[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