[Federal Register Volume 61, Number 239 (Wednesday, December 11, 1996)]
[Rules and Regulations]
[Pages 65157-65177]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-31075]


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NUCLEAR REGULATORY COMMISSION

10 CFR Parts 21, 50, 52, 54 and 100

RIN 3150-AD93


Reactor Site Criteria Including Seismic and Earthquake 
Engineering Criteria for Nuclear Power Plants

AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending 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 rule allows 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. 
This 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 also is 
denying the remaining issue in petition (PRM-50-20) filed by Free 
Environment, Inc. et al.

EFFECTIVE DATE: January 10, 1997.

FOR FURTHER INFORMATION CONTACT: Dr. Andrew J. Murphy, Office of 
Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, telephone (301) 415-6010, concerning the 
seismic and earthquake engineering aspects and Mr. Charles E. Ader, 
Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, telephone (301) 415-5622, 
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 Sections.
VII. Future Regulatory Action.
VIII. Referenced Documents.
IX. Summary of Comments on the Proposed Regulations.
    A. Reactor Siting Criteria (Nonseismic).
    B. Seismic and Earthquake Engineering Criteria.
X. Small Business Regulatory Enforcement Fairness Act
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

[[Page 65158]]

protection of the public health and safety. The Commission is 
addressing the remaining issue in PRM-50-20 as part of this rulemaking 
action.
    Appendix A, ``Seismic and Geologic Siting Criteria for Nuclear 
Power Plants,'' to 10 CFR Part 100 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; January 5, 
1993) 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; March 26, 
1993).
    The second proposed revision to these regulations was published for 
public comment on October 17, 1994 (59 FR 52255). The NRC stated on 
February 8, 1995, (60 FR 7467) that it intended to extend the comment 
period to allow interested persons adequate time to provide comments on 
staff guidance documents. On February 28, 1995, the availability of the 
five draft regulatory guides and three standard review plan sections 
that were developed to provide guidance on meeting the proposed 
regulations was published (60 FR 10880) and the comment period for the 
proposed rule was extended to May 12, 1995 (60 FR 10810).

II. Objectives

    The objectives of this 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 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.
    The new requirements in this rulemaking 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. However, for those operating 
license applicants and holders whose construction permits were issued 
prior to the effective date of this final regulation, the reactor site 
criteria in 10 CFR Part 100, and the seismic and geologic siting 
criteria and the earthquake engineering criteria in Appendix A to 10 
CFR Part 100 would continue to apply in all subsequent proceedings, 
including license amendments and renewal of operating licenses pursuant 
to 10 CFR Part 54.
    Criteria not associated with the selection of the site or 
establishment of the Safe Shutdown Earthquake Ground Motion (SSE) have 
been placed in 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 this final regulation 
does not apply to existing plants, the licensing bases for existing 
nuclear power plants must remain a part of the regulations. Therefore, 
the non-seismic and seismic reactor site criteria for current plants is 
retained as Subpart A and Appendix A to 10 CFR Part 100, respectively. 
The revised reactor site criteria is added as Subpart B in 10 CFR Part 
100 and applies to site applications received on or after the effective 
date of the final regulations. Non-seismic site criteria is added as a 
new Sec. 100.21 to Subpart B in 10 CFR Part 100. The criteria on 
seismic and geologic siting is added as a new Sec. 100.23 to Subpart B 
in 10 CFR Part 100. The dose calculations and the earthquake 
engineering criteria is 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 regulation also makes conforming 
amendments to 10 CFR Parts 21, 50, 52, and 54. Sections 21.3, 
50.49(b)(1), 50.65(b)(1), 52.17(a)(1), and 54.4(a)(1)(iii) are amended 
to reflect changes in Sec. 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

[[Page 65159]]

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 2000 
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, 
that site characteristics are such that adequate security measures to 
protect the plant can be developed, and that physical characteristics 
unique to the proposed site that could pose a significant impediment to 
the development of emergency plans are identified. The Commission has 
also had a long standing policy of siting reactors away from densely 
populated centers, and is continuing this policy in this rule.
    The Commission is incorporating basic reactor site criteria in this 
rule to accomplish the above purposes. The Commission is retaining 
source term and dose calculations to verify the adequacy of a site for 
a specific plant, but source term and dose calculations are 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 is 
referenced in Part 50. Rather, the source term is 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 can be utilized with the 
source term currently used for light-water reactors, or used in 
conjunction with revised accident source terms.
    The 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 will 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 will 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 retaining this requirement, 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 will not be permitted. For plants that do not involve 
standardized designs, the source term and dose calculations will 
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

[[Page 65160]]

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.
    The proposed rule considered two changes in this area.
    First, the Commission proposed 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 proposed use of the total effective dose equivalent, or TEDE, 
was noted as being consistent with Part 20 of the Commission's 
regulations and was also 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 considered 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. The 
Commission proposed 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 supplementary information in the proposed rule noted that 
the current dose criteria are equivalent in risk to 27 rem TEDE, the 
Commission proposed 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.
    However, the Commission specifically requested comments on whether 
the current dose criteria should be modified to utilize the total 
effective dose equivalent or TEDE concept, 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, and 
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.
    Based on the comments received, there was a general consensus that 
the use of the TEDE concept was appropriate, and a nearly unanimous 
opinion that no organ ``capping'' dose was required, since the TEDE 
concept provided the appropriate risk weighting for all body organs.
    With regard to the value to be used as the dose criterion, a number 
of comments were received that the proposed value of 25 rem TEDE 
represented a more restrictive criterion than the current values of 25 
rem whole body and 300 rem to the thyroid gland. These commenters noted 
that the use of organ weighting factors of 1 for the whole body and 
0.03 for the thyroid as given in 10 CFR Part 20, would yield a value of 
34 rem TEDE for whole body and thyroid doses of 25 and 300 rem, 
respectively. This is because the organ weighting factors in 10 CFR 
Part 20 include other effects (e.g., genetic) in addition to latent 
cancer fatality.
    After careful consideration, the Commission has decided to adopt a 
value of 25 rem TEDE as the dose acceptance criterion for the final 
rule. The bases for this decision follows. First, the Commission has 
generally based its regulations on the risk of latent cancer fatality. 
Although a numerical calculation would lead to a value of 27 rem TEDE, 
as noted in the discussion that accompanied the proposed rule, the 
Commission concludes that a value of 25 rem is sufficiently close, and 
that the use of 27 rather than 25 implies an unwarranted numerical 
precision. In addition, in terms of occupational dose, Part 20 also 
permits a once-in-a-lifetime planned special dose of 25 rem TEDE. In 
addition, EPA guidance sets a limit of 25 rem TEDE for workers 
performing emergency service such as lifesaving or protection of large 
populations. While the Commission does not, as noted above, regard this 
dose value as one that is acceptable for members of the public under 
accident conditions, it provides a useful perspective with regard to 
doses that ought not to be exceeded, even for radiation workers under 
emergency conditions.
    The argument that a criterion of 25 rem TEDE in conjunction with 
the organ weighting factors of 10 CFR Part 20 for its calculation 
represents a tightening of the dose criterion, while true in theory, is 
not true in practice. A review of the dose analyses for operating 
plants has shown that the thyroid dose limit of 300 rem has been the 
limiting dose criterion in licensing reviews, and that all operating 
plants would be able to meet a dose criterion of 25 rem TEDE. Hence, 
the Commission concludes that, in practice, use of the organ weighting 
factors of Part 20 together with a dose criterion of 25 rem TEDE, 
represents a relaxation rather than a tightening of the dose criterion. 
In adopting this value, the Commission also rejects the view, advanced 
by some, that the dose calculation is merely a ``reference'' value that 
bears no relation to what might be experienced by an actual person in 
an accident. Although the Commission considers it highly unlikely that 
an actual person would receive such a dose, because of the conservative 
and stylized assumptions employed in its calculation, it is 
conceivable.
    The second change proposed in this area was 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 proposed rule stated 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 proposed 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 an 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 proposed that 
various two-hour periods be examined to assure that the dose to a 
hypothetical individual at the exclusion area boundary would not be in 
excess of 25 rem TEDE over any two-hour period after the onset of 
fission product release.
    A number of comments received in regard to this proposed criterion 
stated that so-called ``sliding'' two-hour

[[Page 65161]]

window for dose evaluation at the exclusion area boundary was 
confusing, illogical, and inappropriate. Several commenters felt it was 
difficult to ascertain which two hour period represented the maximum. 
Others expressed the view that the significance of such a calculation 
was not clearly stated nor understood. For example, one comment 
expressed the view that a dose evaluated for a ``sliding'' two-hour 
period was logically inconsistent since it implied either that an 
individual was not at the exclusion area boundary prior to the 
accident, and approached close to the plant after initiation of the 
accident, contrary to what might be expected, or that the individual 
was, in fact, located at the exclusion area boundary all along, in 
which case the dose contribution received prior to the ``maximum'' two-
hour value was being ignored.
    Although the Commission recognizes that evaluation of the dose to a 
hypothetical individual over any two-hour period may not be entirely 
consistent with the actions of an actual individual in an accident, the 
intent is to assure that the short-term dose to an individual will not 
be in excess of the acceptable value, even where there is some 
variability in the time that an individual might be located at the 
exclusion area boundary. In addition, the dose calculation should not 
be taken too literally with regard to the actions of a real individual, 
but rather is intended primarily as a means to evaluate the 
effectiveness of the plant design and site characteristics in 
mitigating postulated accidents.
    For these reasons, the Commission is retaining the requirement, in 
the final rule, that the dose to an individual located at the nearest 
exclusion area boundary over any two-hour period after the appearance 
of fission products in containment, should not be in excess of 25 rem 
total effective dose equivalent (TEDE).
    (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 is requiring 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 requiring that the atmospheric dispersion characteristics 
of the site 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.
    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 retaining 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 final rule 
requires 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 final rule codifies this 
practice.
    (F) Adequacy of Security Plans. The rule requires 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 will entail a small secure area considerably smaller than 
that envisioned for the exclusion area.
    (G) Emergency Planning. The proposed rule stated that the site 
characteristics should be such that adequate plans to carry out 
protective measures for members of the public in the event of emergency 
could be developed. To avoid any misinterpretation that the Commission 
is adopting emergency planning standards that implicitly overrule or 
may be in conflict with previous Commission decisions (e.g., CLI-90-
02), the language in the final rule has been modified to be consistent 
with that of section 52.17 of the Commission's regulations regarding 
early site permits.
    The Commission's decision in Seabrook on emergency planning, made 
in connection with an operating license review for a site previously 
approved, is being extended in considering site suitability for future 
reactor sites. The Commission, in its Seabrook decision, CLI-90-02, 
reiterated its earlier determination in the Shoreham decision, CLI-86-
13, that the adequacy of an emergency plan is to be determined by the 
sixteen planning standards of 10 CFR 50.47(b), and that these standards 
do not require that an adequate plan achieve a preset minimum radiation 
dose saving or a minimum evacuation time for the plume exposure pathway 
emergency planning zone in the event of a serious accident. Rather, the 
Commission noted that emergency planning is required as a matter of 
prudence and for defense-in-depth, and that the adequacy of an 
emergency plan was to be judged on the basis of its meeting the 16 
planning standards given in 10 CFR 50.47(b). Hence, the characteristics 
of the site, which determine the evacuation time for the plume exposure 
pathway emergency planning zone, have not entered into the 
determination of the adequacy of an emergency plan. Emergency plans 
developed according to the above planning standards will result in 
reasonable assurance that adequate protective measures can be taken in 
the event of emergency.
    It is sufficient that an applicant identify any physical site 
characteristics that could represent a significant impediment to the 
development of emergency plans, primarily to assure that ``A range of 
protective actions have been developed for the plume exposure pathway 
emergency planning zone for

[[Page 65162]]

emergency workers and the public'', as stated in the planning 
standards.
    Accordingly, appropriate sections of the rule (e.g., 
Sec. 100.21(g)) have been modified to state that ``physical 
characteristics unique to the proposed site that could pose a 
significant impediment to the development of emergency plans must be 
identified.'' Except for the deletion of the phrase ``such as egress 
limitations from the area surrounding the site'', this language is 
identical to that in Sec. 52.17(b)(1). This phrase is being deleted 
from Sec. 100.21(g) (but Sec. 52.17(b)(1) remains unchanged), to 
eliminate any confusion that might arise regarding its scope.
    (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, the likelihood that radioactive material would actually 
be carried towards the city is reduced 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 concludes 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 incorporating a two-tier approach with regard to 
population density and reactor sites. The rule requires that reactor 
sites 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, can 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

[[Page 65163]]

the Commission if they present advantages in terms of other 
considerations applicable to the evaluation of proposed sites.
Petition Filed By Free Environment, Inc. et al.
    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 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 (32 km) from the reactor as a result of land 
contamination from a severe accident is very low. A revision of 
Regulatory Guide 4.7 which incorporated this finding that population 
density guidance beyond 20 miles was not needed in the evaluation of 
potential reactor sites was issued for comment at the time of the 
proposed rule. No comments were received on this aspect of the guide.
    Therefore, the Commission concludes that the 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. Population growth in the site vicinity will be periodically 
factored into the emergency plan for the site, but 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 change to 10 CFR Part 50 relocates 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 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 to Appendix A, ``Seismic and Geologic 
Siting Criteria for Nuclear Power Plants,'' to 10 CFR Part 100, are 
associated with the seismic and earthquake engineering criteria 
rulemaking. These changes reflect new information and research results, 
and incorporate the intentions of this regulatory action as defined in 
Section III of this rule. Much of the following discussion remains 
unchanged from that issued for public comment (59 FR 52255) because 
there were no comments which necessitated a major change to the 
regulations and supporting documentation.
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 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
    Appendix A to 10 CFR Part 100 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.

[[Page 65164]]

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 final regulation is 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 final regulation is reduced 
considerably. Thus, the final regulation contains: (a) required 
definitions, (b) a requirement to determine the geological, 
seismological, and engineering characteristics of the proposed site, 
and (c) requirements to determine the Safe Shutdown Earthquake Ground 
Motion (SSE), 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 approach to determine the SSE 
are presented in the following section. The elements are the guidance 
that is described in Regulatory Guide 1.165, ``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 of 
uncertainties 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 not only to 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 final regulation 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.
    Methods acceptable to the NRC staff for implementing the regulation 
are described in Regulatory Guide 1.165, ``Identification and 
Characterization of Seismic Sources and Determination of Safe Shutdown 
Earthquake Ground Motion.'' The key elements of this 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 the regional and site 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 approach requires thorough regional and site-specific 
geoscience investigations. Results of the regional and site-specific 
investigations must be considered in applications 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 
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 as 
necessary.
    It is also necessary to incorporate local site geological factors 
such as structural geology, 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 
characteristics are determined using the procedures outlined in 
Standard Review Plan Section 2.5.2, ``Vibratory Ground Motion,'' 
Revision 3.
    The NRC staff's review approach to evaluate ground motion estimates 
is described in SRP Section 2.5.2, Revision 3. This review takes into 
account the information base developed in licensing more than 100 
plants. 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

[[Page 65165]]

consistency with previous licensing decisions and that the scientific 
bases for decisions are clearly understood. This review approach will 
also assess the fairly complex regional probabilistic modeling, which 
incorporates multiple hypotheses and a multitude of parameters. 
Furthermore, the NRC staff's Safety Evaluation Report 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 final regulation moves 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 final 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. The final regulation also 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 Part 100, 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 Part 100, 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 a compromise is required between design for a broad 
spectrum of unlikely events and optimum design for normal operation. 
Design for a single limiting event (the SSE) and inspection and 
evaluation for earthquakes in excess of some specified limit (the OBE), 
when and if they occur, 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. Furthermore, the final safety evaluation reports 
related to the certification of the System 80+ and the Advanced Boiling 
Water Reactor design (NUREG-1462 and NUREG-1503, respectively) have 
already adopted the single earthquake design philosophy.
    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 final regulation allows 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, it is determined that if an OBE of one-third or less 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 
the 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

[[Page 65166]]

response would be used in the analyses. Section VII of this final 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 NRC staff clarified that guidelines 
should be maintained to ensure the functionality of components, 
equipment, and their supports. In addition, the NRC 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 Shutdown
    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 will be 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 OBE 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 final regulation treats 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) is 
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 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.
    Regulatory Guide 1.166, ``Pre-Earthquake Planning and Immediate 
Nuclear Power Plant Operator Post-Earthquake Actions,'' provides 
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 by the licensee 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 in Regulatory Guide 1.166 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.
    Regulatory Guide 1.167, ``Restart of a Nuclear Power Plant Shut 
Down by a Seismic Event,'' provides 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
    Section 100.23 resolves questions of interpretation. As an example, 
definitions and required investigations stated in the final regulation 
do not contain the phrases in Appendix A to Part 100 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 relocation of the earthquake engineering 
criteria to Appendix S to Part 50 and the 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 references to both the Part 100 and Part 50 dose 
guidelines.

VI. Related Regulatory Guides and Standard Review Plan Sections

    The NRC is developing the following regulatory guides and standard 
review plan sections to provide prospective licensees with the 
necessary guidance for implementing the final regulation. The notice of 
availability for these materials will be published in a later issue of 
the Federal Register.
    1. Regulatory Guide 1.165, ``Identification and Characterization of 
Seismic Sources and Determination of

[[Page 65167]]

Shutdown Earthquake Ground Motions.'' The 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 rule describes the key elements.
    2. Regulatory Guide 1.12, Revision 2, ``Nuclear Power Plant 
Instrumentation for Earthquakes.'' The guide describes seismic 
instrumentation type and location, operability, characteristics, 
installation, actuation, and maintenance that are acceptable to the NRC 
staff.
    3. Regulatory Guide 1.166, ``Pre-Earthquake Planning and Immediate 
Nuclear Power Plant Operator Post-Earthquake Actions.'' The 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. Regulatory Guide 1.167, ``Restart of a Nuclear Power Plant Shut 
Down by a Seismic Event.'' The 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. Standard Review Plan Section 2.5.1, Revision 3, ``Basic Geologic 
and Seismic Information.'' This SRP Section 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. Standard Review Plan Section 2.5.2, Revision 3 ``Vibratory 
Ground Motion.'' This SRP Section describes procedures to assess the 
ground motion potential of seismic sources at the site and to assess 
the adequacy of the SSE.
    7. Standard Review Plan Section 2.5.3, Revision 3, ``Surface 
Faulting.'' This SRP Section 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. Regulatory Guide 4.7, Revision 2, ``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 as final guides without public 
comment subsequent to the publication of the final regulations.
    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. 1.3, ``Assumptions Used for Evaluating the Potential 
Radiological Consequences of a Loss of Coolant Accident for Boiling 
Water Reactors.''
    2. 1.4, ``Assumptions Used for Evaluating the Potential 
Radiological Consequences of a Loss of Coolant Accident for Pressurized 
Water Reactors.''
    3. 1.27, ``Ultimate Heat Sink for Nuclear Power Plants.''
    4. 1.100, ``Seismic Qualification of Electric and Mechanical 
Equipment for Nuclear Power Plants.''
    5. 1.124, ``Service Limits and Loading Combinations for Class 1 
Linear-Type Component Supports.''
    6. 1.130, ``Service Limits and Loading Combinations for Class 1 
Plate-and-Shell-Type Component Supports.''
    7. 1.132, ``Site Investigations for Foundations of Nuclear Power 
Plants.''
    8. 1.138, ``Laboratory Investigations of Soils for Engineering 
Analysis and Design of Nuclear Power Plants.''
    9. 1.142, ``Safety-Related Concrete Structures for Nuclear Power 
Plants (Other than Reactor Vessels and Containments).''
    10. 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 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 rule as set out below.
    Copies of NUREG-0625, NUREG-1061, NUREG-1150, NUREG-1451, NUREG-
1462, NUREG-1503, 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 also are available from the 
National Technical Information Service, 5285 Port Royal Road, 
Springfield, VA 22161. A copy also is 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, P.O. Box 37082, 
Washington, DC 20402-9328. Issued guides also may 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 NRC Public Document Room, 
2120 L Street, NW. (Lower Level), Washington, DC.

IX. Summary of Comments on the Proposed Regulations

A. Reactor Siting Criteria (Nonseismic)

    Eight organizations or individuals commented on the nonseismic 
aspects of the second proposed revision. The first proposed revision 
issued for comment in October 20, 1992, (57 FR 47802) elicited strong 
comments in regard to proposed numerical values of population density 
and a minimum distance to the exclusion area boundary (EAB) in the 
rule. The second proposed revision (October 17, 1994; 59 FR 52255) 
would delete these from the rule by providing guidance on population 
density in a Regulatory Guide and determining the distance to the EAB 
and LPZ by use of source term and dose

[[Page 65168]]

calculations. The rule would contain basic site criteria, without any 
numerical values.
    Several commentors representing the nuclear industry and 
international nuclear organizations stated that the second proposed 
revision was a significant improvement over the first proposed 
revision, while the only public interest group commented that the NRC 
had retreated from decoupling siting and design in response to the 
comments of foreign entities.
    Most comments on the second proposed revision centered on the use 
of total effective dose equivalent (TEDE), the proposed single 
numerical dose acceptance criterion of 25 rem TEDE, the evaluation of 
the maximum dose in any two-hour period, and the question of whether an 
organ capping dose should be adopted.
    Virtually all commenters supported the concept of TEDE and its use. 
However, there were differing views on the proposed numerical dose of 
25 rem and the proposed use of the maximum two-hour period to evaluate 
the dose. Virtually all industry commenters felt that the proposed 
numerical value of 25 rem TEDE was too low and that it represented a 
``ratchet'' since the use of the current dose criteria plus organ 
weighting factors would suggest a value of 34 rem TEDE. In addition, 
all industry commenters believed the ``sliding'' two-hour window for 
dose evaluation to be confusing, illogical and inappropriate. They 
favored a rule that was based upon a two hour period after the onset of 
fission product release, similar in concept to the existing rule. All 
industry commenters opposed the use of an organ capping dose. The only 
public interest group that commented did not object to the use of TEDE, 
favored the proposed dose value of 25 rem, and supported an organ 
capping dose.

B. Seismic and Earthquake Engineering Criteria

    Seven letters were received addressing either the regulations or 
both the regulations and the draft guidance documents identified in 
Section VI (except DG-4003). An additional five letters were received 
addressing only the guidance documents, for a total of twelve comment 
letters. A document, ``Resolution of Public Comments on the Proposed 
Seismic and Earthquake Engineering Criteria for Nuclear Power Plants,'' 
is available explaining the NRC's disposition of the comments received 
on the regulations. A copy of this document has been placed in the NRC 
Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC. 
Single copies are available from Dr. Andrew J. Murphy, Office of 
Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, telephone (301) 415-6010. A second document, 
``Resolution of Public Comments on Draft Regulatory Guides and Standard 
Review Plan Sections Pertaining to the Proposed Seismic and Earthquake 
Engineering Criteria for Nuclear Power Plants,'' will explain the NRC's 
disposition of the comments received on the guidance documents. The 
Federal Register notice announcing the avaliability of the guidance 
documents will also discuss how to obtain copies of the comment 
resolution document.
    A summary of the major comments on the proposed regulations 
follows:
Section III, Genesis (Application)
    Comment: The Department of Energy (Office of Civilian Radioactive 
Waste Management), requests an explicit statement on whether or not 
Sec. 100.23 applies to the Mined Geologic Disposal System (MGDS) and a 
Monitored Retrievable Storage (MRS) facility. The NRC has noted in 
NUREG-1451, ``Staff Technical Position on Investigations to Identify 
Fault Displacement Hazards and Seismic Hazards at a Geologic 
Respository,'' that Appendix A to 10 CFR Part 100 does not apply to a 
geologic repository. NUREG-1451 also notes that the contemplated 
revisions to Part 100 would also not be applicable to a geologic 
repository. Section 72.102(b) requires that, for an MRS located west of 
the Rocky Mountain front or in areas of known potential seismic 
activity in the east, the seismicity be evaluated by the techniques of 
Appendix A to 10 CFR Part 100.
    Response: Although Appendix A to 10 CFR Part 100 is titled 
``Seismic and Geologic Siting Criteria for Nuclear Power Plants,'' it 
is also referenced in two other parts of the regulation. They are (1) 
Part 40, ``Domestic Licensing of Source Material,'' Appendix A, 
``Criteria Relating to the Operation of Uranium Mills and the 
Disposition of Tailings or Waste Produced by the Extraction or 
Concentration of Source Material from Ores Processed Primarily for 
Their Source Material Content,'' Section I, Criterion 4(e), and (2) 
Part 72, ``Licensing Requirements for the Independent Storage of Spent 
Nuclear Fuel and High-Level Radioactive Waste,'' Paragraphs (a)(2), (b) 
and (f)(1) of Sec. 72.102.
    The referenced applicability of Sec. 100.23 to other than power 
reactors, if considered appropriate by the NRC, would be a separate 
rulemaking. That rulemaking would clearly state the applicability of 
Sec. 100.23 to an MRS or other facility. In addition, NUREG-1451 will 
remain the NRC staff technical position on seismic siting issues 
pertaining to an MGDS until it is superseded through a rulemaking, 
revision of NUREG-1451, or other appropriate mechanism.
Section V(B)(5), ``Value of the Operating Basis Earthquake Ground 
Motion (OBE) and Required OBE Analysis.''
    Comment: One commenter, ABB Combustion Engineering Nuclear Systems, 
specifically stated that they agree with the NRC's proposal to not 
require explicit design analysis of the OBE if its peak acceleration is 
less than one-third of the Safe Shutdown Earthquake Ground Motion 
(SSE). The only negative comments, from G.C. Slagis Associates, stated 
that the proposed rule in the area of required OBE analysis is not 
sound, not technically justified, and not appropriate for the design of 
pressure-retaining components. The following are specific comments 
(limited to the design of pressure-retaining components to the ASME 
Boiler and Pressure Vessel Section III rules) that pertain to the 
supplemental information to the proposed regulations, item V(B)(5), 
``Value of the Operating Basis Earthquake Ground Motion (OBE) and 
Required OBE Analysis.''
    (1) Comment: Disagrees with the statement in SECY-79-300 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. It is not feasible to inspect for cyclic damage to 
all the pressure-retaining components. Visually inspecting for 
permanent deformation, or leakage, or failed component supports is 
certainly not adequate to determine cyclic damage.
    Response: The NRC agrees. Postearthquake inspection and evaluation 
guidance is described in Regulatory Guide 1.167 (Draft was DG-1035), 
``Restart of a Nuclear Power Plant Shut Down by an Seismic Event.'' The 
guidance is not limited to visual inspections; it includes inspections, 
tests, and analyses including fatigue analysis.
    (2) Comment: Disagrees with the NRC statement in SECY-090-016 that 
the OBE should not control design. There is a problem with the present 
requirements. Requiring design for five OBE events at one-half SSE is 
unrealistic for most (all?) sites and requires an excessive and 
unnecessary number of seismic supports. The solution is to properly 
define the OBE

[[Page 65169]]

magnitude and the number of events expected during the life of the 
plant and to require design for that loading. OBE may or may not 
control the design. But you cannot assume, before you have the 
seismicity defined and before you have a component design, that OBE 
will not govern the design.
    Response: The NRC has concluded that design requirements based on 
an estimated OBE magnitude at the plant site and the number of events 
expected during the plant life will lead to low design values that will 
not control the design, thus resulting in unnecessary analyses.
    (3) Comment: It is not technically justified to assume that Section 
III components will remain within applicable stress limits (Level B 
limits) at one-third the SSE. The Section III acceptance criteria for 
Level D (for an SSE) is completely different than that for Level B (for 
an OBE). The Level D criteria is based on surviving the extremely-low 
probability SSE load. Gross structural deformations are possible, and 
it is expected that the component will have to be replaced. Cyclic 
effects are not considered. The cyclic effects of the repeated 
earthquakes have to be considered in the design of the component to 
ensure pressure boundary integrity throughout the life of the 
component, especially if the SSE can occur after the lower level 
earthquakes.
    Response: In SECY-93-087, Issue I.M, ``Elimination of Operating-
Basis Earthquake,'' the NRC recognizes that a designer of piping 
systems considers the effects of primary and secondary stresses and 
evaluates fatigue caused by repeated cycles of loading. Primary 
stresses are induced by the inertial effects of vibratory motion. The 
relative motion of anchor points induces secondary stresses. The 
repeating seismic stress cycles induce cyclic effects (fatigue). 
However, after reviewing these aspects, the NRC concludes that, for 
primary stresses, if the OBE is established at one-third the SSE, the 
SSE load combinations control the piping design when the earthquake 
contribution dominates the load combination. Therefore, the NRC 
concludes that eliminating the OBE piping stress load combination for 
primary stresses in piping systems will not significantly reduce 
existing safety margins.
    Eliminating the OBE will, however, directly affect the current 
methods used to evaluate the adequacy of cyclic and secondary stress 
effects in the piping design. Eliminating the OBE from the load 
combination could cause uncertainty in evaluating the cyclic (fatigue) 
effects of earthquake-induced motions in piping systems and the 
relative motion effects of piping anchored to equipment and structures 
at various elevations because both of these effects are currently 
evaluated only for OBE loadings. Accordingly, to account for earthquake 
cycles in the fatigue analysis of piping systems, the staff proposes to 
develop guidelines for selecting a number of SSE cycles at a fraction 
of the peak amplitude of the SSE. These guidelines will provide a level 
of fatigue design for the piping equivalent to that currently provided 
in Standard Review Plan Section 3.9.2.
    Positions pertaining to the elimination of the OBE were proposed in 
SECY-93-087. Commission approval is documented in a memorandum from 
Samuel J. Chilk to James M. Taylor, Subject: SECY-93-087--Policy, 
Technical and Licensing Issues Pertaining to Evolutionary and Advanced 
Light-Water Reactor (ALWR) Designs, dated July 21, 1993.
    (4) Comment: There is one major flaw in the ``SSE only'' design 
approach. The equipment designed for SSE is limited to the equipment 
necessary to assure the integrity of the reactor coolant pressure 
boundary, to shutdown the reactor, and to prevent or mitigate accident 
consequences. The equipment designed for SSE is only part of the 
equipment ``necessary for continued operation without undue risk to the 
health and safety of the public.'' Hence, by this rule, it is possible 
that some equipment necessary for continued operation will not be 
designed for SSE or OBE effects.
    Response: The NRC does not agree that the design approach is 
flawed. It is not possible that some equipment necessary for continued 
safe operation will not be designed for SSE or OBE effects. General 
Design Criterion 2, ``Design Bases for Protection Against Natural 
Phenomena,'' of Appendix A, ``General Design Criteria for Nuclear Power 
Plants,'' to 10 CFR Part 50 requires that nuclear power plant 
structures, systems, and components important to safety be designed to 
withstand the effects of earthquakes without loss of capability to 
perform their safety functions. The criteria in Appendix S to 10 CFR 
Part 50 implement General Design Criterion 2 insofar as it requires 
structures, systems, and components important to safety to withstand 
the effects of earthquakes. Regulatory Guide 1.29, ``Seismic Design 
Classification,'' describes a method acceptable to the NRC for 
identifying and classifying those features of light-water-cooled 
nuclear power plants that should be designed to withstand the effects 
of the SSE. Currently, components which are designed for OBE only 
include components such as waste holdup tanks. As noted in Section VII, 
Future Regulatory Actions, regulatory guides related to these 
components will be revised to provide alternative design requirements.
10 CFR 100.23
    The Nuclear Energy Institute (NEI) congratulated the NRC staff for 
carefully considering and responding to the voluminous and complex 
comments that were provided on the earlier proposed rulemaking package 
(October 20, 1992; 57 FR 47802) and considered that the seismic portion 
of the proposed rulemaking package is nearing maturity and with the 
inclusion of industry's comments (which were principally on the 
guidance documents), has the potential to satisfy the objectives of 
predictable licensing and stable regulations.
    Both NEI and Westinghouse Electric Corporation support the 
regulation format, that is, prescriptive guidance is located in 
regulatory guides or standard review plan sections and not the 
regulation.
    NEI and Westinghouse Electric Corporation support the removal of 
the requirement from the first proposed rulemaking (57 FR 47802) that 
both deterministic and probabilistic evaluations must be conducted to 
determine site suitability and seismic design requirements for the 
site. [Note: the commenters do not agree with the NRC staff's 
deterministic check of the seismic sources and parameters used in the 
LLNL and EPRI probabilistic seismic hazard analyses (Regulatory Guide 
1.165, draft was DG-1032). Also, they do not support the NRC staff's 
deterministic check of the applicants submittal (SRP Section 2.5.2). 
These items are addressed in the document pertaining to comment 
resolution of the draft regulatory guides and standard review plan 
sections.]
    Comment: NEI, Westinghouse Electric Corporation, and Yankee Atomic 
Electric Corporation recommend that the regulation should state that 
for existing sites east of the Rocky Mountain Front (east of 
approximately 105 deg. west longitude), a 0.3g standardized design 
level is acceptable at these sites given confirmatory foundations 
evaluations [Regulatory Guide 1.132, but not the geologic, geophysical, 
seismological investigations in Regulatory Guide 1.165].
    Response: The NRC has determined that the use of a spectral shape 
anchored to 0.3g peak ground acceleration as a standardized design 
level would be

[[Page 65170]]

appropriate for existing central and eastern U.S. sites based on the 
current state of knowledge. However, as new information becomes 
available it may not be appropriate for future licensing decisions. 
Pertinent information such as that described in Regulatory Guide 1.165 
(Draft was DG-1032) is needed to make that assessment. Therefore, it is 
not appropriate to codify the request.
    Comment: NEI recommended a rewording of Paragraph (a), 
Applicability. Although unlikely, an applicant for an operating license 
already holding a construction permit may elect to apply the amended 
methodology and criteria in Subpart B to Part 100.
    Response: The NRC will address this request on a case-by-case basis 
rather than through a generic change to the regulations. This situation 
pertains to a limited number of facilities in various stages of 
construction. Some of the issues that must be addressed by the 
applicant and NRC during the operating license review include 
differences between the design bases derived from the current and 
amended regulations (Appendix A to Part 100 and Sec. 100.23, 
respectively), and earthquake engineering criteria such as, OBE design 
requirements and OBE shutdown requirements.
Appendix S to 10 CFR Part 50
    Support for the NRC position pertaining to the elimination of the 
Operating Basis Earthquake Ground Motion (OBE) response analyses has 
been documented in various NRC publications such as SECY-79-300, SECY-
90-016, SECY-93-087, and NUREG-1061. The final safety evaluation 
reports related to the certification of the System 80+ and the Advanced 
Boiling Water Reactor design (NUREG-1462 and NUREG-1503, respectively) 
have already adopted the single earthquake design philosophy. In 
addition, similar activities are being done in foreign countries, for 
instance, Germany. (Additional discussion is provided in Section 
V(B)(5) of this rule).
    Comment: The American Society of Civil Engineers (ASCE) recommended 
that the seismic design and engineering criteria of ASCE Standard 4, 
``Seismic Analysis of Safety-Related Nuclear Structures and Commentary 
on Standard for Seismic Analysis of Safety-Related Nuclear 
Structures,'' be incorporated by reference into Appendix S to 10 CFR 
Part 50.
    Response: The Commission has determined that new regulations will 
be more streamlined and contain only basic requirements with guidance 
being provided in regulatory guides and, to some extent, in standard 
review plan sections. Both the NRC and industry have experienced 
difficulties in applying prescriptive regulations such as Appendix A to 
10 CFR Part 100 because they inhibit the use of needed latitude in 
judgment. Therefore, it is common NRC practice not to reference 
publications such as ASCE Standard 4 (an analysis, not design standard) 
in its regulations. Rather, publications such as ASCE Standard 4 are 
cited in regulatory guides and standard review plan sections. ASCE 
Standard 4 is cited in the 1989 revision of Standard Review Plan 
Sections 3.7.1, 3.7.2, and 3.7.3.
    Comment: The Department of Energy stated that the required 
consideration of aftershocks in Paragraph IV(B), Surface Deformation, 
is confusing and recommended that it be deleted.
    Response: The NRC agrees. The reference to aftershocks in Paragraph 
IV(b) has been deleted. Paragraphs VI(a), Safe Shutdown Earthquake, and 
VI(B)(3) of Appendix A to Part 100 contain the phrase ``including 
aftershocks.'' The ``including aftershocks'' phrase was removed from 
the Safe Shutdown Earthquake Ground Motion requirements in the proposed 
regulation. The recommended change will make Paragraphs IV(a)(1), 
``Safe Shutdown Earthquake Ground Motion,'' and IV(b), ``Surface 
Deformation, of Appendix S to 10 CFR Part 50 consistent.

X. Small Business Regulatory Enforcement Fairness Act

    In accordance with the Small Business Regulatory Enforcement 
Fairness Act of 1996 the NRC has determined that this action is not a 
major rule and has verified this determination with the Office of 
Information and Regulatory Affairs of OMB.

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 regulation is not 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 public 
health and safety over the current regulations. The 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 baseline inspections of 
structures, equipment, and piping, and with maintenance of seismic 
instrumentation. Baseline 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 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 Dr. Andrew J. 
Murphy, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, telephone (301) 415-6010.

XII. Paperwork Reduction Act Statement

    This final rule amends information collection requirements that are 
subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et 
seq.). These requirements were approved by the Office of Management and 
Budget, approval numbers 3150-0011 and 3150-0093.
    The public reporting burden for this collection of information 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 on any aspect of this 
collection of information, including

[[Page 65171]]

suggestions for reducing the burden, to the Information and Records 
Management Branch (T-6 F33), U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, or by Internet electronic mail to 
[email protected]; 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.

Public Protection Notification

    The NRC may not conduct or sponsor, and a person is not required to 
respond to, a collection of information unless it displays a currently 
valid OMB control number.

XIII. Regulatory Analysis

    The Commission has prepared a regulatory analysis on this 
regulation. The analysis examines the costs and benefits of the 
alternatives considered by the Commission. Interested persons may 
examine a copy of the regulatory analysis at the NRC Public Document 
Room, 2120 L Street NW. (Lower Level), Washington, DC. Single copies of 
the analysis are available from Dr. Andrew J. Murphy, Office of Nuclear 
Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, telephone (301) 415-6010.

XIV. Regulatory Flexibility Certification

    As required by the Regulatory Flexibility Act of 1980, 5 U.S.C. 
605(b), the Commission certifies that this regulation does not have a 
significant economic impact on a substantial number of small entities. 
This regulation affects only the licensing and operation of nuclear 
power plants. The companies that own these plants do not fall within 
the definition of ``small entities'' set forth in the Regulatory 
Flexibility Act or the size standards established by the NRC (April 11, 
1995; 60 FR 18344).

XV. Backfit Analysis

    The NRC has determined that the backfit rule, 10 CFR 50.109, does 
not apply to this regulation, and, therefore, a backfit analysis is not 
required for this regulation because these amendments do not involve 
any provisions that would impose backfits as defined in 10 CFR 
50.109(a)(1). The 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 21

    Nuclear power plants and reactors, Penalties, Radiation protection, 
Reporting and recordkeeping requirements.

10 CFR Part 50

    Antitrust, Classified information, Criminal penalties, 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 54

    Administrative practice and procedure, Age-related degradation, 
Backfitting, Classified information, Criminal penalties, Environmental, 
Nuclear power plants and reactors, Reporting and recordkeeping 
requirements.

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. 552 and 553, the NRC is adopting 
the following amendments to 10 CFR Parts 21, 50, 52, 54, and 100:

PART 21--REPORTING OF DEFECTS AND NONCOMPLIANCE

    1. The authority citation for Part 21 continues to read as follows:

    Authority: Sec. 161, 68 Stat. 948, as amended, sec. 234, 83 
Stat. 444, as amended, sec. 1701, 106 Stat. 2951, 2953 (42 U.S.C. 
2201, 2282, 2297f); secs. 201, as amended, 206, 88 Stat. 1242, as 
amended, 1246 (42 U.S.C. 5841, 5846).
    Section 21.2 also issued under secs. 135, 141, Pub. L. 97-425, 
96 Stat. 2232, 2241 (42 U.S.C. 10155, 10161).

    2. In Sec. 21.3, the definition for Basic component (1)(i)(C) is 
revised to read as follows:


Sec. 21.3  Definitions.

* * * * *
    Basic component. (1)(i) * * *
    (C) The capability to prevent or mitigate the consequences of 
accidents which could result in potential offsite exposures comparable 
to those referred to in Sec. 50.34(a)(1) or Sec. 100.11 of this 
chapter, as applicable.
* * * * *

PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION 
FACILITIES

    3. 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 (42 U.S.C. 5851). Section 50.10 also issued under secs. 101, 
185, 68 Stat. 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).

    4. 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, and revising the definition for Basic component 
(1)(iii) to read as follows:


Sec. 50.2  Definitions.

* * * * *
    Basic component * * *
    (1) * * *
    (iii) The capability to prevent or mitigate the consequences of 
accidents which could result in potential offsite exposures comparable 
to those referred to in Sec. 50.34(a)(1) or Sec. 100.11 of this 
chapter, as applicable.
* * * * *
    Committed dose equivalent means the dose equivalent to organs or 
tissues of

[[Page 65172]]

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 (1000mg/
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 shut down 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, as applicable.
* * * * *
    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).
* * * * *
    5. 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, 50.36b, 50.44, 50.46, 50.47, 50.48, 50.49, 50.54, 50.55, 
50.55a, 50.59, 50.60, 50.61, 50.62, 50.63, 50.64, 50.65, 50.66, 50.71, 
50.72, 50.74, 50.75, 50.80, 50.82, 50.90, 50.91, 50.120, and Appendices 
A, B, E, G, H, I, J, K, M, N, O, Q, R, and S to this part.
* * * * *
    6. 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 January 10, 
1997, 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

[[Page 65173]]

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 rem 7 total effective dose equivalent (TEDE).
---------------------------------------------------------------------------

    \7\ A whole body dose 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 exposure 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 dose to the public 
under accident conditions. Rather, this dose 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 (a)(1)(i), in support of the 
application for a construction permit, or a design approval.
* * * * *
    (12) On or after January 10, 1997, 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 to this part.
    (b) * * *
    (10) On or after January 10, 1997, 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, for those operating 
license applicants and holders whose construction permit was issued 
prior to January 10, 1997, the earthquake engineering criteria in 
Section VI of Appendix A to part 100 of this chapter continues to 
apply.
    (11) On or after January 10, 1997, 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. However, for either an operating 
license applicant or holder whose construction permit was issued prior 
to January 10, 1997, the reactor site criteria in part 100 of this 
chapter and the seismic and geologic siting criteria in Appendix A to 
part 100 of this chapter continues to apply.
* * * * *
    7. In Sec. 50.49, paragraph (b)(1) is revised to read as follows:


Sec. 50.49   Environmental qualification of electric equipment 
important to safety for nuclear power plants.

* * * * *
    (b) * * *
    (1) Safety-related electric equipment.3
---------------------------------------------------------------------------

    \3\ Safety-related electric equipment is referred to as ``Class 
1E'' equipment in IEEE 323-1974. Copies of this standard may be 
obtained from the Institute of Electrical and Electronics Engineers, 
Inc., 345 East 47th Street, New York, NY 10017.
---------------------------------------------------------------------------

    (i) This equipment is that relied upon to remain functional during 
and following design basis events to ensure--
    (A) The integrity of the reactor coolant pressure boundary;
    (B) The capability to shut down the reactor and maintain it in a 
safe shutdown condition; and
    (C) The capability to prevent or mitigate the consequences of 
accidents that could result in potential offsite exposures comparable 
to the guidelines in Sec. 50.34(a)(1) or Sec. 100.11 of this chapter, 
as applicable.
    (ii) Design basis events are defined as conditions of normal 
operation, including anticipated operational occurrences, design basis 
accidents, external events, and natural phenomena for which the plant 
must be designed to ensure functions (b)(1)(i) (A) through (C) of this 
section.
* * * * *
    8. 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 to this part, plant 
shutdown is required as provided in Paragraph IV(a)(3) of Appendix S to 
this part. 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.
    9. In Sec. 50.65, paragraph (b)(1) is revised to read as follows:


Sec. 50.65   Requirements for monitoring the effectiveness of 
maintenance at nuclear power plants

* * * * *
    (b) * * *
    (1) Safety related structures, systems, or components that are 
relied upon to remain functional during and following design basis 
events to ensure the integrity of the reactor coolant pressure 
boundary, the capability to shut down the reactor and maintain it in a 
safe shutdown condition, and the capability to prevent or mitigate the 
consequences of accidents that could result in potential offsite 
exposure comparable to the guidelines in Sec. 50.34(a)(1) or 
Sec. 100.11 of this chapter, as applicable.
* * * * *
    10. 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 January 10, 1997. However, for either an 
operating license applicant or holder whose construction permit was 
issued prior to January 10, 1997, the earthquake engineering 
criteria in Section VI of Appendix A to 10 CFR part 100 continues to 
apply.

I. Introduction

    (a) 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.

[[Page 65174]]

    (b) 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).
    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.
    (i) 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.
    (ii) 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 of this chapter.
    (iii) 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.
    (iv) 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 Operating Basis Earthquake Ground 
Motion, 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 and the licensing basis is 
maintained.
    (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 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

    11. 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).

    12. In Sec. 52.17, the introductory text of paragraph (a)(1) and 
paragraph (a)(1)(vi) are revised to read as follows:

[[Page 65175]]

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) of this 
chapter. 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 54--REQUIREMENTS FOR RENEWAL OF OPERATING LICENSES FOR NUCLEAR 
POWER PLANTS

    13. The authority citation for Part 54 continues to read as 
follows:

    Authority: Secs. 102, 103, 104, 161, 181, 182, 183, 186, 189, 68 
Stat. 936, 937, 938, 948, 953, 954, 955, 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, 202, 206, 88 Stat. 1242, 
1244, as amended (42 U.S.C. 5841, 5842).

    14. In Sec. 54.4, paragraph (a)(1)(iii) is revised to read as 
follows:


Sec. 54.4   Scope.

    (a) * * *
    (1) * * *
    (iii) The capability to prevent or mitigate the consequences of 
accidents that could result in potential offsite exposure comparable to 
the guidelines in Sec. 50.34(a)(1) or Sec. 100.11 of this chapter, as 
applicable.
* * * * *

PART 100--REACTOR SITE CRITERIA

    15. 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).

    16. The table of contents for Part 100 is revised to read as 
follows:

PART 100--REACTOR SITE CRITERIA

Sec.
100.1  Purpose.
100.2  Scope.
100.3  Definitions.
100.4  Communications.
100.8  Information collection requirements: OMB approval.

Subpart A--Evaluation Factors for Stationary Power Reactor Site 
Applications Before January 10, 1997 and for Testing Reactors

100.10  Factors to be considered when evaluating sites.
100.11  Determination of exclusion area, low population zone, and 
population center distance.

Subpart B--Evaluation Factors for Stationary Power Reactor Site 
Applications on or After January 10, 1997

100.20  Factors to be considered when evaluating sites.
100.21  Non-seismic site criteria.
100.23  Geologic and seismic siting criteria.

Appendix A to Part 100--Seismic and Geologic Siting Criteria for 
Nuclear Power Plants

    17. 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, 
that site characteristics are such that adequate security measures to 
protect the plant can be developed, and that physical characteristics 
unique to the proposed site that could pose a significant impediment to 
the development of emergency plans are identified.
    (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.
    18. 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 part 50 or part 52 of this chapter.
    19. 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

[[Page 65176]]

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 
Sec. 50.21(b) or Sec. 50.22 of this chapter designed to produce 
electrical or heat energy.
    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.
    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.
    20. 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 this part 100 should be addressed to the U.S. Nuclear 
Regulatory Commission, ATTN: Document Control Desk, Washington, DC 
20555-0001, 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.
    21. 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 1995 (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 to this part.
    22. The undesignated centerheading preceding Sec. 100.10 is 
removed, Secs. 100.10 and 100.11 are designated as subpart A, and the 
subpart A heading is added to read as follows:

Subpart A--Evaluation Factors for Stationary Power Reactor Site 
Applications Before January 10, 1997 and for Testing Reactors

    23. Subpart B consisting of Secs. 100.20, 100.21 and 100.23 is 
added to part 100 to read as follows:

Subpart B--Evaluation Factors for Stationary Power Reactor Site 
Applications on or After January 10, 1997


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 physical characteristics unique to the proposed site that 
could pose a significant impediment to the development of emergency 
plans are identified.
    (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) Section 100.23, ``Geologic and seismic siting factors,'' 
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) 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) Physical characteristics unique to the proposed site that could 
pose a significant impediment to the development of emergency plans 
must be identified;
    (h) Reactor sites should be located away from very densely 
populated

[[Page 65177]]

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 criteria.

    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 January 10, 1997. However, for either an 
operating license applicant or holder whose construction permit was 
issued prior to January 10, 1997, the seismic and geologic siting 
criteria in Appendix A to Part 100 of this chapter continues to apply.
    (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 2nd day of December, 1996.
    For the Nuclear Regulatory Commission.
John C. Hoyle,
Secretary of the Commission.
[FR Doc. 96-31075 Filed 12-10-96; 8:45 am]
BILLING CODE 7590-01-P