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