Subpart B--Evaluation Factors for Stationary Power Reactor Site Applications on or After [EFFECTIVE DATE OF THE FINAL RULE] Sec. 100.20 Factors to be considered when evaluating sites. 100.21 Non-seismic siting criteria. 100.23 Geologic and seismic siting factors. Sec. 100.20 Factors to be considered when evaluating sites. The Commission will take the following factors into consideration in determining the acceptability of a site for a stationary power reactor: (a) Population density and use characteristics of the site environs, including the exclusion area, the population distribution, and site-related characteristics must be evaluated to determine whether individual as well as societal risk of potential plant accidents is low, and that site-related characteristics would not prevent the development of a plan to carry out suitable protective actions for members of the public in the event of emergency. (b) The nature and proximity of man-related hazards (e.g., airports, dams, transportation routes, military and chemical facilities) must be evaluated to establish site parameters for use in determining whether a plant design can accommodate commonly occurring hazards, and whether the risk of other hazards is very low. (c) Physical characteristics of the site, including seismology, meteorology, geology, and hydrology. (1) Sec. 100.23, ``Geologic and seismic siting factors,'' of this part describes the criteria and nature of investigations required to obtain the geologic and seismic data necessary to determine the suitability of the proposed site and the plant design bases. (2) Meteorological characteristics of the site that are necessary for safety analysis or that may have an impact upon plant design (such as maximum probable wind speed and precipitation) must be identified and characterized. (3) Factors important to hydrological radionuclide transport such as soil, sediment, and rock characteristics, adsorption and retention coefficients, ground water velocity, and distances to the nearest surface body of water) must be obtained from on-site measurements. The maximum probable flood along with the potential for seismically induced floods discussed in Sec. 100.23(d)(3) of this part must be estimated using historical data. Sec. 100.21 Non-seismic siting criteria. Applications for site approval for commercial power reactors shall demonstrate that the proposed site meets the following criteria: (a) Every site must have an exclusion area and a low population zone, as defined in Sec. 100.3; (b) The population center distance, as defined in Sec. 100.3, must be at least one and one-third times the distance from the reactor to the outer boundary of (the low population zone. In applying this guide, the boundary of the population center shall be determined upon consideration of population distribution. Political boundaries are not controlling in the application of this guide; (c) Site atmospheric dispersion characteristics must be evaluated and dispersion parameters established such that: (1) Radiological effluent release limits associated with normal operation from the type of facility proposed to be located at the site can be met for any individual located offsite; and (2) Radiological dose consequences of postulated accidents shall meet the criteria set forth in Sec. 50.34(a)(1) of this chapter for the type of facility proposed to be located at the site; (d) The physical characteristics of the site, including meteorology, geology, seismology, and hydrology must be evaluated and site parameters established such that potential threats from such physical characteristics will pose no undue risk to the type of facility proposed to be located at the site; (e) Potential hazards associated with nearby transportation routes, industrial and military facilities must be evaluated and site parameters established such that potential hazards from such routes and facilities will pose no undue risk to the type of facility proposed to be located at the site; (f) Site characteristics must be such that adequate security plans and measures can be developed; (g) Site characteristics must be such that adequate plans to take protective actions for members of the public in the event of emergency can be developed: (h) Reactor sites should be located away from very densely populated centers. Areas of low population density are, generally, preferred. However, in determining the acceptability of a particular site located away from a very densely populated center but not in an area of low density, consideration will be given to safety, environmental, economic, or other factors, which may result in the site being found acceptable.\3\ --------------------------------------------------------------------------- \3\Examples of these factors include, but are not limited to, such factors as the higher population density site having superior seismic characteristics, better access to skilled labor for construction, better rail and highway access, shorter transmission line requirements, or less environmental impact on undeveloped areas, wetlands or endangered species, etc. Some of these factors are included in, or impact, the other criteria included in this section. --------------------------------------------------------------------------- Sec. 100.23 Geologic and seismic siting factors. This section sets forth the principal geologic and seismic considerations that guide the Commission in its evaluation of the suitability of a proposed site and adequacy of the design bases established in consideration of the geologic and seismic characteristics of the proposed site, such that, there is a reasonable assurance that a nuclear power plant can be constructed and operated at the proposed site without undue risk to the health and safety of the public. Applications to engineering design are contained in appendix S to part 50 of this chapter. (a) Applicability. The requirements in paragraphs (c) and (d) of this section apply to applicants for an early site permit or combined license pursuant to part 52 of this chapter, or a construction permit or operating license for a nuclear power plant pursuant to Part 50 of this chapter on or after [EFFECTIVE DATE OF THE FINAL RULE]. However, if the construction permit was issued prior to [EFFECTIVE DATE OF THE FINAL RULE], the operating license applicant shall comply with the seismic and geologic siting criteria in appendix A to part 100 of this chapter. (b) Commencement of construction. The investigations required in paragraph (c) of this section are within the scope of investigations permitted by Sec. 50.10(c)(1) of this chapter. (c) Geological, seismological, and engineering characteristics. The geological, seismological, and engineering characteristics of a site and its environs must be investigated in sufficient scope and detail to permit an adequate evaluation of the proposed site, to provide sufficient information to support evaluations performed to arrive at estimates of the Safe Shutdown Earthquake Ground Motion, and to permit adequate engineering solutions to actual or potential geologic and seismic effects at the proposed site. The size of the region to be investigated and the type of data pertinent to the investigations must be determined based on the nature of the region surrounding the proposed site. Data on the vibratory ground motion, tectonic surface deformation, nontectonic deformation, earthquake recurrence rates, fault geometry and slip rates, site foundation material, and seismically induced floods and water waves must be obtained by reviewing pertinent literature and carrying out field investigations. However, each applicant shall investigate all geologic and seismic factors (for example, volcanic activity) that may affect the design and operation of the proposed nuclear power plant irrespective of whether such factors are explicitly included in this section. (d) Geologic and seismic siting factors. The geologic and seismic siting factors considered for design must include a determination of the Safe Shutdown Earthquake Ground Motion for the site, the potential for surface tectonic and nontectonic deformations, the design bases for seismically induced floods and water waves, and other design conditions as stated in paragraph (d)(4) of this section. (1) Determination of the Safe Shutdown Earthquake Ground Motion. The Safe Shutdown Earthquake Ground Motion for the site is characterized by both horizontal and vertical free-field ground motion response spectra at the free ground surface. The Safe Shutdown Earthquake Ground Motion for the site is determined considering the results of the investigations required by paragraph (c) of this section. Uncertainties are inherent in such estimates. These uncertainties must be addressed through an appropriate analysis, such as a probabilistic seismic hazard analysis or suitable sensitivity analyses. Paragraph IV(a)(1) of appendix S to part 50 of this chapter defines the minimum Safe Shutdown Earthquake Ground Motion for design. (2) Determination of the potential for surface tectonic and nontectonic deformations. Sufficient geological, seismological, and geophysical data must be provided to clearly establish whether there is a potential for surface deformation. (3) Determination of design bases for seismically induced floods and water waves. The size of seismically induced floods and water waves that could affect a site from either locally or distantly generated seismic activity must be determined. (4) Determination of siting factors for other design conditions. Siting factors for other design conditions that must be evaluated include soil and rock stability, liquefaction potential, natural and artificial slope stability, cooling water supply, and remote safety- related structure siting. Each applicant shall evaluate all siting factors and potential causes of failure, such as, the physical properties of the materials underlying the site, ground disruption, and the effects of vibratory ground motion that may affect the design and operation of the proposed nuclear power plant. Dated at Rockville, Maryland, this 11th day of October. For the Nuclear Regulatory Commission. John C. Hoyle, Acting Secretary of the Commission. [FR Doc. 94-25585 Filed 10-14-94; 8:45 am] BILLING CODE 7590-01-P