[Federal Register Volume 61, Number 239 (Wednesday, December 11, 1996)]
[Rules and Regulations]
[Pages 65157-65177]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-31075]
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NUCLEAR REGULATORY COMMISSION
10 CFR Parts 21, 50, 52, 54 and 100
RIN 3150-AD93
Reactor Site Criteria Including Seismic and Earthquake
Engineering Criteria for Nuclear Power Plants
AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its
regulations to update the criteria used in decisions regarding power
reactor siting, including geologic, seismic, and earthquake engineering
considerations for future nuclear power plants. The rule allows NRC to
benefit from experience gained in the application of the procedures and
methods set forth in the current regulation and to incorporate the
rapid advancements in the earth sciences and earthquake engineering.
This rule primarily consists of two separate changes, namely, the
source term and dose considerations, and the seismic and earthquake
engineering considerations of reactor siting. The Commission also is
denying the remaining issue in petition (PRM-50-20) filed by Free
Environment, Inc. et al.
EFFECTIVE DATE: January 10, 1997.
FOR FURTHER INFORMATION CONTACT: Dr. Andrew J. Murphy, Office of
Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, telephone (301) 415-6010, concerning the
seismic and earthquake engineering aspects and Mr. Charles E. Ader,
Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, telephone (301) 415-5622,
concerning other siting aspects.
SUPPLEMENTARY INFORMATION:
I. Background.
II. Objectives.
III. Genesis.
IV. Alternatives.
V. Major Changes.
A. Reactor Siting Criteria (Nonseismic).
B. Seismic and Earthquake Engineering Criteria.
VI. Related Regulatory Guides and Standard Review Plan Sections.
VII. Future Regulatory Action.
VIII. Referenced Documents.
IX. Summary of Comments on the Proposed Regulations.
A. Reactor Siting Criteria (Nonseismic).
B. Seismic and Earthquake Engineering Criteria.
X. Small Business Regulatory Enforcement Fairness Act
XI. Finding of No Significant Environmental Impact: Availability.
XII. Paperwork Reduction Act Statement.
XIII. Regulatory Analysis.
XIV. Regulatory Flexibility Certification.
XV. Backfit Analysis.
I. Background
The present regulation regarding reactor site criteria (10 CFR Part
100) was promulgated April 12, 1962 (27 FR 3509). NRC staff guidance on
exclusion area and low population zone sizes as well as population
density was issued in Regulatory Guide 4.7, ``General Site Suitability
Criteria for Nuclear Power Stations,'' published for comment in
September 1974. Revision 1 to this guide was issued in November 1975.
On June 1, 1976, the Public Interest Research Group (PIRG) filed a
petition for rulemaking (PRM-100-2) requesting that the NRC incorporate
minimum exclusion area and low population zone distances and population
density limits into the regulations. On April 28, 1977, Free
Environment, Inc. et al., filed a petition for rulemaking (PRM-50-20).
The remaining issue of this petition requests that the central Iowa
nuclear project and other reactors be sited at least 40 miles from
major population centers. In August 1978, the Commission directed the
NRC staff to develop a general policy statement on nuclear power
reactor siting. The ``Report of the Siting Policy Task Force'' (NUREG-
0625) was issued in August 1979 and provided recommendations regarding
siting of future nuclear power reactors. In the 1980 Authorization Act
for the NRC, the Congress directed the NRC to decouple siting from
design and to specify demographic criteria for siting. On July 29, 1980
(45 FR 50350), the NRC issued an Advance Notice of Proposed Rulemaking
(ANPRM) regarding revision of the reactor site criteria, which
discussed the recommendations of the Siting Policy Task Force and
sought public comments. The proposed rulemaking was deferred by the
Commission in December 1981 to await development of a Safety Goal and
improved research on accident source terms. On August 4, 1986 (51 FR
23044), the NRC issued its Policy Statement on Safety Goals that stated
quantitative health objectives with regard to both prompt and latent
cancer fatality risks. On December 14, 1988 (53 FR 50232), the NRC
denied PRM-100-2 on the basis that it would unnecessarily restrict
NRC's regulatory siting policies and would not result in a substantial
increase in the overall
[[Page 65158]]
protection of the public health and safety. The Commission is
addressing the remaining issue in PRM-50-20 as part of this rulemaking
action.
Appendix A, ``Seismic and Geologic Siting Criteria for Nuclear
Power Plants,'' to 10 CFR Part 100 was originally issued as a proposed
regulation on November 25, 1971 (36 FR 22601), published as a final
regulation on November 13, 1973 (38 FR 31279), and became effective on
December 13, 1973. There have been two amendments to 10 CFR Part 100,
Appendix A. The first amendment, issued November 27, 1973 (38 FR
32575), corrected the final regulation by adding the legend under the
diagram. The second amendment resulted from a petition for rulemaking
(PRM 100-1) requesting that an opinion be issued that would interpret
and clarify Appendix A with respect to the determination of the Safe
Shutdown Earthquake. A notice of filing of the petition was published
on May 14, 1975 (40 FR 20983). The substance of the petitioner's
proposal was accepted and published as an immediately effective final
regulation on January 10, 1977 (42 FR 2052).
The first proposed revision to these regulations was published for
public comment on October 20, 1992, (57 FR 47802). The availability of
the five draft regulatory guides and the standard review plan section
that were developed to provide guidance on meeting the proposed
regulations was published on November 25, 1992, (57 FR 55601). The
comment period for the proposed regulations was extended two times.
First, the NRC staff initiated an extension (58 FR 271; January 5,
1993) from February 17, 1993 to March 24, 1993, to be consistent with
the comment period on the draft regulatory guides and standard review
plan section. Second, in response to a request from the public, the
comment period was extended to June 1, 1993 (58 FR 16377; March 26,
1993).
The second proposed revision to these regulations was published for
public comment on October 17, 1994 (59 FR 52255). The NRC stated on
February 8, 1995, (60 FR 7467) that it intended to extend the comment
period to allow interested persons adequate time to provide comments on
staff guidance documents. On February 28, 1995, the availability of the
five draft regulatory guides and three standard review plan sections
that were developed to provide guidance on meeting the proposed
regulations was published (60 FR 10880) and the comment period for the
proposed rule was extended to May 12, 1995 (60 FR 10810).
II. Objectives
The objectives of this regulatory action are to--
1. State basic site criteria for future sites that, based upon
experience and importance to risk, have been shown as key to protecting
public health and safety;
2. Provide a stable regulatory basis for seismic and geologic
siting and applicable earthquake engineering design of future nuclear
power plants that will update and clarify regulatory requirements and
provide a flexible structure to permit consideration of new technical
understandings; and
3. Relocate source term and dose requirements that apply primarily
to plant design into 10 CFR Part 50.
III. Genesis
The regulatory action reflects changes that are intended to (1)
benefit from the experience gained in applying the existing regulation
and from research; (2) resolve interpretive questions; (3) provide
needed regulatory flexibility to incorporate state-of-the-art
improvements in the geosciences and earthquake engineering; and (4)
simplify the language to a more ``plain English'' text.
The new requirements in this rulemaking apply to applicants who
apply for a construction permit, operating license, preliminary design
approval, final design approval, manufacturing license, early site
permit, design certification, or combined license on or after the
effective date of the final regulations. However, for those operating
license applicants and holders whose construction permits were issued
prior to the effective date of this final regulation, the reactor site
criteria in 10 CFR Part 100, and the seismic and geologic siting
criteria and the earthquake engineering criteria in Appendix A to 10
CFR Part 100 would continue to apply in all subsequent proceedings,
including license amendments and renewal of operating licenses pursuant
to 10 CFR Part 54.
Criteria not associated with the selection of the site or
establishment of the Safe Shutdown Earthquake Ground Motion (SSE) have
been placed in 10 CFR Part 50. This action is consistent with the
location of other design requirements in 10 CFR Part 50.
Because the revised criteria presented in this final regulation
does not apply to existing plants, the licensing bases for existing
nuclear power plants must remain a part of the regulations. Therefore,
the non-seismic and seismic reactor site criteria for current plants is
retained as Subpart A and Appendix A to 10 CFR Part 100, respectively.
The revised reactor site criteria is added as Subpart B in 10 CFR Part
100 and applies to site applications received on or after the effective
date of the final regulations. Non-seismic site criteria is added as a
new Sec. 100.21 to Subpart B in 10 CFR Part 100. The criteria on
seismic and geologic siting is added as a new Sec. 100.23 to Subpart B
in 10 CFR Part 100. The dose calculations and the earthquake
engineering criteria is located in 10 CFR Part 50 (Sec. 50.34(a) and
Appendix S, respectively). Because Appendix S is not self executing,
applicable sections of Part 50 (Sec. 50.34 and Sec. 50.54) are revised
to reference Appendix S. The regulation also makes conforming
amendments to 10 CFR Parts 21, 50, 52, and 54. Sections 21.3,
50.49(b)(1), 50.65(b)(1), 52.17(a)(1), and 54.4(a)(1)(iii) are amended
to reflect changes in Sec. 50.34(a)(1) and 10 CFR Part 100.
IV. Alternatives
The first alternative considered by the Commission was to continue
using current regulations for site suitability determinations. This is
not considered an acceptable alternative. Accident source terms and
dose calculations currently primarily influence plant design
requirements rather than siting. It is desirable to state basic site
criteria which, through importance to risk, have been shown to be key
to assuring public health and safety. Further, significant advances in
understanding severe accident behavior, including fission product
release and transport, as well as in the earth sciences and in
earthquake engineering have taken place since the promulgation of the
present regulation and deserve to be reflected in the regulations.
The second alternative considered was replacement of the existing
regulation with an entirely new regulation. This is not an acceptable
alternative because the provisions of the existing regulations form
part of the licensing bases for many of the operating nuclear power
plants and others that are in various stages of obtaining operating
licenses. Therefore, these provisions should remain in force and
effect.
The approach of establishing the revised requirements in new
sections to 10 CFR Part 100 and relocating plant design requirements to
10 CFR Part 50 while retaining the existing regulation was chosen as
the best alternative. The public will benefit from a clearer, more
uniform, and more consistent licensing process that incorporates
updated information and is subject to fewer interpretations. The NRC
staff will
[[Page 65159]]
benefit from improved regulatory implementation (both technical and
legal), fewer interpretive debates, and increased regulatory
flexibility. Applicants will derive the same benefits in addition to
avoiding licensing delays caused by unclear regulatory requirements.
V. Major Changes
A. Reactor Siting Criteria (Nonseismic)
Since promulgation of the reactor site criteria in 1962, the
Commission has approved more than 75 sites for nuclear power reactors
and has had an opportunity to review a number of others. In addition,
light-water commercial power reactors have accumulated about 2000
reactor-years of operating experience in the United States. As a result
of these site reviews and operational experience, a great deal of
insight has been gained regarding the design and operation of nuclear
power plants as well as the site factors that influence risk. In
addition, an extensive research effort has been conducted to understand
accident phenomena, including fission product release and transport.
This extensive operational experience together with the insights gained
from recent severe accident research as well as numerous risk studies
on radioactive material releases to the environment under severe
accident conditions have all confirmed that present commercial power
reactor design, construction, operation and siting is expected to
effectively limit risk to the public to very low levels. These risk
studies include the early ``Reactor Safety Study'' (WASH-1400),
published in 1975, many Probabilistic Risk Assessment (PRA) studies
conducted on individual plants as well as several specialized studies,
and the recent ``Severe Accident Risks: An Assessment for Five U.S.
Nuclear Power Plants,'' (NUREG-1150), issued in 1990. Advanced reactor
designs currently under review are expected to result in even lower
risk and improved safety compared to existing plants. Hence, the
substantial base of knowledge regarding power reactor siting, design,
construction and operation reflects that the primary factors that
determine public health and safety are the reactor design, construction
and operation.
Siting factors and criteria, however, are important in assuring
that radiological doses from normal operation and postulated accidents
will be acceptably low, that natural phenomena and potential man-made
hazards will be appropriately accounted for in the design of the plant,
that site characteristics are such that adequate security measures to
protect the plant can be developed, and that physical characteristics
unique to the proposed site that could pose a significant impediment to
the development of emergency plans are identified. The Commission has
also had a long standing policy of siting reactors away from densely
populated centers, and is continuing this policy in this rule.
The Commission is incorporating basic reactor site criteria in this
rule to accomplish the above purposes. The Commission is retaining
source term and dose calculations to verify the adequacy of a site for
a specific plant, but source term and dose calculations are relocated
to Part 50, since experience has shown that these calculations have
tended to influence plant design aspects such as containment leak rate
or filter performance rather than siting. No specific source term is
referenced in Part 50. Rather, the source term is required to be one
that is ``* * * assumed to result in substantial meltdown of the core
with subsequent release into the containment of appreciable quantities
of fission products.'' Hence, this guidance can be utilized with the
source term currently used for light-water reactors, or used in
conjunction with revised accident source terms.
The relocation of source term and dose calculations to Part 50
represent a partial decoupling of siting from accident source term and
dose calculations. The siting criteria are envisioned to be utilized
together with standardized plant designs whose features will be
certified in a separate design certification rulemaking procedure. Each
of the standardized designs will specify an atmospheric dilution factor
that would be required to be met, in order to meet the dose criteria at
the exclusion area boundary. For a given standardized design, a site
having relatively poor dispersion characteristics would require a
larger exclusion area distance than one having good dispersion
characteristics. Additional design features would be discouraged in a
standardized design to compensate for otherwise poor site conditions.
Although individual plant tradeoffs will be discouraged for a given
standardized design, a different standardized design could require a
different atmospheric dilution factor. For custom plants that do not
involve a standardized design, the source term and dose criteria will
continue to provide assurance that the site is acceptable for the
proposed design.
Rationale for Individual Criteria
(A) Exclusion Area. An exclusion area surrounding the immediate
vicinity of the plant has been a requirement for siting power reactors
from the very beginning. This area provides a high degree of protection
to the public from a variety of potential plant accidents and also
affords protection to the plant from potential man-related hazards. The
Commission considers an exclusion area to be an essential feature of a
reactor site and is retaining this requirement, in Part 50, to verify
that an applicant's proposed exclusion area distance is adequate to
assure that the radiological dose to an individual will be acceptably
low in the event of a postulated accident. However, as noted above, if
source term and dose calculations are used in conjunction with
standardized designs, unlimited plant tradeoffs to compensate for poor
site conditions will not be permitted. For plants that do not involve
standardized designs, the source term and dose calculations will
provide assurance that the site is acceptable for the proposed design.
The present regulation requires that the exclusion area be of such
size that an individual located at any point on its boundary for two
hours immediately following onset of the postulated fission product
release would not receive a total radiation dose in excess of 25 rem to
the whole body or 300 rem to the thyroid gland. A footnote in the
present regulation notes that a whole body dose of 25 rem has been
stated to correspond numerically to the once in a lifetime accidental
or emergency dose to radiation workers which could be disregarded in
the determination of their radiation exposure status (NBS Handbook 69
dated June 5, 1959). However, the same footnote also clearly states
that the Commission's use of this value does not imply that it
considers it to be an acceptable limit for an emergency dose to the
public under accident conditions, but only that it represents a
reference value to be used for evaluating plant features and site
characteristics intended to mitigate the radiological consequences of
accidents in order to provide assurance of low risk to the public under
postulated accidents. The Commission, based upon extensive experience
in applying this criterion, and in recognition of the conservatism of
the assumptions in its application (a large fission product release
within containment associated with major core damage, maximum allowable
containment leak rate, a postulated single failure of any of the
fission product cleanup systems, such as the containment sprays,
adverse site
[[Page 65160]]
meteorological dispersion characteristics, an individual presumed to be
located at the boundary of the exclusion area at the centerline of the
plume for two hours without protective actions), believes that this
criterion has clearly resulted in an adequate level of protection. As
an illustration of the conservatism of this assessment, the maximum
whole body dose received by an actual individual during the Three Mile
Island accident in March 1979, which involved major core damage, was
estimated to be about 0.1 rem.
The proposed rule considered two changes in this area.
First, the Commission proposed that the use of different doses for
the whole body and thyroid gland be replaced by a single value of 25
rem, total effective dose equivalent (TEDE).
The proposed use of the total effective dose equivalent, or TEDE,
was noted as being consistent with Part 20 of the Commission's
regulations and was also based upon two considerations. First, since it
utilizes a risk consistent methodology to assess the radiological
impact of all relevant nuclides upon all body organs, use of TEDE
promotes a uniformity and consistency in assessing radiation risk that
may not exist with the separate whole body and thyroid organ dose
values in the present regulation. Second, use of TEDE lends itself
readily to the application of updated accident source terms, which can
vary not only with plant design, but in which additional nuclides,
besides the noble gases and iodine are predicted to be released into
containment.
The Commission considered the current dose criteria of 25 rem whole
body and 300 rem thyroid with the intent of selecting a TEDE numerical
value equivalent to the risk implied by the current dose criteria. The
Commission proposed to use the risk of latent cancer fatality as the
appropriate risk measure since quantitative health objectives (QHOs)
for it have been established in the Commission's Safety Goal policy.
Although the supplementary information in the proposed rule noted that
the current dose criteria are equivalent in risk to 27 rem TEDE, the
Commission proposed to use 25 rem TEDE as the dose criterion for plant
evaluation purposes, since this value is essentially the same level of
risk as the current criteria.
However, the Commission specifically requested comments on whether
the current dose criteria should be modified to utilize the total
effective dose equivalent or TEDE concept, whether a TEDE value of 25
rem (consistent with latent cancer fatality), or 34 rem (consistent
with latent cancer incidence), or some other value should be used, and
whether the dose criterion should also include a ``capping''
limitation, that is, an additional requirement that the dose to any
individual organ not be in excess of some fraction of the total.
Based on the comments received, there was a general consensus that
the use of the TEDE concept was appropriate, and a nearly unanimous
opinion that no organ ``capping'' dose was required, since the TEDE
concept provided the appropriate risk weighting for all body organs.
With regard to the value to be used as the dose criterion, a number
of comments were received that the proposed value of 25 rem TEDE
represented a more restrictive criterion than the current values of 25
rem whole body and 300 rem to the thyroid gland. These commenters noted
that the use of organ weighting factors of 1 for the whole body and
0.03 for the thyroid as given in 10 CFR Part 20, would yield a value of
34 rem TEDE for whole body and thyroid doses of 25 and 300 rem,
respectively. This is because the organ weighting factors in 10 CFR
Part 20 include other effects (e.g., genetic) in addition to latent
cancer fatality.
After careful consideration, the Commission has decided to adopt a
value of 25 rem TEDE as the dose acceptance criterion for the final
rule. The bases for this decision follows. First, the Commission has
generally based its regulations on the risk of latent cancer fatality.
Although a numerical calculation would lead to a value of 27 rem TEDE,
as noted in the discussion that accompanied the proposed rule, the
Commission concludes that a value of 25 rem is sufficiently close, and
that the use of 27 rather than 25 implies an unwarranted numerical
precision. In addition, in terms of occupational dose, Part 20 also
permits a once-in-a-lifetime planned special dose of 25 rem TEDE. In
addition, EPA guidance sets a limit of 25 rem TEDE for workers
performing emergency service such as lifesaving or protection of large
populations. While the Commission does not, as noted above, regard this
dose value as one that is acceptable for members of the public under
accident conditions, it provides a useful perspective with regard to
doses that ought not to be exceeded, even for radiation workers under
emergency conditions.
The argument that a criterion of 25 rem TEDE in conjunction with
the organ weighting factors of 10 CFR Part 20 for its calculation
represents a tightening of the dose criterion, while true in theory, is
not true in practice. A review of the dose analyses for operating
plants has shown that the thyroid dose limit of 300 rem has been the
limiting dose criterion in licensing reviews, and that all operating
plants would be able to meet a dose criterion of 25 rem TEDE. Hence,
the Commission concludes that, in practice, use of the organ weighting
factors of Part 20 together with a dose criterion of 25 rem TEDE,
represents a relaxation rather than a tightening of the dose criterion.
In adopting this value, the Commission also rejects the view, advanced
by some, that the dose calculation is merely a ``reference'' value that
bears no relation to what might be experienced by an actual person in
an accident. Although the Commission considers it highly unlikely that
an actual person would receive such a dose, because of the conservative
and stylized assumptions employed in its calculation, it is
conceivable.
The second change proposed in this area was in regard to the time
period that a hypothetical individual is assumed to be at the exclusion
area boundary. While the duration of the time period remains at a value
of two hours, the proposed rule stated that this time period not be
fixed in regard to the appearance of fission products within
containment, but that various two-hour periods be examined with the
objective that the dose to an individual not be in excess of 25 rem
TEDE for any two-hour period after the appearance of fission products
within containment. The Commission proposed this change to reflect
improved understanding of fission product release into the containment
under severe accident conditions. For an assumed instantaneous release
of fission products, as contemplated by the present rule, the two hour
period that commences with the onset of the fission product release
clearly results in the highest dose to an individual offsite. Improved
understanding of severe accidents shows that fission product releases
to the containment do not occur instantaneously, and that the bulk of
the releases may not take place for about an hour or more. Hence, the
two-hour period commencing with the onset of fission product release
may not represent the highest dose that an individual could be exposed
to over any two-hour period. As a result, the Commission proposed that
various two-hour periods be examined to assure that the dose to a
hypothetical individual at the exclusion area boundary would not be in
excess of 25 rem TEDE over any two-hour period after the onset of
fission product release.
A number of comments received in regard to this proposed criterion
stated that so-called ``sliding'' two-hour
[[Page 65161]]
window for dose evaluation at the exclusion area boundary was
confusing, illogical, and inappropriate. Several commenters felt it was
difficult to ascertain which two hour period represented the maximum.
Others expressed the view that the significance of such a calculation
was not clearly stated nor understood. For example, one comment
expressed the view that a dose evaluated for a ``sliding'' two-hour
period was logically inconsistent since it implied either that an
individual was not at the exclusion area boundary prior to the
accident, and approached close to the plant after initiation of the
accident, contrary to what might be expected, or that the individual
was, in fact, located at the exclusion area boundary all along, in
which case the dose contribution received prior to the ``maximum'' two-
hour value was being ignored.
Although the Commission recognizes that evaluation of the dose to a
hypothetical individual over any two-hour period may not be entirely
consistent with the actions of an actual individual in an accident, the
intent is to assure that the short-term dose to an individual will not
be in excess of the acceptable value, even where there is some
variability in the time that an individual might be located at the
exclusion area boundary. In addition, the dose calculation should not
be taken too literally with regard to the actions of a real individual,
but rather is intended primarily as a means to evaluate the
effectiveness of the plant design and site characteristics in
mitigating postulated accidents.
For these reasons, the Commission is retaining the requirement, in
the final rule, that the dose to an individual located at the nearest
exclusion area boundary over any two-hour period after the appearance
of fission products in containment, should not be in excess of 25 rem
total effective dose equivalent (TEDE).
(B) Site Dispersion Factors. Site dispersion factors have been
utilized to provide an assessment of dose to an individual as a result
of a postulated accident. Since the Commission is requiring that a
verification be made that the exclusion area distance is adequate to
assure that the guideline dose to a hypothetical individual will not be
exceeded under postulated accident conditions, as well as to assure
that radiological limits are met under normal operating conditions, the
Commission is requiring that the atmospheric dispersion characteristics
of the site be evaluated, and that site dispersion factors based upon
this evaluation be determined and used in assessing radiological
consequences of normal operations as well as accidents.
(C) Low Population Zone. The present regulation requires that a low
population zone (LPZ) be defined immediately beyond the exclusion area.
Residents are permitted in this area, but the number and density must
be such that there is a reasonable probability that appropriate
protective measures could be taken in their behalf in the event of a
serious accident. In addition, the nearest densely populated center
containing more than about 25,000 residents must be located no closer
than one and one-third times the outer boundary of the LPZ. Finally,
the dose to a hypothetical individual located at the outer boundary of
the LPZ over the entire course of the accident must not be in excess of
the dose values given in the regulation.
While the Commission considers that the siting functions intended
for the LPZ, namely, a low density of residents and the feasibility of
taking protective actions, have been accomplished by other regulations
or can be accomplished by other guidance, the Commission continues to
believe that a requirement that limits the radiological consequences
over the course of the accident provides a useful evaluation of the
plant's long-term capability to mitigate postulated accidents. For this
reason, the Commission is retaining the requirement that the dose
consequences be evaluated at the outer boundary of the LPZ over the
course of the postulated accident and that these not be in excess of 25
rem TEDE.
(D) Physical Characteristics of the Site. It has been required that
physical characteristics of the site, such as the geology, seismology,
hydrology, meteorology characteristics be considered in the design and
construction of any plant proposed to be located there. The final rule
requires that these characteristics be evaluated and that site
parameters, such as design basis flood conditions or tornado wind
loadings be established for use in evaluating any plant to be located
on that site in order to ensure that the occurrence of such physical
phenomena would pose no undue hazard.
(E) Nearby Transportation Routes, Industrial and Military
Facilities. As for natural phenomena, it has been a long-standing NRC
staff practice to review man-related activities in the site vicinity to
provide assurance that potential hazards associated with such
facilities or transportation routes will pose no undue risk to any
plant proposed to be located at the site. The final rule codifies this
practice.
(F) Adequacy of Security Plans. The rule requires that the
characteristics of the site be such that adequate security plans and
measures for the plant could be developed. The Commission envisions
that this will entail a small secure area considerably smaller than
that envisioned for the exclusion area.
(G) Emergency Planning. The proposed rule stated that the site
characteristics should be such that adequate plans to carry out
protective measures for members of the public in the event of emergency
could be developed. To avoid any misinterpretation that the Commission
is adopting emergency planning standards that implicitly overrule or
may be in conflict with previous Commission decisions (e.g., CLI-90-
02), the language in the final rule has been modified to be consistent
with that of section 52.17 of the Commission's regulations regarding
early site permits.
The Commission's decision in Seabrook on emergency planning, made
in connection with an operating license review for a site previously
approved, is being extended in considering site suitability for future
reactor sites. The Commission, in its Seabrook decision, CLI-90-02,
reiterated its earlier determination in the Shoreham decision, CLI-86-
13, that the adequacy of an emergency plan is to be determined by the
sixteen planning standards of 10 CFR 50.47(b), and that these standards
do not require that an adequate plan achieve a preset minimum radiation
dose saving or a minimum evacuation time for the plume exposure pathway
emergency planning zone in the event of a serious accident. Rather, the
Commission noted that emergency planning is required as a matter of
prudence and for defense-in-depth, and that the adequacy of an
emergency plan was to be judged on the basis of its meeting the 16
planning standards given in 10 CFR 50.47(b). Hence, the characteristics
of the site, which determine the evacuation time for the plume exposure
pathway emergency planning zone, have not entered into the
determination of the adequacy of an emergency plan. Emergency plans
developed according to the above planning standards will result in
reasonable assurance that adequate protective measures can be taken in
the event of emergency.
It is sufficient that an applicant identify any physical site
characteristics that could represent a significant impediment to the
development of emergency plans, primarily to assure that ``A range of
protective actions have been developed for the plume exposure pathway
emergency planning zone for
[[Page 65162]]
emergency workers and the public'', as stated in the planning
standards.
Accordingly, appropriate sections of the rule (e.g.,
Sec. 100.21(g)) have been modified to state that ``physical
characteristics unique to the proposed site that could pose a
significant impediment to the development of emergency plans must be
identified.'' Except for the deletion of the phrase ``such as egress
limitations from the area surrounding the site'', this language is
identical to that in Sec. 52.17(b)(1). This phrase is being deleted
from Sec. 100.21(g) (but Sec. 52.17(b)(1) remains unchanged), to
eliminate any confusion that might arise regarding its scope.
(H) Siting Away From Densely Populated Centers. Population density
considerations beyond the exclusion area have been required since
issuance of Part 100 in 1962. The current rule requires a ``low
population zone'' (LPZ) beyond the immediate exclusion area. The LPZ
boundary must be of such a size that an individual located at its outer
boundary must not receive a dose in excess of the values given in Part
100 over the course of the accident. While numerical values of
population or population density are not specified for this region, the
regulation also requires that the nearest boundary of a densely
populated center of about 25,000 or more persons be located no closer
than one and one-third times the LPZ outer boundary. Part 100 has no
population criteria other than the size of the LPZ and the proximity of
the nearest population center, but notes that ``where very large cities
are involved, a greater distance may be necessary.''
Whereas the exclusion area size is based upon limitation of
individual risk, population density requirements serve to set societal
risk limitations and reflect consideration of accidents beyond the
design basis, or severe accidents. Such accidents were clearly a
consideration in the original issuance of Part 100, since the Statement
of Considerations (27 FR 3509; April 12, 1962) noted that:
Further, since accidents of greater potential hazard than those
commonly postulated as representing an upper limit are conceivable,
although highly improbable, it was considered desirable to provide
for protection against excessive exposure doses to people in large
centers, where effective protective measures might not be feasible *
* * Hence, the population center distance was added as a site
requirement.
Limitation of population density beyond the exclusion area has the
following benefits:
(a) It facilitates emergency preparedness and planning; and
(b) It reduces potential doses to large numbers of people and
reduces property damage in the event of severe accidents.
Although the Commission's Safety Goal policy provides guidance on
individual risk limitations, in the form of the Quantitative Health
Objectives (QHO), it provides no guidance with regard to societal risk
limitations and therefore cannot be used to ascertain whether a
particular population density would meet the Safety Goal.
However, results of severe accident risk studies, particularly
those obtained from NUREG-1150, can provide useful insights for
considering potential criteria for population density. Severe accidents
having the highest consequences are those where core-melt together with
early bypass of or containment failure occurs. Such an event would
likely lead to a ``large release'' (without defining this precisely).
Based upon NUREG-1150, the probability of a core-melt accident together
with early containment failure or bypass for some current generation
LWRs is estimated to be between 10-5 and 10-6 per reactor
year. For future plants, this value is expected to be less than
10-6 per reactor year.
If a reactor was located nearer to a large city than current NRC
practice permitted, the likelihood of exposing a large number of people
to significant releases of radioactive material would be about the same
as the probability of a core-melt and early containment failure, that
is, less than 10-6 per reactor year for future reactor designs. It
is worth noting that events having the very low likelihood of about
10-6 per reactor year or lower have been regarded in past
licensing actions to be ``incredible'', and as such, have not been
required to be incorporated into the design basis of the plant. Hence,
based solely upon accident likelihood, it might be argued that siting a
reactor nearer to a large city than current NRC practice would pose no
undue risk.
If, however, a reactor were sited away from large cities, the
likelihood of the city being affected would be reduced because of two
factors. First, the likelihood that radioactive material would actually
be carried towards the city is reduced because it is likely that the
wind will blow in a direction away from the city. Second, the
radiological dose consequences would also be reduced with distance
because the radioactive material becomes increasingly diluted by the
atmosphere and the inventory becomes depleted due to the natural
processes of fallout and rainout before reaching the city. Analyses
indicate that if a reactor were located at distances ranging from 10 to
about 20 miles away from a city, depending upon its size, the
likelihood of exposure of large numbers of people within the city would
be reduced by factors of ten to one hundred or more compared with
locating a reactor very close to a city.
In summary, next-generation reactors are expected to have risk
characteristics sufficiently low that the safety of the public is
reasonably assured by the reactor and plant design and operation
itself, resulting in a very low likelihood of occurrence of a severe
accident. Such a plant can satisfy the QHOs of the Safety Goal with a
very small exclusion area distance (as low as 0.1 miles). The
consequences of design basis accidents, analyzed using revised source
terms and with a realistic evaluation of engineered safety features,
are likely to be found acceptable at distances of 0.25 miles or less.
With regard to population density beyond the exclusion area, siting a
reactor closer to a densely populated city than is current NRC practice
would pose a very low risk to the populace.
Nevertheless, the Commission concludes that defense-in-depth
considerations and the additional enhancement in safety to be gained by
siting reactors away from densely populated centers should be
maintained.
The Commission is incorporating a two-tier approach with regard to
population density and reactor sites. The rule requires that reactor
sites be located away from very densely populated centers, and that
areas of low population density are, generally, preferred. The
Commission believes that a site not falling within these two
categories, although not preferred, can be found acceptable under
certain conditions.
The Commission is not establishing specific numerical criteria for
evaluation of population density in siting future reactor facilities
because the acceptability of a specific site from the standpoint of
population density must be considered in the overall context of safety
and environmental considerations. The Commission's intent is to assure
that a site that has significant safety, environmental or economic
advantages is not rejected solely because it has a higher population
density than other available sites. Population density is but one
factor that must be balanced against the other advantages and
disadvantages of a particular site in determining the site's
acceptability. Thus, it must be recognized that sites with higher
population density, so long as they are located away from very densely
populated centers, can be approved by
[[Page 65163]]
the Commission if they present advantages in terms of other
considerations applicable to the evaluation of proposed sites.
Petition Filed By Free Environment, Inc. et al.
On April 28, 1977, Free Environment, Inc. et al., filed a petition
for rulemaking (PRM-50-20) requesting, among other things, that ``the
central Iowa nuclear project and other reactors be sited at least 40
miles from major population centers.'' The petitioner also stated that
``locating reactors in sparsely-populated areas * * * has been endorsed
in non-binding NRC guidelines for reactor siting.'' The petitioner did
not specify what constituted a major population center. The only NRC
guidelines concerning population density in regard to reactor siting
are in Regulatory Guide 4.7, issued in 1974, and revised in 1975, prior
to the date of the petition. This guide states population density
values of 500 persons per square mile out to a distance of 30 miles
from the reactor, not 40 miles.
Regulatory Guide 4.7 does provide effective separation from
population centers of various sizes. Under this guide, a population
center of about 25,000 or more residents should be no closer than 4
miles (6.4 km) from a reactor because a density of 500 persons per
square mile within this distance would yield a total population of
about 25,000 persons. Similarly, a city of 100,000 or more residents
should be no closer than about 10 miles (16 km); a city of 500,000 or
more persons should be no closer than about 20 miles (32 km), and a
city of 1,000,000 or more persons should be no closer than about 30
miles (50 km) from the reactor.
The Commission has examined these guidelines with regard to the
Safety Goal. The Safety Goal quantitative health objective in regard to
latent cancer fatality states that, within a distance of ten miles (16
km) from the reactor, the risk to the population of latent cancer
fatality from nuclear power plant operation, including accidents,
should not exceed one-tenth of one percent of the likelihood of latent
cancer fatalities from all other causes. In addition to the risks of
latent cancer fatalities, the Commission has also investigated the
likelihood and extent of land contamination arising from the release of
long-lived radioactive species, such as cesium-137, in the event of a
severe reactor accident.
The results of these analyses indicate that the latent cancer
fatality quantitative health objective noted is met for current plant
designs. From analysis done in support of this proposed change in
regulation, the likelihood of permanent relocation of people located
more than about 20 miles (32 km) from the reactor as a result of land
contamination from a severe accident is very low. A revision of
Regulatory Guide 4.7 which incorporated this finding that population
density guidance beyond 20 miles was not needed in the evaluation of
potential reactor sites was issued for comment at the time of the
proposed rule. No comments were received on this aspect of the guide.
Therefore, the Commission concludes that the NRC staff guidance in
Regulatory Guide 4.7 provide a means of locating reactors away from
population centers, including ``major'' population centers, depending
upon their size, that would limit societal consequences significantly,
in the event of a severe accident. The Commission finds that granting
of the petitioner's request to specify population criteria out to 40
miles would not substantially reduce the risks to the public. As noted,
the Commission also believes that a higher population density site
could be found to be acceptable, compared to a lower population density
site, provided there were safety, environmental, or economic advantages
to the higher population site. Granting of the petitioner's request
would neglect this possibility and would make population density the
sole criterion of site acceptability. For these reasons, the Commission
has decided not to adopt the proposal by Free Environment,
Incorporated.
The Commission also notes that future population growth around a
nuclear power plant site, as in other areas of the region, is expected
but cannot be predicted with great accuracy, particularly in the long-
term. Population growth in the site vicinity will be periodically
factored into the emergency plan for the site, but since higher
population density sites are not unacceptable, per se, the Commission
does not intend to consider license conditions or restrictions upon an
operating reactor solely upon the basis that the population density
around it may reach or exceed levels that were not expected at the time
of site approval. Finally, the Commission wishes to emphasize that
population considerations as well as other siting requirements apply
only for the initial siting for new plants and will not be used in
evaluating applications for the renewal of existing nuclear power plant
licenses.
Change to 10 CFR Part 50
The change to 10 CFR Part 50 relocates from 10 CFR Part 100 the
dose requirements for each applicant at specified distances. Because
these requirements affect reactor design rather than siting, they are
more appropriately located in 10 CFR Part 50.
These requirements apply to future applicants for a construction
permit, design certification, or an operating license. The Commission
will consider after further experience in the review of certified
designs whether more specific requirements need to be developed
regarding revised accident source terms and severe accident insights.
B. Seismic and Earthquake Engineering Criteria
The following major changes to Appendix A, ``Seismic and Geologic
Siting Criteria for Nuclear Power Plants,'' to 10 CFR Part 100, are
associated with the seismic and earthquake engineering criteria
rulemaking. These changes reflect new information and research results,
and incorporate the intentions of this regulatory action as defined in
Section III of this rule. Much of the following discussion remains
unchanged from that issued for public comment (59 FR 52255) because
there were no comments which necessitated a major change to the
regulations and supporting documentation.
1. Separate Siting From Design
Criteria not associated with site suitability or establishment of
the Safe Shutdown Earthquake Ground Motion (SSE) have been placed into
10 CFR Part 50. This action is consistent with the location of other
design requirements in 10 CFR Part 50. Because the revised criteria
presented in the regulation will not be applied to existing plants, the
licensing basis for existing nuclear power plants must remain part of
the regulations. The criteria on seismic and geologic siting would be
designated as a new Sec. 100.23 to Subpart B in 10 CFR Part 100.
Criteria on earthquake engineering would be designated as a new
Appendix S, ``Earthquake Engineering Criteria for Nuclear Power
Plants,'' to 10 CFR Part 50.
2. Remove Detailed Guidance From the Regulation
Appendix A to 10 CFR Part 100 contains both requirements and
guidance on how to satisfy the requirements. For example, Section IV,
``Required Investigations,'' of Appendix A, states that investigations
are required for vibratory ground motion, surface faulting, and
seismically induced floods and water waves. Appendix A then provides
detailed guidance on what constitutes an acceptable investigation.
[[Page 65164]]
A similar situation exists in Section V, ``Seismic and Geologic Design
Bases,'' of Appendix A.
Geoscience assessments require considerable latitude in judgment.
This latitude in judgment is needed because of limitations in data and
the state-of-the-art of geologic and seismic analyses and because of
the rapid evolution taking place in the geosciences in terms of
accumulating knowledge and in modifying concepts. This need appears to
have been recognized when the existing regulation was developed. The
existing regulation states that it is based on limited geophysical and
geological information and will be revised as necessary when more
complete information becomes available.
However, having geoscience assessments detailed and cast in a
regulation has created difficulty for applicants and the staff in terms
of inhibiting the use of needed latitude in judgment. Also, it has
inhibited flexibility in applying basic principles to new situations
and the use of evolving methods of analyses (for instance,
probabilistic) in the licensing process.
The final regulation is streamlined, becoming a new section in
Subpart B to 10 CFR Part 100 rather than a new appendix to Part 100.
Also, the level of detail presented in the final regulation is reduced
considerably. Thus, the final regulation contains: (a) required
definitions, (b) a requirement to determine the geological,
seismological, and engineering characteristics of the proposed site,
and (c) requirements to determine the Safe Shutdown Earthquake Ground
Motion (SSE), to determine the potential for surface deformation, and
to determine the design bases for seismically induced floods and water
waves. The guidance documents describe how to carry out these required
determinations. The key elements of the approach to determine the SSE
are presented in the following section. The elements are the guidance
that is described in Regulatory Guide 1.165, ``Identification and
Characterization of Seismic Sources and Determination of Safe Shutdown
Earthquake Ground Motions.''
3. Uncertainties and Probabilistic Methods
The existing approach for determining a Safe Shutdown Earthquake
Ground Motion (SSE) for a nuclear reactor site, embodied in Appendix A
to 10 CFR Part 100, relies on a ``deterministic'' approach. Using this
deterministic approach, an applicant develops a single set of
earthquake sources, develops for each source a postulated earthquake to
be used as the source of ground motion that can affect the site,
locates the postulated earthquake according to prescribed rules, and
then calculates ground motions at the site.
Although this approach has worked reasonably well for the past two
decades, in the sense that SSEs for plants sited with this approach are
judged to be suitably conservative, the approach has not explicitly
recognized uncertainties in geosciences parameters. Because of
uncertainties about earthquake phenomena (especially in the eastern
United States), there have often been differences of opinion and
differing interpretations among experts as to the largest earthquakes
to be considered and ground-motion models to be used, thus often making
the licensing process relatively unstable.
Over the past decade, analysis methods for incorporating these
different interpretations have been developed and used. These
``probabilistic'' methods have been designed to allow explicit
incorporation of different models for zonation, earthquake size, ground
motion, and other parameters. The advantage of using these
probabilistic methods is their ability not only to incorporate
different models and different data sets, but also to weight them using
judgments as to the validity of the different models and data sets, and
thereby providing an explicit expression for the uncertainty in the
ground motion estimates and a means of assessing sensitivity to various
input parameters. Another advantage of the probabilistic method is the
target exceedance probability is set by examining the design bases of
more recently licensed nuclear power plants.
The final regulation explicitly recognizes that there are inherent
uncertainties in establishing the seismic and geologic design
parameters and allows for the option of using a probabilistic seismic
hazard methodology capable of propagating uncertainties as a means to
address these uncertainties. The rule further recognizes that the
nature of uncertainty and the appropriate approach to account for it
depend greatly on the tectonic regime and parameters, such as, the
knowledge of seismic sources, the existence of historical and recorded
data, and the understanding of tectonics. Therefore, methods other than
the probabilistic methods, such as sensitivity analyses, may be
adequate for some sites to account for uncertainties.
Methods acceptable to the NRC staff for implementing the regulation
are described in Regulatory Guide 1.165, ``Identification and
Characterization of Seismic Sources and Determination of Safe Shutdown
Earthquake Ground Motion.'' The key elements of this approach are:
--Conduct site-specific and regional geoscience investigations,
--Target exceedance probability is set by examining the design bases of
more recently licensed nuclear power plants,
--Conduct probabilistic seismic hazard analysis and determine ground
motion level corresponding to the target exceedance probability
--Determine if information from the regional and site geoscience
investigations change probabilistic results,
--Determine site-specific spectral shape and scale this shape to the
ground motion level determined above,
--NRC staff review using all available data including insights and
information from previous licensing experience, and
--Update the data base and reassess probabilistic methods at least
every ten years.
Thus, the approach requires thorough regional and site-specific
geoscience investigations. Results of the regional and site-specific
investigations must be considered in applications of the probabilistic
method. The current probabilistic methods, the NRC sponsored study
conducted by Lawrence Livermore National Laboratory (LLNL) or the
Electric Power Research Institute (EPRI) seismic hazard study, are
regional studies without detailed information on any specific location.
The regional and site-specific investigations provide detailed
information to update the database of the hazard methodology as
necessary.
It is also necessary to incorporate local site geological factors
such as structural geology, stratigraphy, and topography and to account
for site-specific geotechnical properties in establishing the design
basis ground motion. In order to incorporate local site factors and
advances in ground motion attenuation models, ground motion
characteristics are determined using the procedures outlined in
Standard Review Plan Section 2.5.2, ``Vibratory Ground Motion,''
Revision 3.
The NRC staff's review approach to evaluate ground motion estimates
is described in SRP Section 2.5.2, Revision 3. This review takes into
account the information base developed in licensing more than 100
plants. Although the basic premise in establishing the target
exceedance probability is that the current design levels are adequate,
a staff review further assures that there is
[[Page 65165]]
consistency with previous licensing decisions and that the scientific
bases for decisions are clearly understood. This review approach will
also assess the fairly complex regional probabilistic modeling, which
incorporates multiple hypotheses and a multitude of parameters.
Furthermore, the NRC staff's Safety Evaluation Report should provide a
clear basis for the staff's decisions and facilitate communication with
nonexperts.
4. Safe Shutdown Earthquake
The existing regulation (10 CFR Part 100, Appendix A, Section
V(a)(1)(iv)) states ``The maximum vibratory accelerations of the Safe
Shutdown Earthquake at each of the various foundation locations of the
nuclear power plant structures at a given site shall be determined * *
*'' The location of the seismic input motion control point as stated in
the existing regulation has led to confrontations with many applicants
that believe this stipulation is inconsistent with good engineering
fundamentals.
The final regulation moves the location of the seismic input motion
control point from the foundation-level to the free-field at the free
ground surface. The 1975 version of the Standard Review Plan placed the
control motion in the free-field. The final regulation is also
consistent with the resolution of Unresolved Safety Issue (USI) A-40,
``Seismic Design Criteria'' (August 1989), that resulted in the
revision of Standard Review Plan Sections 2.5.2, 3.7.1, 3.7.2, and
3.7.3. The final regulation also requires that the horizontal component
of the Safe Shutdown Earthquake Ground Motion in the free-field at the
foundation level of the structures must be an appropriate response
spectrum considering the site geotechnical properties, with a peak
ground acceleration of at least 0.1g.
5. Value of the Operating Basis Earthquake Ground Motion (OBE) and
Required OBE Analyses
The existing regulation (10 CFR Part 100, Appendix A, Section
V(a)(2)) states that the maximum vibratory ground motion of the OBE is
at least one half the maximum vibratory ground motion of the Safe
Shutdown Earthquake ground motion. Also, the existing regulation (10
CFR Part 100, Appendix A, Section VI(a)(2)) states that the engineering
method used to insure that structures, systems, and components are
capable of withstanding the effects of the OBE shall involve the use of
either a suitable dynamic analysis or a suitable qualification test. In
some cases, for instance piping, these multi-facets of the OBE in the
existing regulation made it possible for the OBE to have more design
significance than the SSE. A decoupling of the OBE and SSE has been
suggested in several documents. For instance, the NRC staff, SECY-79-
300, suggested that a compromise is required between design for a broad
spectrum of unlikely events and optimum design for normal operation.
Design for a single limiting event (the SSE) and inspection and
evaluation for earthquakes in excess of some specified limit (the OBE),
when and if they occur, may be the most sound regulatory approach.
NUREG-1061, ``Report of the U.S. Nuclear Regulatory Commission Piping
Review Committee,'' Vol.5, April 1985, (Table 10.1) ranked a decoupling
of the OBE and SSE as third out of six high priority changes. In SECY-
90-016, ``Evolutionary Light Water Reactor (LWR) Certification Issues
and Their Relationship to Current Regulatory Requirements,'' the NRC
staff states that it agrees that the OBE should not control the design
of safety systems. Furthermore, the final safety evaluation reports
related to the certification of the System 80+ and the Advanced Boiling
Water Reactor design (NUREG-1462 and NUREG-1503, respectively) have
already adopted the single earthquake design philosophy.
Activities equivalent to OBE-SSE decoupling are also being done in
foreign countries. For instance, in Germany their new design standard
requires only one design basis earthquake (equivalent to the SSE). They
require an inspection-level earthquake (for shutdown) of 0.4 SSE. This
level was set so that the vibratory ground motion should not induce
stresses exceeding the allowable stress limits originally required for
the OBE design.
The final regulation allows the value of the OBE to be set at (i)
one-third or less of the SSE, where OBE requirements are satisfied
without an explicit response or design analyses being performed, or
(ii) a value greater than one-third of the SSE, where analysis and
design are required. There are two issues the applicant should consider
in selecting the value of the OBE: first, plant shutdown is required if
vibratory ground motion exceeding that of the OBE occurs (discussed
below in Item 6, Required Plant Shutdown), and second, the amount of
analyses associated with the OBE. An applicant may determine that at
one-third of the SSE level, the probability of exceeding the OBE
vibratory ground motion is too high, and the cost associated with plant
shutdown for inspections and testing of equipment and structures prior
to restarting the plant is unacceptable. Therefore, the applicant may
voluntarily select an OBE value at some higher fraction of the SSE to
avoid plant shutdowns. However, if an applicant selects an OBE value at
a fraction of the SSE higher than one-third, a suitable analysis shall
be performed to demonstrate that the requirements associated with the
OBE are satisfied. The design shall take into account soil-structure
interaction effects and the expected duration of the vibratory ground
motion. The requirement associated with the OBE is that all structures,
systems, and components of the nuclear power plant necessary for
continued operation without undue risk to the health and safety of the
public shall remain functional and within applicable stress, strain and
deformation limits when subjected to the effects of the OBE in
combination with normal operating loads.
As stated, it is determined that if an OBE of one-third or less of
the SSE is used, the requirements of the OBE can be satisfied without
the applicant performing any explicit response analyses. In this case,
the OBE serves the function of an inspection and shutdown earthquake.
Some minimal design checks and the applicability of this position to
seismic base isolation of buildings are discussed below. There is high
confidence that, at this ground-motion level with other postulated
concurrent loads, most critical structures, systems, and components
will not exceed currently used design limits. This is ensured, in part,
because PRA insights will be used to support a margins-type assessment
of seismic events. A PRA-based seismic margins analysis will consider
sequence-level High Confidence, Low Probability of Failures (HCLPFs)
and fragilities for all sequences leading to core damage or containment
failures up to approximately one and two-thirds the ground motion
acceleration of the design basis SSE (Reference: Item II.N, Site-
Specific Probabilistic Risk Assessment and Analysis of External Events,
memorandum from Samuel J. Chilk to James M. Taylor, Subject: SECY-93-
087--Policy, Technical, and Licensing Issues Pertaining to Evolutionary
and Advance Light-Water Reactor (ALWR) Designs, dated July 21, 1993).
There are situations associated with current analyses where only
the OBE is associated with the design requirements, for example, the
ultimate heat sink (see Regulatory Guide 1.27, ``Ultimate Heat Sink for
Nuclear Power Plants''). In these situations, a value expressed as a
fraction of the SSE
[[Page 65166]]
response would be used in the analyses. Section VII of this final rule
identifies existing guides that would be revised technically to
maintain the existing design philosophy.
In SECY-93-087, ``Policy, Technical, and Licensing Issues
Pertaining to Evolutionary and Advance Light-Water Reactor (ALWR)
Designs,'' the NRC staff requested Commission approval on 42 technical
and policy issues pertaining to either evolutionary LWRs, passive LWRs,
or both. The issue pertaining to the elimination of the OBE is
designated I.M. The NRC staff identified actions necessary for the
design of structures, systems, and components when the OBE design
requirement is eliminated. The NRC staff clarified that guidelines
should be maintained to ensure the functionality of components,
equipment, and their supports. In addition, the NRC staff clarified how
certain design requirements are to be considered for buildings and
structures that are currently designed for the OBE, but not the SSE.
Also, the NRC staff has evaluated the effect on safety of eliminating
the OBE from the design load combinations for selected structures,
systems, and components and has developed proposed criteria for an
analysis using only the SSE. Commission approval is documented in the
Chilk to Taylor memorandum dated July 21, 1993, cited above.
More than one earthquake response analysis for a seismic base
isolated nuclear power plant design may be necessary to ensure adequate
performance at all earthquake levels. Decisions pertaining to the
response analyses associated with base isolated facilities will be
handled on a case by case basis.
6. Required Plant Shutdown
The current regulation (Section V(a)(2)) states that if vibratory
ground motion exceeding that of the OBE occurs, shutdown of the nuclear
power plant will be required. The supplementary information to the
final regulation (published November 13, 1973; 38 FR 31279, Item 6e)
includes the following statement: ``A footnote has been added to
Sec. 50.36(c)(2) of 10 CFR Part 50 to assure that each power plant is
aware of the limiting condition of operation which is imposed under
Section V(2) of Appendix A to 10 CFR Part 100. This limitation requires
that if vibratory ground motion exceeding that of the OBE occurs,
shutdown of the nuclear power plant will be required. Prior to resuming
operations, the licensee will be required to demonstrate to the
Commission that no functional damage has occurred to those features
necessary for continued operation without undue risk to the health and
safety of the public.'' At that time, it was the intention of the
Commission to treat the OBE as a limiting condition of operation. From
the statement in the Supplementary Information, the Commission directed
applicants to specifically review 10 CFR Part 100 to be aware of this
intention in complying with the requirements of 10 CFR 50.36. Thus, the
requirement to shut down if an OBE occurs was expected to be
implemented by being included among the technical specifications
submitted by applicants after the adoption of Appendix A. In fact,
applicants did not include OBE shutdown requirements in their technical
specifications.
The final regulation treats plant shutdown associated with
vibratory ground motion exceeding the OBE or significant plant damage
as a condition in every operating license. A new Sec. 50.54(ff) is
added to the regulations to require a process leading to plant shutdown
for licensees of nuclear power plants that comply with the earthquake
engineering criteria in Paragraph IV(a)(3) of Appendix S, ``Earthquake
Engineering Criteria for Nuclear Power Plants,'' to 10 CFR Part 50.
Immediate shutdown could be required until it is determined that
structures, systems, and components needed for safe shutdown are still
functional.
Regulatory Guide 1.166, ``Pre-Earthquake Planning and Immediate
Nuclear Power Plant Operator Post-Earthquake Actions,'' provides
guidance acceptable to the NRC staff for determining whether or not
vibratory ground motion exceeding the OBE ground motion or significant
plant damage had occurred and the timing of nuclear power plant
shutdown. The guidance is based on criteria developed by the Electric
Power Research Institute (EPRI). The decision to shut down the plant
should be made by the licensee within eight hours after the earthquake.
The data from the seismic instrumentation, coupled with information
obtained from a plant walk down, are used to make the determination of
when the plant should be shut down, if it has not already been shut
down by operational perturbations resulting from the seismic event. The
guidance in Regulatory Guide 1.166 is based on two assumptions, first,
that the nuclear power plant has operable seismic instrumentation,
including the equipment and software required to process the data
within four hours after an earthquake, and second, that the operator
walk down inspections can be performed in approximately four to eight
hours depending on the number of personnel conducting the inspection.
The regulation also includes a provision that requires the licensee to
consult with the Commission and to propose a plan for the timely, safe
shutdown of the nuclear power plant if systems, structures, or
components necessary for a safe shutdown or to maintain a safe shutdown
are not available.
Regulatory Guide 1.167, ``Restart of a Nuclear Power Plant Shut
Down by a Seismic Event,'' provides guidelines that are acceptable to
the NRC staff for performing inspections and tests of nuclear power
plant equipment and structures prior to plant restart. This guidance is
also based on EPRI reports. Prior to resuming operations, the licensee
must demonstrate to the Commission that no functional damage has
occurred to those features necessary for continued operation without
undue risk to the health and safety of the public. The results of post-
shutdown inspections, operability checks, and surveillance tests must
be documented in written reports and submitted to the Director, Office
of Nuclear Reactor Regulation. The licensee shall not resume operation
until authorized to do so by the Director, Office of Nuclear Reactor
Regulation.
7. Clarify Interpretations
Section 100.23 resolves questions of interpretation. As an example,
definitions and required investigations stated in the final regulation
do not contain the phrases in Appendix A to Part 100 that were more
applicable to only the western part of the United States.
The institutional definition for ``safety-related structures,
systems, and components'' is drawn from Appendix A to Part 100 under
III(c) and VI(a). With the relocation of the earthquake engineering
criteria to Appendix S to Part 50 and the relocation and modification
to dose guidelines in Sec. 50.34(a)(1), the definition of safety-
related structures, systems, and components is included in Part 50
definitions with references to both the Part 100 and Part 50 dose
guidelines.
VI. Related Regulatory Guides and Standard Review Plan Sections
The NRC is developing the following regulatory guides and standard
review plan sections to provide prospective licensees with the
necessary guidance for implementing the final regulation. The notice of
availability for these materials will be published in a later issue of
the Federal Register.
1. Regulatory Guide 1.165, ``Identification and Characterization of
Seismic Sources and Determination of
[[Page 65167]]
Shutdown Earthquake Ground Motions.'' The guide provides general
guidance and recommendations, describes acceptable procedures and
provides a list of references that present acceptable methodologies to
identify and characterize capable tectonic sources and seismogenic
sources. Section V.B.3 of this rule describes the key elements.
2. Regulatory Guide 1.12, Revision 2, ``Nuclear Power Plant
Instrumentation for Earthquakes.'' The guide describes seismic
instrumentation type and location, operability, characteristics,
installation, actuation, and maintenance that are acceptable to the NRC
staff.
3. Regulatory Guide 1.166, ``Pre-Earthquake Planning and Immediate
Nuclear Power Plant Operator Post-Earthquake Actions.'' The guide
provides guidelines that are acceptable to the NRC staff for a timely
evaluation of the recorded seismic instrumentation data and to
determine whether or not plant shutdown is required.
4. Regulatory Guide 1.167, ``Restart of a Nuclear Power Plant Shut
Down by a Seismic Event.'' The guide provides guidelines that are
acceptable to the NRC staff for performing inspections and tests of
nuclear power plant equipment and structures prior to restart of a
plant that has been shut down because of a seismic event.
5. Standard Review Plan Section 2.5.1, Revision 3, ``Basic Geologic
and Seismic Information.'' This SRP Section describes procedures to
assess the adequacy of the geologic and seismic information cited in
support of the applicant's conclusions concerning the suitability of
the plant site.
6. Standard Review Plan Section 2.5.2, Revision 3 ``Vibratory
Ground Motion.'' This SRP Section describes procedures to assess the
ground motion potential of seismic sources at the site and to assess
the adequacy of the SSE.
7. Standard Review Plan Section 2.5.3, Revision 3, ``Surface
Faulting.'' This SRP Section describes procedures to assess the
adequacy of the applicant's submittal related to the existence of a
potential for surface faulting affecting the site.
8. Regulatory Guide 4.7, Revision 2, ``General Site Suitability
Criteria for Nuclear Power Plants.'' This guide discusses the major
site characteristics related to public health and safety and
environmental issues that the NRC staff considers in determining the
suitability of sites.
VII. Future Regulatory Action
Several existing regulatory guides will be revised to incorporate
editorial changes or maintain the existing design or analysis
philosophy. These guides will be issued as final guides without public
comment subsequent to the publication of the final regulations.
The following regulatory guides will be revised to incorporate
editorial changes, for example to reference new sections to Part 100 or
Appendix S to Part 50. No technical changes will be made in these
regulatory guides.
1. 1.57, ``Design Limits and Loading Combinations for Metal Primary
Reactor Containment System Components.''
2. 1.59, ``Design Basis Floods for Nuclear Power Plants.''
3. 1.60, ``Design Response Spectra for Seismic Design of Nuclear
Power Plants.''
4. 1.83, ``Inservice Inspection of Pressurized Water Reactor Steam
Generator Tubes.''
5. 1.92, ``Combining Modal Responses and Spatial Components in
Seismic Response Analysis.''
6. 1.102, ``Flood Protection for Nuclear Power Plants.''
7. 1.121, ``Bases for Plugging Degraded PWR Steam Generator
Tubes.''
8. 1.122, ``Development of Floor Design Response Spectra for
Seismic Design of Floor-Supported Equipment or Components.''
The following regulatory guides will be revised to update the
design or analysis philosophy, for example, to change OBE to a fraction
of the SSE:
1. 1.3, ``Assumptions Used for Evaluating the Potential
Radiological Consequences of a Loss of Coolant Accident for Boiling
Water Reactors.''
2. 1.4, ``Assumptions Used for Evaluating the Potential
Radiological Consequences of a Loss of Coolant Accident for Pressurized
Water Reactors.''
3. 1.27, ``Ultimate Heat Sink for Nuclear Power Plants.''
4. 1.100, ``Seismic Qualification of Electric and Mechanical
Equipment for Nuclear Power Plants.''
5. 1.124, ``Service Limits and Loading Combinations for Class 1
Linear-Type Component Supports.''
6. 1.130, ``Service Limits and Loading Combinations for Class 1
Plate-and-Shell-Type Component Supports.''
7. 1.132, ``Site Investigations for Foundations of Nuclear Power
Plants.''
8. 1.138, ``Laboratory Investigations of Soils for Engineering
Analysis and Design of Nuclear Power Plants.''
9. 1.142, ``Safety-Related Concrete Structures for Nuclear Power
Plants (Other than Reactor Vessels and Containments).''
10. 1.143, ``Design Guidance for Radioactive Waste Management
Systems, Structures, and Components Installed in Light-Water-Cooled
Nuclear Power Plants.''
Minor and conforming changes to other Regulatory Guides and
standard review plan sections as a result of changes in the nonseismic
criteria are also planned. If substantive changes are made during the
revisions, the applicable guides will be issued for public comment as
draft guides.
VIII. Referenced Documents
An interested person may examine or obtain copies of the documents
referenced in this rule as set out below.
Copies of NUREG-0625, NUREG-1061, NUREG-1150, NUREG-1451, NUREG-
1462, NUREG-1503, and NUREG/CR-2239 may be purchased from the
Superintendent of Documents, U.S. Government Printing Office, Mail Stop
SSOP, Washington, DC 20402-9328. Copies also are available from the
National Technical Information Service, 5285 Port Royal Road,
Springfield, VA 22161. A copy also is available for inspection and
copying for a fee in the NRC Public Document Room, 2120 L Street, NW.
(Lower Level), Washington, DC.
Copies of issued regulatory guides may be purchased from the
Government Printing Office (GPO) at the current GPO price. Information
on current GPO prices may be obtained by contacting the Superintendent
of Documents, U.S. Government Printing Office, P.O. Box 37082,
Washington, DC 20402-9328. Issued guides also may be purchased from the
National Technical Information Service on a standing order basis.
Details on this service may be obtained by writing NTIS, 5826 Port
Royal Road, Springfield, VA 22161.
SECY 79-300, SECY 90-016, SECY 93-087, and WASH-1400 are available
for inspection and copying for a fee at the NRC Public Document Room,
2120 L Street, NW. (Lower Level), Washington, DC.
IX. Summary of Comments on the Proposed Regulations
A. Reactor Siting Criteria (Nonseismic)
Eight organizations or individuals commented on the nonseismic
aspects of the second proposed revision. The first proposed revision
issued for comment in October 20, 1992, (57 FR 47802) elicited strong
comments in regard to proposed numerical values of population density
and a minimum distance to the exclusion area boundary (EAB) in the
rule. The second proposed revision (October 17, 1994; 59 FR 52255)
would delete these from the rule by providing guidance on population
density in a Regulatory Guide and determining the distance to the EAB
and LPZ by use of source term and dose
[[Page 65168]]
calculations. The rule would contain basic site criteria, without any
numerical values.
Several commentors representing the nuclear industry and
international nuclear organizations stated that the second proposed
revision was a significant improvement over the first proposed
revision, while the only public interest group commented that the NRC
had retreated from decoupling siting and design in response to the
comments of foreign entities.
Most comments on the second proposed revision centered on the use
of total effective dose equivalent (TEDE), the proposed single
numerical dose acceptance criterion of 25 rem TEDE, the evaluation of
the maximum dose in any two-hour period, and the question of whether an
organ capping dose should be adopted.
Virtually all commenters supported the concept of TEDE and its use.
However, there were differing views on the proposed numerical dose of
25 rem and the proposed use of the maximum two-hour period to evaluate
the dose. Virtually all industry commenters felt that the proposed
numerical value of 25 rem TEDE was too low and that it represented a
``ratchet'' since the use of the current dose criteria plus organ
weighting factors would suggest a value of 34 rem TEDE. In addition,
all industry commenters believed the ``sliding'' two-hour window for
dose evaluation to be confusing, illogical and inappropriate. They
favored a rule that was based upon a two hour period after the onset of
fission product release, similar in concept to the existing rule. All
industry commenters opposed the use of an organ capping dose. The only
public interest group that commented did not object to the use of TEDE,
favored the proposed dose value of 25 rem, and supported an organ
capping dose.
B. Seismic and Earthquake Engineering Criteria
Seven letters were received addressing either the regulations or
both the regulations and the draft guidance documents identified in
Section VI (except DG-4003). An additional five letters were received
addressing only the guidance documents, for a total of twelve comment
letters. A document, ``Resolution of Public Comments on the Proposed
Seismic and Earthquake Engineering Criteria for Nuclear Power Plants,''
is available explaining the NRC's disposition of the comments received
on the regulations. A copy of this document has been placed in the NRC
Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC.
Single copies are available from Dr. Andrew J. Murphy, Office of
Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, telephone (301) 415-6010. A second document,
``Resolution of Public Comments on Draft Regulatory Guides and Standard
Review Plan Sections Pertaining to the Proposed Seismic and Earthquake
Engineering Criteria for Nuclear Power Plants,'' will explain the NRC's
disposition of the comments received on the guidance documents. The
Federal Register notice announcing the avaliability of the guidance
documents will also discuss how to obtain copies of the comment
resolution document.
A summary of the major comments on the proposed regulations
follows:
Section III, Genesis (Application)
Comment: The Department of Energy (Office of Civilian Radioactive
Waste Management), requests an explicit statement on whether or not
Sec. 100.23 applies to the Mined Geologic Disposal System (MGDS) and a
Monitored Retrievable Storage (MRS) facility. The NRC has noted in
NUREG-1451, ``Staff Technical Position on Investigations to Identify
Fault Displacement Hazards and Seismic Hazards at a Geologic
Respository,'' that Appendix A to 10 CFR Part 100 does not apply to a
geologic repository. NUREG-1451 also notes that the contemplated
revisions to Part 100 would also not be applicable to a geologic
repository. Section 72.102(b) requires that, for an MRS located west of
the Rocky Mountain front or in areas of known potential seismic
activity in the east, the seismicity be evaluated by the techniques of
Appendix A to 10 CFR Part 100.
Response: Although Appendix A to 10 CFR Part 100 is titled
``Seismic and Geologic Siting Criteria for Nuclear Power Plants,'' it
is also referenced in two other parts of the regulation. They are (1)
Part 40, ``Domestic Licensing of Source Material,'' Appendix A,
``Criteria Relating to the Operation of Uranium Mills and the
Disposition of Tailings or Waste Produced by the Extraction or
Concentration of Source Material from Ores Processed Primarily for
Their Source Material Content,'' Section I, Criterion 4(e), and (2)
Part 72, ``Licensing Requirements for the Independent Storage of Spent
Nuclear Fuel and High-Level Radioactive Waste,'' Paragraphs (a)(2), (b)
and (f)(1) of Sec. 72.102.
The referenced applicability of Sec. 100.23 to other than power
reactors, if considered appropriate by the NRC, would be a separate
rulemaking. That rulemaking would clearly state the applicability of
Sec. 100.23 to an MRS or other facility. In addition, NUREG-1451 will
remain the NRC staff technical position on seismic siting issues
pertaining to an MGDS until it is superseded through a rulemaking,
revision of NUREG-1451, or other appropriate mechanism.
Section V(B)(5), ``Value of the Operating Basis Earthquake Ground
Motion (OBE) and Required OBE Analysis.''
Comment: One commenter, ABB Combustion Engineering Nuclear Systems,
specifically stated that they agree with the NRC's proposal to not
require explicit design analysis of the OBE if its peak acceleration is
less than one-third of the Safe Shutdown Earthquake Ground Motion
(SSE). The only negative comments, from G.C. Slagis Associates, stated
that the proposed rule in the area of required OBE analysis is not
sound, not technically justified, and not appropriate for the design of
pressure-retaining components. The following are specific comments
(limited to the design of pressure-retaining components to the ASME
Boiler and Pressure Vessel Section III rules) that pertain to the
supplemental information to the proposed regulations, item V(B)(5),
``Value of the Operating Basis Earthquake Ground Motion (OBE) and
Required OBE Analysis.''
(1) Comment: Disagrees with the statement in SECY-79-300 that
design for a single limiting event and inspection and evaluation for
earthquakes in excess of some specified limit may be the most sound
regulatory approach. It is not feasible to inspect for cyclic damage to
all the pressure-retaining components. Visually inspecting for
permanent deformation, or leakage, or failed component supports is
certainly not adequate to determine cyclic damage.
Response: The NRC agrees. Postearthquake inspection and evaluation
guidance is described in Regulatory Guide 1.167 (Draft was DG-1035),
``Restart of a Nuclear Power Plant Shut Down by an Seismic Event.'' The
guidance is not limited to visual inspections; it includes inspections,
tests, and analyses including fatigue analysis.
(2) Comment: Disagrees with the NRC statement in SECY-090-016 that
the OBE should not control design. There is a problem with the present
requirements. Requiring design for five OBE events at one-half SSE is
unrealistic for most (all?) sites and requires an excessive and
unnecessary number of seismic supports. The solution is to properly
define the OBE
[[Page 65169]]
magnitude and the number of events expected during the life of the
plant and to require design for that loading. OBE may or may not
control the design. But you cannot assume, before you have the
seismicity defined and before you have a component design, that OBE
will not govern the design.
Response: The NRC has concluded that design requirements based on
an estimated OBE magnitude at the plant site and the number of events
expected during the plant life will lead to low design values that will
not control the design, thus resulting in unnecessary analyses.
(3) Comment: It is not technically justified to assume that Section
III components will remain within applicable stress limits (Level B
limits) at one-third the SSE. The Section III acceptance criteria for
Level D (for an SSE) is completely different than that for Level B (for
an OBE). The Level D criteria is based on surviving the extremely-low
probability SSE load. Gross structural deformations are possible, and
it is expected that the component will have to be replaced. Cyclic
effects are not considered. The cyclic effects of the repeated
earthquakes have to be considered in the design of the component to
ensure pressure boundary integrity throughout the life of the
component, especially if the SSE can occur after the lower level
earthquakes.
Response: In SECY-93-087, Issue I.M, ``Elimination of Operating-
Basis Earthquake,'' the NRC recognizes that a designer of piping
systems considers the effects of primary and secondary stresses and
evaluates fatigue caused by repeated cycles of loading. Primary
stresses are induced by the inertial effects of vibratory motion. The
relative motion of anchor points induces secondary stresses. The
repeating seismic stress cycles induce cyclic effects (fatigue).
However, after reviewing these aspects, the NRC concludes that, for
primary stresses, if the OBE is established at one-third the SSE, the
SSE load combinations control the piping design when the earthquake
contribution dominates the load combination. Therefore, the NRC
concludes that eliminating the OBE piping stress load combination for
primary stresses in piping systems will not significantly reduce
existing safety margins.
Eliminating the OBE will, however, directly affect the current
methods used to evaluate the adequacy of cyclic and secondary stress
effects in the piping design. Eliminating the OBE from the load
combination could cause uncertainty in evaluating the cyclic (fatigue)
effects of earthquake-induced motions in piping systems and the
relative motion effects of piping anchored to equipment and structures
at various elevations because both of these effects are currently
evaluated only for OBE loadings. Accordingly, to account for earthquake
cycles in the fatigue analysis of piping systems, the staff proposes to
develop guidelines for selecting a number of SSE cycles at a fraction
of the peak amplitude of the SSE. These guidelines will provide a level
of fatigue design for the piping equivalent to that currently provided
in Standard Review Plan Section 3.9.2.
Positions pertaining to the elimination of the OBE were proposed in
SECY-93-087. Commission approval is documented in a memorandum from
Samuel J. Chilk to James M. Taylor, Subject: SECY-93-087--Policy,
Technical and Licensing Issues Pertaining to Evolutionary and Advanced
Light-Water Reactor (ALWR) Designs, dated July 21, 1993.
(4) Comment: There is one major flaw in the ``SSE only'' design
approach. The equipment designed for SSE is limited to the equipment
necessary to assure the integrity of the reactor coolant pressure
boundary, to shutdown the reactor, and to prevent or mitigate accident
consequences. The equipment designed for SSE is only part of the
equipment ``necessary for continued operation without undue risk to the
health and safety of the public.'' Hence, by this rule, it is possible
that some equipment necessary for continued operation will not be
designed for SSE or OBE effects.
Response: The NRC does not agree that the design approach is
flawed. It is not possible that some equipment necessary for continued
safe operation will not be designed for SSE or OBE effects. General
Design Criterion 2, ``Design Bases for Protection Against Natural
Phenomena,'' of Appendix A, ``General Design Criteria for Nuclear Power
Plants,'' to 10 CFR Part 50 requires that nuclear power plant
structures, systems, and components important to safety be designed to
withstand the effects of earthquakes without loss of capability to
perform their safety functions. The criteria in Appendix S to 10 CFR
Part 50 implement General Design Criterion 2 insofar as it requires
structures, systems, and components important to safety to withstand
the effects of earthquakes. Regulatory Guide 1.29, ``Seismic Design
Classification,'' describes a method acceptable to the NRC for
identifying and classifying those features of light-water-cooled
nuclear power plants that should be designed to withstand the effects
of the SSE. Currently, components which are designed for OBE only
include components such as waste holdup tanks. As noted in Section VII,
Future Regulatory Actions, regulatory guides related to these
components will be revised to provide alternative design requirements.
10 CFR 100.23
The Nuclear Energy Institute (NEI) congratulated the NRC staff for
carefully considering and responding to the voluminous and complex
comments that were provided on the earlier proposed rulemaking package
(October 20, 1992; 57 FR 47802) and considered that the seismic portion
of the proposed rulemaking package is nearing maturity and with the
inclusion of industry's comments (which were principally on the
guidance documents), has the potential to satisfy the objectives of
predictable licensing and stable regulations.
Both NEI and Westinghouse Electric Corporation support the
regulation format, that is, prescriptive guidance is located in
regulatory guides or standard review plan sections and not the
regulation.
NEI and Westinghouse Electric Corporation support the removal of
the requirement from the first proposed rulemaking (57 FR 47802) that
both deterministic and probabilistic evaluations must be conducted to
determine site suitability and seismic design requirements for the
site. [Note: the commenters do not agree with the NRC staff's
deterministic check of the seismic sources and parameters used in the
LLNL and EPRI probabilistic seismic hazard analyses (Regulatory Guide
1.165, draft was DG-1032). Also, they do not support the NRC staff's
deterministic check of the applicants submittal (SRP Section 2.5.2).
These items are addressed in the document pertaining to comment
resolution of the draft regulatory guides and standard review plan
sections.]
Comment: NEI, Westinghouse Electric Corporation, and Yankee Atomic
Electric Corporation recommend that the regulation should state that
for existing sites east of the Rocky Mountain Front (east of
approximately 105 deg. west longitude), a 0.3g standardized design
level is acceptable at these sites given confirmatory foundations
evaluations [Regulatory Guide 1.132, but not the geologic, geophysical,
seismological investigations in Regulatory Guide 1.165].
Response: The NRC has determined that the use of a spectral shape
anchored to 0.3g peak ground acceleration as a standardized design
level would be
[[Page 65170]]
appropriate for existing central and eastern U.S. sites based on the
current state of knowledge. However, as new information becomes
available it may not be appropriate for future licensing decisions.
Pertinent information such as that described in Regulatory Guide 1.165
(Draft was DG-1032) is needed to make that assessment. Therefore, it is
not appropriate to codify the request.
Comment: NEI recommended a rewording of Paragraph (a),
Applicability. Although unlikely, an applicant for an operating license
already holding a construction permit may elect to apply the amended
methodology and criteria in Subpart B to Part 100.
Response: The NRC will address this request on a case-by-case basis
rather than through a generic change to the regulations. This situation
pertains to a limited number of facilities in various stages of
construction. Some of the issues that must be addressed by the
applicant and NRC during the operating license review include
differences between the design bases derived from the current and
amended regulations (Appendix A to Part 100 and Sec. 100.23,
respectively), and earthquake engineering criteria such as, OBE design
requirements and OBE shutdown requirements.
Appendix S to 10 CFR Part 50
Support for the NRC position pertaining to the elimination of the
Operating Basis Earthquake Ground Motion (OBE) response analyses has
been documented in various NRC publications such as SECY-79-300, SECY-
90-016, SECY-93-087, and NUREG-1061. The final safety evaluation
reports related to the certification of the System 80+ and the Advanced
Boiling Water Reactor design (NUREG-1462 and NUREG-1503, respectively)
have already adopted the single earthquake design philosophy. In
addition, similar activities are being done in foreign countries, for
instance, Germany. (Additional discussion is provided in Section
V(B)(5) of this rule).
Comment: The American Society of Civil Engineers (ASCE) recommended
that the seismic design and engineering criteria of ASCE Standard 4,
``Seismic Analysis of Safety-Related Nuclear Structures and Commentary
on Standard for Seismic Analysis of Safety-Related Nuclear
Structures,'' be incorporated by reference into Appendix S to 10 CFR
Part 50.
Response: The Commission has determined that new regulations will
be more streamlined and contain only basic requirements with guidance
being provided in regulatory guides and, to some extent, in standard
review plan sections. Both the NRC and industry have experienced
difficulties in applying prescriptive regulations such as Appendix A to
10 CFR Part 100 because they inhibit the use of needed latitude in
judgment. Therefore, it is common NRC practice not to reference
publications such as ASCE Standard 4 (an analysis, not design standard)
in its regulations. Rather, publications such as ASCE Standard 4 are
cited in regulatory guides and standard review plan sections. ASCE
Standard 4 is cited in the 1989 revision of Standard Review Plan
Sections 3.7.1, 3.7.2, and 3.7.3.
Comment: The Department of Energy stated that the required
consideration of aftershocks in Paragraph IV(B), Surface Deformation,
is confusing and recommended that it be deleted.
Response: The NRC agrees. The reference to aftershocks in Paragraph
IV(b) has been deleted. Paragraphs VI(a), Safe Shutdown Earthquake, and
VI(B)(3) of Appendix A to Part 100 contain the phrase ``including
aftershocks.'' The ``including aftershocks'' phrase was removed from
the Safe Shutdown Earthquake Ground Motion requirements in the proposed
regulation. The recommended change will make Paragraphs IV(a)(1),
``Safe Shutdown Earthquake Ground Motion,'' and IV(b), ``Surface
Deformation, of Appendix S to 10 CFR Part 50 consistent.
X. Small Business Regulatory Enforcement Fairness Act
In accordance with the Small Business Regulatory Enforcement
Fairness Act of 1996 the NRC has determined that this action is not a
major rule and has verified this determination with the Office of
Information and Regulatory Affairs of OMB.
XI. Finding of No Significant Environmental Impact: Availability
The Commission has determined under the National Environmental
Policy Act of 1969, as amended, and the Commission's regulations in
Subpart A of 10 CFR Part 51, that this regulation is not a major
Federal action significantly affecting the quality of the human
environment and therefore an environmental impact statement is not
required.
The revisions associated with the reactor siting criteria in 10 CFR
Part 100 and the relocation of the plant design requirements from 10
CFR Part 100 to 10 CFR Part 50 have been evaluated against the current
requirements. The Commission has concluded that relocating the
requirement for a dose calculation to Part 50 and adding more specific
site criteria to Part 100 does not decrease the protection of public
health and safety over the current regulations. The amendments do not
affect nonradiological plant effluents and have no other environmental
impact.
The addition of Sec. 100.23 to 10 CFR Part 100, and the addition of
Appendix S to 10 CFR Part 50, will not change the radiological
environmental impact offsite. Onsite occupational radiation exposure
associated with inspection and maintenance will not change. These
activities are principally associated with baseline inspections of
structures, equipment, and piping, and with maintenance of seismic
instrumentation. Baseline inspections are needed to differentiate
between pre-existing conditions at the nuclear power plant and
earthquake related damage. The structures, equipment and piping
selected for these inspections are those routinely examined by plant
operators during normal plant walkdowns and inspections. Routine
maintenance of seismic instrumentation ensures its operability during
earthquakes. The location of the seismic instrumentation is similar to
that in the existing nuclear power plants. The amendments do not affect
nonradiological plant effluents and have no other environmental impact.
The environmental assessment and finding of no significant impact
on which this determination is based are available for inspection at
the NRC Public Document Room, 2120 L Street NW. (Lower Level),
Washington, DC. Single copies of the environmental assessment and
finding of no significant impact are available from Dr. Andrew J.
Murphy, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, telephone (301) 415-6010.
XII. Paperwork Reduction Act Statement
This final rule amends information collection requirements that are
subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et
seq.). These requirements were approved by the Office of Management and
Budget, approval numbers 3150-0011 and 3150-0093.
The public reporting burden for this collection of information is
estimated to average 800,000 hours per response, including the time for
reviewing instructions, searching existing data sources, gathering and
maintaining the data needed, and completing and reviewing the
collection of information. Send comments on any aspect of this
collection of information, including
[[Page 65171]]
suggestions for reducing the burden, to the Information and Records
Management Branch (T-6 F33), U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, or by Internet electronic mail to
[email protected]; and to the Desk Officer, Office of Information and
Regulatory Affairs, NEOB-10202 (3150-0011 and 3150-0093), Office of
Management and Budget, Washington, DC 20503.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a collection of information unless it displays a currently
valid OMB control number.
XIII. Regulatory Analysis
The Commission has prepared a regulatory analysis on this
regulation. The analysis examines the costs and benefits of the
alternatives considered by the Commission. Interested persons may
examine a copy of the regulatory analysis at the NRC Public Document
Room, 2120 L Street NW. (Lower Level), Washington, DC. Single copies of
the analysis are available from Dr. Andrew J. Murphy, Office of Nuclear
Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, telephone (301) 415-6010.
XIV. Regulatory Flexibility Certification
As required by the Regulatory Flexibility Act of 1980, 5 U.S.C.
605(b), the Commission certifies that this regulation does not have a
significant economic impact on a substantial number of small entities.
This regulation affects only the licensing and operation of nuclear
power plants. The companies that own these plants do not fall within
the definition of ``small entities'' set forth in the Regulatory
Flexibility Act or the size standards established by the NRC (April 11,
1995; 60 FR 18344).
XV. Backfit Analysis
The NRC has determined that the backfit rule, 10 CFR 50.109, does
not apply to this regulation, and, therefore, a backfit analysis is not
required for this regulation because these amendments do not involve
any provisions that would impose backfits as defined in 10 CFR
50.109(a)(1). The regulation would apply only to applicants for future
nuclear power plant construction permits, preliminary design approval,
final design approval, manufacturing licenses, early site reviews,
operating licenses, and combined operating licenses.
List of Subjects
10 CFR Part 21
Nuclear power plants and reactors, Penalties, Radiation protection,
Reporting and recordkeeping requirements.
10 CFR Part 50
Antitrust, Classified information, Criminal penalties, Fire
protection, Intergovernmental relations, Nuclear power plants and
reactors, Radiation protection, Reactor siting criteria, Reporting and
recordkeeping requirements.
10 CFR Part 52
Administrative practice and procedure, Antitrust, Backfitting,
Combined license, Early site permit, Emergency planning, Fees,
Inspection, Limited work authorization, Nuclear power plants and
reactors, Probabilistic risk assessment, Prototype, Reactor siting
criteria, Redress of site, Reporting and recordkeeping requirements,
Standard design, Standard design certification.
10 CFR Part 54
Administrative practice and procedure, Age-related degradation,
Backfitting, Classified information, Criminal penalties, Environmental,
Nuclear power plants and reactors, Reporting and recordkeeping
requirements.
10 CFR Part 100
Nuclear power plants and reactors, Reactor siting criteria.
For the reasons set out in the preamble and under the authority of
the Atomic Energy Act of 1954, as amended, the Energy Reorganization
Act of 1974, as amended, and 5 U.S.C. 552 and 553, the NRC is adopting
the following amendments to 10 CFR Parts 21, 50, 52, 54, and 100:
PART 21--REPORTING OF DEFECTS AND NONCOMPLIANCE
1. The authority citation for Part 21 continues to read as follows:
Authority: Sec. 161, 68 Stat. 948, as amended, sec. 234, 83
Stat. 444, as amended, sec. 1701, 106 Stat. 2951, 2953 (42 U.S.C.
2201, 2282, 2297f); secs. 201, as amended, 206, 88 Stat. 1242, as
amended, 1246 (42 U.S.C. 5841, 5846).
Section 21.2 also issued under secs. 135, 141, Pub. L. 97-425,
96 Stat. 2232, 2241 (42 U.S.C. 10155, 10161).
2. In Sec. 21.3, the definition for Basic component (1)(i)(C) is
revised to read as follows:
Sec. 21.3 Definitions.
* * * * *
Basic component. (1)(i) * * *
(C) The capability to prevent or mitigate the consequences of
accidents which could result in potential offsite exposures comparable
to those referred to in Sec. 50.34(a)(1) or Sec. 100.11 of this
chapter, as applicable.
* * * * *
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
3. The authority citation for Part 50 continues to read as follows:
Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68
Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234,
83 Stat. 1244, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201,
2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88
Stat. 1242, as amended, 1244, 1246, (42 U.S.C. 5841, 5842, 5846).
Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat.
2951 (42 U.S.C. 5851). Section 50.10 also issued under secs. 101,
185, 68 Stat. 955 as amended (42 U.S.C. 2131, 2235), sec. 102, Pub.
L. 91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 50.54(dd)
and 50.103 also issued under sec. 108, 68 Stat. 939, as amended (42
U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56 also issued
under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a,
50.55a and Appendix Q also issued under sec. 102, Pub. L. 91-190, 83
Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued
under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 50.58,
50.91 and 50.92 also issued under Pub. L. 97-415, 96 Stat. 2073 (42
U.S.C. 2239). Section 50.78 also issued under sec. 122, 68 Stat. 939
(42 U.S.C. 2152). Sections 50.80-50.81 also issued under sec. 184,
68 Stat. 954, as amended (42 U.S.C. 2234). Appendix F also issued
under sec. 187, 68 Stat. 955 (42 U.S.C. 2237).
4. Section 50.2 is amended by adding in alphabetical order the
definitions for Committed dose equivalent, Committed effective dose
equivalent, Deep-dose equivalent, Exclusion area, Low population zone,
Safety-related structures, systems, and components and Total effective
dose equivalent, and revising the definition for Basic component
(1)(iii) to read as follows:
Sec. 50.2 Definitions.
* * * * *
Basic component * * *
(1) * * *
(iii) The capability to prevent or mitigate the consequences of
accidents which could result in potential offsite exposures comparable
to those referred to in Sec. 50.34(a)(1) or Sec. 100.11 of this
chapter, as applicable.
* * * * *
Committed dose equivalent means the dose equivalent to organs or
tissues of
[[Page 65172]]
reference that will be received from an intake of radioactive material
by an individual during the 50-year period following the intake.
Committed effective dose equivalent is the sum of the products of
the weighting factors applicable to each of the body organs or tissues
that are irradiated and the committed dose equivalent to these organs
or tissues.
* * * * *
Deep-dose equivalent, which applies to external whole-body
exposure, is the dose equivalent at a tissue depth of 1 cm (1000mg/
cm2).
* * * * *
Exclusion area means that area surrounding the reactor, in which
the reactor licensee has the authority to determine all activities
including exclusion or removal of personnel and property from the area.
This area may be traversed by a highway, railroad, or waterway,
provided these are not so close to the facility as to interfere with
normal operations of the facility and provided appropriate and
effective arrangements are made to control traffic on the highway,
railroad, or waterway, in case of emergency, to protect the public
health and safety. Residence within the exclusion area shall normally
be prohibited. In any event, residents shall be subject to ready
removal in case of necessity. Activities unrelated to operation of the
reactor may be permitted in an exclusion area under appropriate
limitations, provided that no significant hazards to the public health
and safety will result.
* * * * *
Low population zone means the area immediately surrounding the
exclusion area which contains residents, the total number and density
of which are such that there is a reasonable probability that
appropriate protective measures could be taken in their behalf in the
event of a serious accident. These guides do not specify a permissible
population density or total population within this zone because the
situation may vary from case to case. Whether a specific number of
people can, for example, be evacuated from a specific area, or
instructed to take shelter, on a timely basis will depend on many
factors such as location, number and size of highways, scope and extent
of advance planning, and actual distribution of residents within the
area.
* * * * *
Safety-related structures, systems, and components means those
structures, systems, and components that are relied on to remain
functional during and following design basis (postulated) events to
assure:
(1) The integrity of the reactor coolant pressure boundary;
(2) The capability to shut down the reactor and maintain it in a
safe shutdown condition; and
(3) The capability to prevent or mitigate the consequences of
accidents which could result in potential offsite exposures comparable
to the applicable guideline exposures set forth in Sec. 50.34(a)(1) or
Sec. 100.11 of this chapter, as applicable.
* * * * *
Total effective dose equivalent (TEDE) means the sum of the deep-
dose equivalent (for external exposures) and the committed effective
dose equivalent (for internal exposures).
* * * * *
5. In Sec. 50.8, paragraph (b) is revised to read as follows:
Sec. 50.8 Information collection requirements: OMB approval.
* * * * *
(b) The approved information collection requirements contained in
this part appear in Secs. 50.30, 50.33, 50.33a, 50.34, 50.34a, 50.35,
50.36, 50.36a, 50.36b, 50.44, 50.46, 50.47, 50.48, 50.49, 50.54, 50.55,
50.55a, 50.59, 50.60, 50.61, 50.62, 50.63, 50.64, 50.65, 50.66, 50.71,
50.72, 50.74, 50.75, 50.80, 50.82, 50.90, 50.91, 50.120, and Appendices
A, B, E, G, H, I, J, K, M, N, O, Q, R, and S to this part.
* * * * *
6. In Sec. 50.34, footnotes 6, 7, and 8 are redesignated as
footnotes 8, 9 and 10 and paragraph (a)(1) is revised and paragraphs
(a)(12), (b)(10), and (b)(11) are added to read as follows:
Sec. 50.34 Contents of applications; technical information.
(a) * * *
(1) Stationary power reactor applicants for a construction permit
pursuant to this part, or a design certification or combined license
pursuant to part 52 of this chapter who apply on or after January 10,
1997, shall comply with paragraph (a)(1)(ii) of this section. All other
applicants for a construction permit pursuant to this part or a design
certification or combined license pursuant to part 52 of this chapter,
shall comply with paragraph (a)(1)(i) of this section.
(i) A description and safety assessment of the site on which the
facility is to be located, with appropriate attention to features
affecting facility design. Special attention should be directed to the
site evaluation factors identified in part 100 of this chapter. The
assessment must contain an analysis and evaluation of the major
structures, systems and components of the facility which bear
significantly on the acceptability of the site under the site
evaluation factors identified in part 100 of this chapter, assuming
that the facility will be operated at the ultimate power level which is
contemplated by the applicant. With respect to operation at the
projected initial power level, the applicant is required to submit
information prescribed in paragraphs (a)(2) through (a)(8) of this
section, as well as the information required by this paragraph, in
support of the application for a construction permit, or a design
approval.
(ii) A description and safety assessment of the site and a safety
assessment of the facility. It is expected that reactors will reflect
through their design, construction and operation an extremely low
probability for accidents that could result in the release of
significant quantities of radioactive fission products. The following
power reactor design characteristics and proposed operation will be
taken into consideration by the Commission:
(A) Intended use of the reactor including the proposed maximum
power level and the nature and inventory of contained radioactive
materials;
(B) The extent to which generally accepted engineering standards
are applied to the design of the reactor;
(C) The extent to which the reactor incorporates unique, unusual or
enhanced safety features having a significant bearing on the
probability or consequences of accidental release of radioactive
materials;
(D) The safety features that are to be engineered into the facility
and those barriers that must be breached as a result of an accident
before a release of radioactive material to the environment can occur.
Special attention must be directed to plant design features intended to
mitigate the radiological consequences of accidents. In performing this
assessment, an applicant shall assume a fission product release 6
from the core into the containment assuming that the facility is
operated at the ultimate power level contemplated. The applicant shall
perform an evaluation and analysis of the postulated fission product
release, using the expected demonstrable containment leak rate and any
fission
[[Page 65173]]
product cleanup systems intended to mitigate the consequences of the
accidents, together with applicable site characteristics, including
site meteorology, to evaluate the offsite radiological consequences.
Site characteristics must comply with part 100 of this chapter. The
evaluation must determine that:
---------------------------------------------------------------------------
\6\ The fission product release assumed for this evaluation
should be based upon a major accident, hypothesized for purposes of
site analysis or postulated from considerations of possible
accidental events. Such accidents have generally been assumed to
result in substantial meltdown of the core with subsequent release
into the containment of appreciable quantities of fission products.
---------------------------------------------------------------------------
(1) An individual located at any point on the boundary of the
exclusion area for any 2 hour period following the onset of the
postulated fission product release, would not receive a radiation dose
in excess of 25 rem 7 total effective dose equivalent (TEDE).
---------------------------------------------------------------------------
\7\ A whole body dose of 25 rem has been stated to correspond
numerically to the once in a lifetime accidental or emergency dose
for radiation workers which, according to NCRP recommendations at
the time could be disregarded in the determination of their
radiation exposure status (see NBS Handbook 69 dated June 5, 1959).
However, its use is not intended to imply that this number
constitutes an acceptable limit for an emergency dose to the public
under accident conditions. Rather, this dose value has been set
forth in this section as a reference value, which can be used in the
evaluation of plant design features with respect to postulated
reactor accidents, in order to assure that such designs provide
assurance of low risk of public exposure to radiation, in the event
of such accidents.
---------------------------------------------------------------------------
(2) An individual located at any point on the outer boundary of the
low population zone, who is exposed to the radioactive cloud resulting
from the postulated fission product release (during the entire period
of its passage) would not receive a radiation dose in excess of 25 rem
total effective dose equivalent (TEDE);
(E) With respect to operation at the projected initial power level,
the applicant is required to submit information prescribed in
paragraphs (a)(2) through (a)(8) of this section, as well as the
information required by this paragraph (a)(1)(i), in support of the
application for a construction permit, or a design approval.
* * * * *
(12) On or after January 10, 1997, stationary power reactor
applicants who apply for a construction permit pursuant to this part,
or a design certification or combined license pursuant to part 52 of
this chapter, as partial conformance to General Design Criterion 2 of
Appendix A to this part, shall comply with the earthquake engineering
criteria in Appendix S to this part.
(b) * * *
(10) On or after January 10, 1997, stationary power reactor
applicants who apply for an operating license pursuant to this part, or
a design certification or combined license pursuant to part 52 of this
chapter, as partial conformance to General Design Criterion 2 of
Appendix A to this part, shall comply with the earthquake engineering
criteria of Appendix S to this part. However, for those operating
license applicants and holders whose construction permit was issued
prior to January 10, 1997, the earthquake engineering criteria in
Section VI of Appendix A to part 100 of this chapter continues to
apply.
(11) On or after January 10, 1997, stationary power reactor
applicants who apply for an operating license pursuant to this part, or
a combined license pursuant to part 52 of this chapter, shall provide a
description and safety assessment of the site and of the facility as in
Sec. 50.34(a)(1)(ii) of this part. However, for either an operating
license applicant or holder whose construction permit was issued prior
to January 10, 1997, the reactor site criteria in part 100 of this
chapter and the seismic and geologic siting criteria in Appendix A to
part 100 of this chapter continues to apply.
* * * * *
7. In Sec. 50.49, paragraph (b)(1) is revised to read as follows:
Sec. 50.49 Environmental qualification of electric equipment
important to safety for nuclear power plants.
* * * * *
(b) * * *
(1) Safety-related electric equipment.3
---------------------------------------------------------------------------
\3\ Safety-related electric equipment is referred to as ``Class
1E'' equipment in IEEE 323-1974. Copies of this standard may be
obtained from the Institute of Electrical and Electronics Engineers,
Inc., 345 East 47th Street, New York, NY 10017.
---------------------------------------------------------------------------
(i) This equipment is that relied upon to remain functional during
and following design basis events to ensure--
(A) The integrity of the reactor coolant pressure boundary;
(B) The capability to shut down the reactor and maintain it in a
safe shutdown condition; and
(C) The capability to prevent or mitigate the consequences of
accidents that could result in potential offsite exposures comparable
to the guidelines in Sec. 50.34(a)(1) or Sec. 100.11 of this chapter,
as applicable.
(ii) Design basis events are defined as conditions of normal
operation, including anticipated operational occurrences, design basis
accidents, external events, and natural phenomena for which the plant
must be designed to ensure functions (b)(1)(i) (A) through (C) of this
section.
* * * * *
8. In Sec. 50.54, paragraph (ff) is added to read as follows:
Sec. 50.54 Conditions of licenses.
* * * * *
(ff) For licensees of nuclear power plants that have implemented
the earthquake engineering criteria in Appendix S to this part, plant
shutdown is required as provided in Paragraph IV(a)(3) of Appendix S to
this part. Prior to resuming operations, the licensee shall demonstrate
to the Commission that no functional damage has occurred to those
features necessary for continued operation without undue risk to the
health and safety of the public and the licensing basis is maintained.
9. In Sec. 50.65, paragraph (b)(1) is revised to read as follows:
Sec. 50.65 Requirements for monitoring the effectiveness of
maintenance at nuclear power plants
* * * * *
(b) * * *
(1) Safety related structures, systems, or components that are
relied upon to remain functional during and following design basis
events to ensure the integrity of the reactor coolant pressure
boundary, the capability to shut down the reactor and maintain it in a
safe shutdown condition, and the capability to prevent or mitigate the
consequences of accidents that could result in potential offsite
exposure comparable to the guidelines in Sec. 50.34(a)(1) or
Sec. 100.11 of this chapter, as applicable.
* * * * *
10. Appendix S to Part 50 is added to read as follows:
Appendix S to Part 50--Earthquake Engineering Criteria for Nuclear
Power Plants
General Information
This appendix applies to applicants for a design certification
or combined license pursuant to part 52 of this chapter or a
construction permit or operating license pursuant to part 50 of this
chapter on or after January 10, 1997. However, for either an
operating license applicant or holder whose construction permit was
issued prior to January 10, 1997, the earthquake engineering
criteria in Section VI of Appendix A to 10 CFR part 100 continues to
apply.
I. Introduction
(a) Each applicant for a construction permit, operating license,
design certification, or combined license is required by Sec. 50.34
(a)(12), (b)(10), and General Design Criterion 2 of Appendix A to
this part to design nuclear power plant structures, systems, and
components important to safety to withstand the effects of natural
phenomena, such as earthquakes, without loss of capability to
perform their safety functions. Also, as specified in
Sec. 50.54(ff), nuclear power plants that have implemented the
earthquake engineering criteria described herein must shut down if
the criteria in Paragraph IV(a)(3) of this appendix are exceeded.
[[Page 65174]]
(b) These criteria implement General Design Criterion 2 insofar
as it requires structures, systems, and components important to
safety to withstand the effects of earthquakes.
II. Scope
The evaluations described in this appendix are within the scope
of investigations permitted by Sec. 50.10(c)(1).
III. Definitions
As used in these criteria:
Combined license means a combined construction permit and
operating license with conditions for a nuclear power facility
issued pursuant to Subpart C of Part 52 of this chapter.
Design Certification means a Commission approval, issued
pursuant to Subpart B of Part 52 of this chapter, of a standard
design for a nuclear power facility. A design so approved may be
referred to as a ``certified standard design.''
The Operating Basis Earthquake Ground Motion (OBE) is the
vibratory ground motion for which those features of the nuclear
power plant necessary for continued operation without undue risk to
the health and safety of the public will remain functional. The
Operating Basis Earthquake Ground Motion is only associated with
plant shutdown and inspection unless specifically selected by the
applicant as a design input.
A response spectrum is a plot of the maximum responses
(acceleration, velocity, or displacement) of idealized single-
degree-of-freedom oscillators as a function of the natural
frequencies of the oscillators for a given damping value. The
response spectrum is calculated for a specified vibratory motion
input at the oscillators' supports.
The Safe Shutdown Earthquake Ground Motion (SSE) is the
vibratory ground motion for which certain structures, systems, and
components must be designed to remain functional.
The structures, systems, and components required to withstand
the effects of the Safe Shutdown Earthquake Ground Motion or surface
deformation are those necessary to assure:
(1) The integrity of the reactor coolant pressure boundary;
(2) The capability to shut down the reactor and maintain it in a
safe shutdown condition; or
(3) The capability to prevent or mitigate the consequences of
accidents that could result in potential offsite exposures
comparable to the guideline exposures of Sec. 50.34(a)(1).
Surface deformation is distortion of geologic strata at or near
the ground surface by the processes of folding or faulting as a
result of various earth forces. Tectonic surface deformation is
associated with earthquake processes.
IV. Application To Engineering Design
The following are pursuant to the seismic and geologic design
basis requirements of Sec. 100.23 of this chapter:
(a) Vibratory Ground Motion.
(1) Safe Shutdown Earthquake Ground Motion.
(i) The Safe Shutdown Earthquake Ground Motion must be
characterized by free-field ground motion response spectra at the
free ground surface. In view of the limited data available on
vibratory ground motions of strong earthquakes, it usually will be
appropriate that the design response spectra be smoothed spectra.
The horizontal component of the Safe Shutdown Earthquake Ground
Motion in the free-field at the foundation level of the structures
must be an appropriate response spectrum with a peak ground
acceleration of at least 0.1g.
(ii) The nuclear power plant must be designed so that, if the
Safe Shutdown Earthquake Ground Motion occurs, certain structures,
systems, and components will remain functional and within applicable
stress, strain, and deformation limits. In addition to seismic
loads, applicable concurrent normal operating, functional, and
accident-induced loads must be taken into account in the design of
these safety-related structures, systems, and components. The design
of the nuclear power plant must also take into account the possible
effects of the Safe Shutdown Earthquake Ground Motion on the
facility foundations by ground disruption, such as fissuring,
lateral spreads, differential settlement, liquefaction, and
landsliding, as required in Sec. 100.23 of this chapter.
(iii) The required safety functions of structures, systems, and
components must be assured during and after the vibratory ground
motion associated with the Safe Shutdown Earthquake Ground Motion
through design, testing, or qualification methods.
(iv) The evaluation must take into account soil-structure
interaction effects and the expected duration of vibratory motion.
It is permissible to design for strain limits in excess of yield
strain in some of these safety-related structures, systems, and
components during the Safe Shutdown Earthquake Ground Motion and
under the postulated concurrent loads, provided the necessary safety
functions are maintained.
(2) Operating Basis Earthquake Ground Motion.
(i) The Operating Basis Earthquake Ground Motion must be
characterized by response spectra. The value of the Operating Basis
Earthquake Ground Motion must be set to one of the following
choices:
(A) One-third or less of the Safe Shutdown Earthquake Ground
Motion design response spectra. The requirements associated with
this Operating Basis Earthquake Ground Motion in Paragraph
(a)(2)(i)(B)(I ) can be satisfied without the applicant performing
explicit response or design analyses, or
(B) A value greater than one-third of the Safe Shutdown
Earthquake Ground Motion design response spectra. Analysis and
design must be performed to demonstrate that the requirements
associated with this Operating Basis Earthquake Ground Motion in
Paragraph (a)(2)(i)(B)(I) are satisfied. The design must take into
account soil-structure interaction effects and the duration of
vibratory ground motion.
(I) When subjected to the effects of the Operating Basis
Earthquake Ground Motion in combination with normal operating loads,
all structures, systems, and components of the nuclear power plant
necessary for continued operation without undue risk to the health
and safety of the public must remain functional and within
applicable stress, strain, and deformation limits.
(3) Required Plant Shutdown. If vibratory ground motion
exceeding that of the Operating Basis Earthquake Ground Motion or if
significant plant damage occurs, the licensee must shut down the
nuclear power plant. If systems, structures, or components necessary
for the safe shutdown of the nuclear power plant are not available
after the occurrence of the Operating Basis Earthquake Ground
Motion, the licensee must consult with the Commission and must
propose a plan for the timely, safe shutdown of the nuclear power
plant. Prior to resuming operations, the licensee must demonstrate
to the Commission that no functional damage has occurred to those
features necessary for continued operation without undue risk to the
health and safety of the public and the licensing basis is
maintained.
(4) Required Seismic Instrumentation. Suitable instrumentation
must be provided so that the seismic response of nuclear power plant
features important to safety can be evaluated promptly after an
earthquake.
(b) Surface Deformation. The potential for surface deformation
must be taken into account in the design of the nuclear power plant
by providing reasonable assurance that in the event of deformation,
certain structures, systems, and components will remain functional.
In addition to surface deformation induced loads, the design of
safety features must take into account seismic loads and applicable
concurrent functional and accident-induced loads. The design
provisions for surface deformation must be based on its postulated
occurrence in any direction and azimuth and under any part of the
nuclear power plant, unless evidence indicates this assumption is
not appropriate, and must take into account the estimated rate at
which the surface deformation may occur.
(c) Seismically Induced Floods and Water Waves and Other Design
Conditions. Seismically induced floods and water waves from either
locally or distantly generated seismic activity and other design
conditions determined pursuant to Sec. 100.23 of this chapter must
be taken into account in the design of the nuclear power plant so as
to prevent undue risk to the health and safety of the public.
Part 52--Early Site Permits; Standard Design Certifications; and
Combined Licenses for Nuclear Power Plants
11. The authority citation for Part 52 continues to read as
follows:
Authority: Secs. 103, 104, 161, 182, 183, 186, 189, 68 Stat.
936, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 1244,
as amended (42 U.S.C. 2133, 2201, 2232, 2233, 2236, 2239, 2282);
secs. 201, 202, 206, 88 Stat. 1242, 1244, 1246, as amended (42
U.S.C. 5841, 5842, 5846).
12. In Sec. 52.17, the introductory text of paragraph (a)(1) and
paragraph (a)(1)(vi) are revised to read as follows:
[[Page 65175]]
Sec. 52.17 Contents of applications.
(a)(1) The application must contain the information required by
Sec. 50.33 (a) through (d), the information required by Sec. 50.34
(a)(12) and (b)(10), and to the extent approval of emergency plans is
sought under paragraph (b)(2)(ii) of this section, the information
required by Sec. 50.33 (g) and (j), and Sec. 50.34 (b)(6)(v) of this
chapter. The application must also contain a description and safety
assessment of the site on which the facility is to be located. The
assessment must contain an analysis and evaluation of the major
structures, systems, and components of the facility that bear
significantly on the acceptability of the site under the radiological
consequence evaluation factors identified in Sec. 50.34(a)(1) of this
chapter. Site characteristics must comply with part 100 of this
chapter. In addition, the application should describe the following:
* * * * *
(vi) The seismic, meteorological, hydrologic, and geologic
characteristics of the proposed site;
* * * * *
PART 54--REQUIREMENTS FOR RENEWAL OF OPERATING LICENSES FOR NUCLEAR
POWER PLANTS
13. The authority citation for Part 54 continues to read as
follows:
Authority: Secs. 102, 103, 104, 161, 181, 182, 183, 186, 189, 68
Stat. 936, 937, 938, 948, 953, 954, 955, as amended, sec. 234, 83
Stat. 1244, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201,
2232, 2233, 2236, 2239, 2282); secs. 201, 202, 206, 88 Stat. 1242,
1244, as amended (42 U.S.C. 5841, 5842).
14. In Sec. 54.4, paragraph (a)(1)(iii) is revised to read as
follows:
Sec. 54.4 Scope.
(a) * * *
(1) * * *
(iii) The capability to prevent or mitigate the consequences of
accidents that could result in potential offsite exposure comparable to
the guidelines in Sec. 50.34(a)(1) or Sec. 100.11 of this chapter, as
applicable.
* * * * *
PART 100--REACTOR SITE CRITERIA
15. The authority citation for Part 100 continues to read as
follows:
Authority: Secs. 103, 104, 161, 182, 68 Stat. 936, 937, 948,
953, as amended (42 U.S.C. 2133, 2134, 2201, 2232); sec. 201, as
amended, 202, 88 Stat. 1242, as amended, 1244 (42 U.S.C. 5841,
5842).
16. The table of contents for Part 100 is revised to read as
follows:
PART 100--REACTOR SITE CRITERIA
Sec.
100.1 Purpose.
100.2 Scope.
100.3 Definitions.
100.4 Communications.
100.8 Information collection requirements: OMB approval.
Subpart A--Evaluation Factors for Stationary Power Reactor Site
Applications Before January 10, 1997 and for Testing Reactors
100.10 Factors to be considered when evaluating sites.
100.11 Determination of exclusion area, low population zone, and
population center distance.
Subpart B--Evaluation Factors for Stationary Power Reactor Site
Applications on or After January 10, 1997
100.20 Factors to be considered when evaluating sites.
100.21 Non-seismic site criteria.
100.23 Geologic and seismic siting criteria.
Appendix A to Part 100--Seismic and Geologic Siting Criteria for
Nuclear Power Plants
17. Section 100.1 is revised to read as follows:
Sec. 100.1 Purpose.
(a) The purpose of this part is to establish approval requirements
for proposed sites for stationary power and testing reactors subject to
part 50 or part 52 of this chapter.
(b) There exists a substantial base of knowledge regarding power
reactor siting, design, construction and operation. This base reflects
that the primary factors that determine public health and safety are
the reactor design, construction and operation.
(c) Siting factors and criteria are important in assuring that
radiological doses from normal operation and postulated accidents will
be acceptably low, that natural phenomena and potential man-made
hazards will be appropriately accounted for in the design of the plant,
that site characteristics are such that adequate security measures to
protect the plant can be developed, and that physical characteristics
unique to the proposed site that could pose a significant impediment to
the development of emergency plans are identified.
(d) This approach incorporates the appropriate standards and
criteria for approval of stationary power and testing reactor sites.
The Commission intends to carry out a traditional defense-in-depth
approach with regard to reactor siting to ensure public safety. Siting
away from densely populated centers has been and will continue to be an
important factor in evaluating applications for site approval.
18. Section 100.2 is revised to read as follows:
Sec. 100.2 Scope.
The siting requirements contained in this part apply to
applications for site approval for the purpose of constructing and
operating stationary power and testing reactors pursuant to the
provisions of part 50 or part 52 of this chapter.
19. Section 100.3 is revised to read as follows:
Sec. 100.3 Definitions.
As used in this part:
Combined license means a combined construction permit and operating
license with conditions for a nuclear power facility issued pursuant to
subpart C of part 52 of this chapter.
Early Site Permit means a Commission approval, issued pursuant to
subpart A of part 52 of this chapter, for a site or sites for one or
more nuclear power facilities.
Exclusion area means that area surrounding the reactor, in which
the reactor licensee has the authority to determine all activities
including exclusion or removal of personnel and property from the area.
This area may be traversed by a highway, railroad, or waterway,
provided these are not so close to the facility as to interfere with
normal operations of the facility and provided appropriate and
effective arrangements are made to control traffic on the highway,
railroad, or waterway, in case of emergency, to protect the public
health and safety. Residence within the exclusion area shall normally
be prohibited. In any event, residents shall be subject to ready
removal in case of necessity. Activities unrelated to operation of the
reactor may be permitted in an exclusion area under appropriate
limitations, provided that no significant hazards to the public health
and safety will result.
Low population zone means the area immediately surrounding the
exclusion area which contains residents, the total number and density
of which are such that there is a reasonable probability that
appropriate protective measures could be taken in their behalf in the
event of a serious accident. These guides do not specify a permissible
population density or total population within this zone because the
situation may vary from case to case. Whether a specific number of
people can, for example, be evacuated from a specific area, or
instructed to take shelter, on a timely basis will depend on many
factors such as location, number and size of highways, scope and extent
of
[[Page 65176]]
advance planning, and actual distribution of residents within the area.
Population center distance means the distance from the reactor to
the nearest boundary of a densely populated center containing more than
about 25,000 residents.
Power reactor means a nuclear reactor of a type described in
Sec. 50.21(b) or Sec. 50.22 of this chapter designed to produce
electrical or heat energy.
Response spectrum is a plot of the maximum responses (acceleration,
velocity, or displacement) of idealized single-degree-of-freedom
oscillators as a function of the natural frequencies of the oscillators
for a given damping value. The response spectrum is calculated for a
specified vibratory motion input at the oscillators' supports.
Safe Shutdown Earthquake Ground Motion is the vibratory ground
motion for which certain structures, systems, and components must be
designed pursuant to appendix S to part 50 of this chapter to remain
functional.
Surface deformation is distortion of geologic strata at or near the
ground surface by the processes of folding or faulting as a result of
various earth forces. Tectonic surface deformation is associated with
earthquake processes.
Testing reactor means a testing facility as defined in Sec. 50.2 of
this chapter.
20. Section 100.4 is added to read as follows:
Sec. 100.4 Communications.
Except where otherwise specified in this part, all correspondence,
reports, applications, and other written communications submitted
pursuant to this part 100 should be addressed to the U.S. Nuclear
Regulatory Commission, ATTN: Document Control Desk, Washington, DC
20555-0001, and copies sent to the appropriate Regional Office and
Resident Inspector. Communications and reports may be delivered in
person at the Commission's offices at 2120 L Street, NW., Washington,
DC, or at 11555 Rockville Pike, Rockville, Maryland.
21. Section 100.8 is revised to read as follows:
Sec. 100.8 Information collection requirements: OMB approval.
(a) The Nuclear Regulatory Commission has submitted the information
collection requirements contained in this part to the Office of
Management and Budget (OMB) for approval as required by the Paperwork
Reduction Act of 1995 (44 U.S.C. 3501 et seq.). OMB has approved the
information collection requirements contained in this part under
control number 3150-0093.
(b) The approved information collection requirements contained in
this part appear in Sec. 100.23 and appendix A to this part.
22. The undesignated centerheading preceding Sec. 100.10 is
removed, Secs. 100.10 and 100.11 are designated as subpart A, and the
subpart A heading is added to read as follows:
Subpart A--Evaluation Factors for Stationary Power Reactor Site
Applications Before January 10, 1997 and for Testing Reactors
23. Subpart B consisting of Secs. 100.20, 100.21 and 100.23 is
added to part 100 to read as follows:
Subpart B--Evaluation Factors for Stationary Power Reactor Site
Applications on or After January 10, 1997
Sec. 100.20 Factors to be considered when evaluating sites.
The Commission will take the following factors into consideration
in determining the acceptability of a site for a stationary power
reactor:
(a) Population density and use characteristics of the site
environs, including the exclusion area, the population distribution,
and site-related characteristics must be evaluated to determine whether
individual as well as societal risk of potential plant accidents is
low, and that physical characteristics unique to the proposed site that
could pose a significant impediment to the development of emergency
plans are identified.
(b) The nature and proximity of man-related hazards (e.g.,
airports, dams, transportation routes, military and chemical
facilities) must be evaluated to establish site parameters for use in
determining whether a plant design can accommodate commonly occurring
hazards, and whether the risk of other hazards is very low.
(c) Physical characteristics of the site, including seismology,
meteorology, geology, and hydrology.
(1) Section 100.23, ``Geologic and seismic siting factors,''
describes the criteria and nature of investigations required to obtain
the geologic and seismic data necessary to determine the suitability of
the proposed site and the plant design bases.
(2) Meteorological characteristics of the site that are necessary
for safety analysis or that may have an impact upon plant design (such
as maximum probable wind speed and precipitation) must be identified
and characterized.
(3) Factors important to hydrological radionuclide transport (such
as soil, sediment, and rock characteristics, adsorption and retention
coefficients, ground water velocity, and distances to the nearest
surface body of water) must be obtained from on-site measurements. The
maximum probable flood along with the potential for seismically induced
floods discussed in Sec. 100.23 (d)(3) must be estimated using
historical data.
Sec. 100.21 Non-seismic siting criteria.
Applications for site approval for commercial power reactors shall
demonstrate that the proposed site meets the following criteria:
(a) Every site must have an exclusion area and a low population
zone, as defined in Sec. 100.3;
(b) The population center distance, as defined in Sec. 100.3, must
be at least one and one-third times the distance from the reactor to
the outer boundary of the low population zone. In applying this guide,
the boundary of the population center shall be determined upon
consideration of population distribution. Political boundaries are not
controlling in the application of this guide;
(c) Site atmospheric dispersion characteristics must be evaluated
and dispersion parameters established such that:
(1) Radiological effluent release limits associated with normal
operation from the type of facility proposed to be located at the site
can be met for any individual located offsite; and
(2) Radiological dose consequences of postulated accidents shall
meet the criteria set forth in Sec. 50.34(a)(1) of this chapter for the
type of facility proposed to be located at the site;
(d) The physical characteristics of the site, including
meteorology, geology, seismology, and hydrology must be evaluated and
site parameters established such that potential threats from such
physical characteristics will pose no undue risk to the type of
facility proposed to be located at the site;
(e) Potential hazards associated with nearby transportation routes,
industrial and military facilities must be evaluated and site
parameters established such that potential hazards from such routes and
facilities will pose no undue risk to the type of facility proposed to
be located at the site;
(f) Site characteristics must be such that adequate security plans
and measures can be developed;
(g) Physical characteristics unique to the proposed site that could
pose a significant impediment to the development of emergency plans
must be identified;
(h) Reactor sites should be located away from very densely
populated
[[Page 65177]]
centers. Areas of low population density are, generally, preferred.
However, in determining the acceptability of a particular site located
away from a very densely populated center but not in an area of low
density, consideration will be given to safety, environmental,
economic, or other factors, which may result in the site being found
acceptable 3.
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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 criteria.
This section sets forth the principal geologic and seismic
considerations that guide the Commission in its evaluation of the
suitability of a proposed site and adequacy of the design bases
established in consideration of the geologic and seismic
characteristics of the proposed site, such that, there is a reasonable
assurance that a nuclear power plant can be constructed and operated at
the proposed site without undue risk to the health and safety of the
public. Applications to engineering design are contained in appendix S
to part 50 of this chapter.
(a) Applicability. The requirements in paragraphs (c) and (d) of
this section apply to applicants for an early site permit or combined
license pursuant to Part 52 of this chapter, or a construction permit
or operating license for a nuclear power plant pursuant to Part 50 of
this chapter on or after January 10, 1997. However, for either an
operating license applicant or holder whose construction permit was
issued prior to January 10, 1997, the seismic and geologic siting
criteria in Appendix A to Part 100 of this chapter continues to apply.
(b) Commencement of construction. The investigations required in
paragraph (c) of this section are within the scope of investigations
permitted by Sec. 50.10(c)(1) of this chapter.
(c) Geological, seismological, and engineering characteristics. The
geological, seismological, and engineering characteristics of a site
and its environs must be investigated in sufficient scope and detail to
permit an adequate evaluation of the proposed site, to provide
sufficient information to support evaluations performed to arrive at
estimates of the Safe Shutdown Earthquake Ground Motion, and to permit
adequate engineering solutions to actual or potential geologic and
seismic effects at the proposed site. The size of the region to be
investigated and the type of data pertinent to the investigations must
be determined based on the nature of the region surrounding the
proposed site. Data on the vibratory ground motion, tectonic surface
deformation, nontectonic deformation, earthquake recurrence rates,
fault geometry and slip rates, site foundation material, and
seismically induced floods and water waves must be obtained by
reviewing pertinent literature and carrying out field investigations.
However, each applicant shall investigate all geologic and seismic
factors (for example, volcanic activity) that may affect the design and
operation of the proposed nuclear power plant irrespective of whether
such factors are explicitly included in this section.
(d) Geologic and seismic siting factors. The geologic and seismic
siting factors considered for design must include a determination of
the Safe Shutdown Earthquake Ground Motion for the site, the potential
for surface tectonic and nontectonic deformations, the design bases for
seismically induced floods and water waves, and other design conditions
as stated in paragraph (d)(4) of this section.
(1) Determination of the Safe Shutdown Earthquake Ground Motion.
The Safe Shutdown Earthquake Ground Motion for the site is
characterized by both horizontal and vertical free-field ground motion
response spectra at the free ground surface. The Safe Shutdown
Earthquake Ground Motion for the site is determined considering the
results of the investigations required by paragraph
(c) of this section. Uncertainties are inherent in such estimates.
These uncertainties must be addressed through an appropriate analysis,
such as a probabilistic seismic hazard analysis or suitable sensitivity
analyses. Paragraph IV(a)(1) of appendix S to part 50 of this chapter
defines the minimum Safe Shutdown Earthquake Ground Motion for design.
(2) Determination of the potential for surface tectonic and
nontectonic deformations. Sufficient geological, seismological, and
geophysical data must be provided to clearly establish whether there is
a potential for surface deformation.
(3) Determination of design bases for seismically induced floods
and water waves. The size of seismically induced floods and water waves
that could affect a site from either locally or distantly generated
seismic activity must be determined.
(4) Determination of siting factors for other design conditions.
Siting factors for other design conditions that must be evaluated
include soil and rock stability, liquefaction potential, natural and
artificial slope stability, cooling water supply, and remote safety-
related structure siting. Each applicant shall evaluate all siting
factors and potential causes of failure, such as, the physical
properties of the materials underlying the site, ground disruption, and
the effects of vibratory ground motion that may affect the design and
operation of the proposed nuclear power plant.
Dated at Rockville, Maryland, this 2nd day of December, 1996.
For the Nuclear Regulatory Commission.
John C. Hoyle,
Secretary of the Commission.
[FR Doc. 96-31075 Filed 12-10-96; 8:45 am]
BILLING CODE 7590-01-P