ML20140F466
| ML20140F466 | |
| Person / Time | |
|---|---|
| Issue date: | 05/29/1992 |
| From: | Beckjord E NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Rasin W NUCLEAR ENERGY INSTITUTE (FORMERLY NUCLEAR MGMT & |
| Shared Package | |
| ML20007G200 | List: |
| References | |
| FRN-57FR47802, RULE-PR-100, RULE-PR-50, RULE-PR-52 AD93-1-020, AD93-1-20, NUDOCS 9705050014 | |
| Download: ML20140F466 (2) | |
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- j. AP%-1 F'_Pf4 8 UNITED STATES 8 % NUCLEAR REGULATORY COMMISSION l 5 WASHINGTON, D. C. 20555 l
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\ S*** / MAY 2 91992 Mr. William H. Rasin '
l Vice President and Director
! Technical Division Nuclear Management and Resources Council Suite 300, 1776 Eye Street, N.W.
Washington, DC 20006-2496 l
" ear Mr. Rasin:
l This letter is in response to your letter of May 8,1992, pertaining to the proposed revision of Appendix A, " Seismic and Geologic Siting Criteria for Nuclear Power Plants," to 10 CFR Part 100, " Reactor Site Criteria." Specifi-l cally, your letter provided early industry comments on the proposed regula-tion, designated Appendix B to 10 CFR Part 100, that had been placed in the NRC Public Document Room on January 21, 1992 and focused on the use of both probabilistic and deterministic evaluations in seismic and geologic siting of
- nuclear power plants.
There are divergent views on the role probabilistic seismic hazard aialysis i should play in the licensing arena. For example, the NRC Advisory Committee on Reactor Safeguards (ACRS) in their February 14, 1992 letter states that the l need for a probabilistic approach was successfully argued by the staff but not l for the use of a dual approach. Your letter also acknowledges that the ACRS is not convinced that the proposed dual approach is either necessary or desirable.
There is a general consensus within the NRC staff that the revised seismic and geological siting criteria should allow considerations for a probabilistic l hazard analysis. There is also a general belief that the probabilistic analysis should be calibrated against the past practices for siting and licensing the current generation of nuclear power plants. There is a general consensus that ground motions should be calculated using deterministic methods once the controlling earthquakes are determined. With regard to the role of the probabilistic analysis, views range from an advocacy of a predominantly probabilistic analysis to the probabilistic/ deterministic procedure proposed in the regulation and supporting regulatory guide (DG-1015, " Identification and Characterization of Seismic Sources, Expected Maximum Earthquakes and Ground Motion") to a predominantly deterministic approach as used currently.
The staff within the Office of Nuclear Regulatory Research has reviewed your comments and is in general agreement with some of the clarifying language
. suggested in your letter. In particular, it is our intent that the degree of reliance that should be placed on probabilistic or deterministic procedures will depend on the type of tectonic region within which the site is located.
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William H. Rasin 2 MM 29 19E l
! In draft Regulatory Guide DG-1015, the staff has attempted to achieve this l l intent by.providing separate discussions on eastern and western U.S. sites. l As discussed in the February 5 and 7 ACRS meetings, the staff is planning to 5 l ' include, within the Supplemental Information portion of the Federal Reaister notice, specific questions regarding the use of probabilistic seismic hazard analysis and the balance between tlie. deterministic and probabilistic evalua-
' tions. These questions address issues very closely related to the issues l raised in your letter.
It is the staff's intent to seek comments from a variety of sources prior to '
finalizing its position on this issue. Be assured that your comments received l to date, future comments that you may provide, and other comments received :
- during the public comment period will receive serious consideration during the ,
j development of the final regulation and supporting regulatory guides. ;
I f I want to acknowledge the interest of NUMARC staff and the Ad Hoc Advisory Committee on the Appendix A Revision formed by NUMARC. I urge you to continue
, your active participation in this rulemaking activity. I appreciate your support. ;
Sincerely, ,
! N O ,t G .hhM Eric S. Beckjord, Virector Office of Nuclear Regulatory Research l
cc: L. C. Shao i R. J. Bosnak A. J. Murphy T..E. Murley J. E. Richardson i
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June 12, 1992 RULEMAKING ISSUE SECv-92-21s (Notation Vote) l f_qt: The Commissioners f_rps: James M. Taylor Executive Director for Operations Subiect: REVISION OF 10 CFR PART 100, REVISIONS T0 10 CFR PART 50, NEW APPENDIX B TO 10 CFR PART 100 AND NEW APPENDIX S TO 10 CFR PART 50 Puroose: To obtain Commission approval to publish for public comment proposed revisions to reactor siting regulations '
and associated regulatory guides for use by future applicants that will decouple siting from plant design and also reflect advancements in the earth sciences and earthquake engineering with regard to siting power reactors.
Summary: This proposed rule change to 10 CFR Part 100, " Reactor Site Criteria," is intended to accomplish three major objectives. The first change would add a new section to Part 100 (designated Subpart B) for future plants that would eliminate the use of a postulated accident source term and the use of dose calculations in the determination of acceptability of a nuclear power plant site. The existing requirements would be retained for existing plants and test reactors. The proposed regulation would set a minimum size for the exclusion area and would set population density criteria to be used only in the siting decision process for future reactor sites. The proposed population density criteria would not be upper limits of acceptability, but, if exceeded, would require consideration of alternative sites having lower population densities. The requirement for a low population zone (LPZ) would be deleted from 10 CFR Part 100 for future plants. Requirements regarding the evaluation of man-related hazards and the feasibility of carrying out protective actions in the event of a radiological emergency would be added to 10 CFR Part 100. ~
Contact:
Soffer, RES NOTE: TO BE MADE PUBLICLY AVAILABLE WHEN THE FINAL SRM IS MADE AVAILABLE i Dr. Andrew J. Murphy, RES i
492-3860
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L The Commissioners 2 The second change is to revise Appendix A " Seismic and' Geologic Siting Criteria for Nuclear Power Plants," to 10 L CFR Part 100 to reflect current understanding and advancements in the earth sciences and earthquake engineering with regard to reactor siting. The proposed regulation would require the use of both probabilistic and deterministic evaluations in reactor siting. Also, '4 detailed guidance on acceptable investigations or design bases would be deleted from the regulation and placed in a regulatory guide. The revised. criteria would not be applied to existing plants. Therefore, the proposed criteria will be designated Appendix B so that the licensing basis for existing plants is maintained in Appendix A.
The third part of this rulemaking effort is revisions to l Part 50. One portion of the Part 50 revision would add, t on an interim basis, the source term and' dose calculations being deleted from Part 100. The source term and dose calculations to be added to Part 50 would be used for evaluating plant features, not site suitability. A second ,
portion would transfer all seismic criteria not associated l with the selection of the site or establishment of the !
Safe Shutdown Earthquake Ground Motion (SSE) from Appendix A to Part 100 to Appendix S to Part 50. Section 50.54 has been revised to require plant shutdown if vibratory ground motion exceeding that of an Operating Basis Earthquake l Ground Motion (OBE) or significant plant damage occurs.
I Backaround: L Reactor Sitina Criteria (Nonseismic):
l The present criteria regarding reactor siting were issued in April 1962. There were only a few small power reactors operating at that time.' The present regulation requires !
that every reactor have an exclusion area that normally l has no permantnt residents; transient use is permitted. A l low population zone immediately beyond the exclusion area is also required, within which protective actions can be taken. The regulation recognizes the importance of !
accident considerations in reactor siting; hence, a key element is determining the size of the exclusion area via 4
the postulation of a large accidental fission product release within containment and the evaluation of the radiological consequences, in terms of doses. Doses are calculated for two hypothetical individuals located at any
, point (generally, the closest point) on the exclusion area i boundary and at the outer radius of the low population i
zone; these doses are required to be within specified limits (25 rem to the whole body and 300 rem to the l thyroid gland). In addition, the nearest population
- center, containing about 25,000 or more residents, may be no closer than one and one-third times the outer radius of L - _ , - _
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The Commissioners- 3 the low population zone. The effect of these requirements '
is to set both individual and, to some extent, societal '
limits on dose (and implicitly on risk) without cetting numerical criteria on exclusion area and low population zone size. Numerical limits on population are also not specified. However, since 1975, Regulatory Guide 4.7,
" General Site Suitability Criteria for Nuclear Power Stations," has provided guidelines on acceptable exclusion i area distance and population density and has been used in the review of sites.
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 l into the regulations. On A 28, 1977 Free Environment, Inc, et. al., filed pril 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. In August '
1978, the Commission directed the 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 1979 and provided recommendations in this regard.. On July 29, 1980, the NRC issued an Advance Notice of Proposed Rulemaking (ANPR) (45 FR 50350) regarding revision of the reactor siting 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 l development of the Safety Goal and improved research on i
accident source terms. On August 4, 1986, the Policy i Statement on Safety Goals was issued (51 FR 23044). On l
November 29, 1988, the PIRG petition was denied (28 NRC 829)=on the basis that it would unnecessarily restrict NRC's regulatory siting policies and would not result in a substantial increase in the overall protection of the public health and safety. Since the proposed regulation would include population density criteria for future nuclear power reactor sites, the staff concludes that the i petition filed by Free Environment,~Inc. would be l addressed as part of this rulemaking. The staff plans to l
L send a letter to the petitioner informing him of the actions taken in regard to his petition at the time a proposed rule is issued for comment.
In SECY-90-341, dated October 4, 1990, and in a subsequent memorandum from J. Taylor to the Commissioners dated i
December 13, 1990, the staff proposed to decouple siting from plant design for future plants via a two step 4
l rulemaking. Step one is to modify Part 100 to directly
!. address the site criteria while moving the dose t
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! The Commissioners 4 requirements currently in Part 100 to Part 50 on an [
- r. interim basis. Step two is to update Part 50 to reflect l current source term information and to replace the interim !
. dose requirements with updated design criteria. The Commission, in a Staff Requirements Memorandum (SRM) dated January 25, 1991, approved the staff recommendation. This :
paper presents step one of the proposed regulation change.
l L Seismic Sitina and Earthouake Enaineerina Criteria !
1 Appendix A, " Seismic and Geologic Siting Criteria for i Nuclear Power Plants," to 10 CFR Part 100, " Reactor Site i
' Criteria," was originally issued as a proposed rule on November 25,1971 (36 FR 22601), published as a final rule ,
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 rule i by adding the legend under the diagram. The second i amendment resulted from a petition for rulemaking (PRM .
100-1) requesting that an opinion be issued on interpreting and clarifying Appendix A with respect to the {
j 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 rule on January 10, 1977 (42 FR 2052).
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Discussion: The proposed regulation changes included with this paper :
i primarily involve two related but basically separate >
changes. The first change involves the nonseismic portion ;
of the reactor site criteria, 10 CFR Part-100. The second change involves updating the seismic and earth sciences j siting criteria in Appendix A to Part 100. i L Reactor Sitina Criteria (Nonseismic) ;
I The proposed revision to Part 100 would retain the current !
criteria for existing plants and nonpower reactors, including the dose requirements. The current criteria cre 4
designated subpart A and apply to nonpower reactors and to i plants currently licensed or applying for a license prior j to the effective data of the proposed regulation. A new i subpart B would be added to Part 100. Subpart B would r
contain the proposed requirements for power reactor !
applicants after the effective date of the proposed i regulation. Part 52 Appendix Q would be amended to note l the potential for revisiting the population density and t maneade hazard potential for renewal of early site i permits. !
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I l 'The Commissioners 5 l
- These proposed changes are based.on current' staff practice i and have been derived from the guidelines in Regulatory Guide 4.7, " General Site Suitability Criteria for Nuclear {
i Power Stations." ' Experience in implementing this Guide 4
over the past 17 years has shown.that application of this siting guidance is expected to result in low risk to the 1
public while allowing a good selection of sites in all. -
regions of the nation. It also reflects the Commission's.
! policy to keep reactors away from densely populated centers. In addition, risk studies conducted over the 4 same period on radioactive material releases under ;
j accident conditions (e.g., the Reactor Safety Study, WASH- ;
'1400 as well as NUREG-Il50) have confirmed the i acceptability of the present practice in limiting risk to I j i the public. l l In developing the proposed changes, the staff considered i i the Commission's Safety Goal Policy Statement along with l' the recommendation of the Siting Policy Task Force (NUREG-4625) of 1979. The proposed regulation would :
[ require a minimum exclusion area distance of 0.4 miles ,
- (640 meters) for future stationary power reactors. Staff '
experience has shown that,'given an exclusion area of this
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i size, plant engineered safety features (e.g., sprays, 1 filters and allowable containment leak rates) can be
. provided to satisfy the dose guidelines of Part 100. It .
i should be noted that 25 of the 75 current power reactor i i sites in the U.S. have exclusion area distances smaller j than 0.4 miles, although all of the sites meet the dose i 4 guidelines of Part 100. If adopted, this proposed
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i rulechange would preclude siting additional reactors at ,
,. those 25 sites. I
!'; The proposed regulation would also include population criteria and would state that at the time of initial site approval (as in the case of an early site permit), offsite j population density values averaged over any radial distance out to 30 miles-(50 kilometers) should not exceed 500 people per square mile. In addition, the projected offsite population density 40 years after the time of I initial site approval or early site permit renewal should not exceed 1000 people per square mile out to a radial distance of 30 miles.
Based on staff experience, the proposed population density values provide reasonable criteria to guide initial siting decisions, and if exceeded, to require consideration of alternative sites having lower population densities.
However, because severe accident risk considerations show that low risk can be achieved for sites having significantly higher population densities than these siting decision values, the proposed population density
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'6[ d The Commissioners 6 values do not' represent upper limits of acceptability.
For this reason, the staff does not propose that' license conditions or restrictions be imposed on operating reactors if the population density around an operating reactor reaches or exceeds the proposed siting decision values during a plant's operating lifetime. Currently, about eight reactor sites are estimated to have population densities in excess of these proposed values. All of these were reviewed and approved prior to the issuance of
, Regulatory Guide 4.7 in 1975.
During discussion of the proposed rule, the Committee to Review Generic Requirements (CRGR) raised concerns regarding the advisability of putting-numerical values of population density in the rule itself, rather than retaining these in a regulatory guide, as present.
Population density values in the rule would have the benefit of limiting litigation of these values in individual site hearings. However, some staff believe that stating numerical values in the rule may imply a greater precision than warranted and might also pose an.
impediment to revision if an improved future basis led to revised densities. In addition, even though the population densities are. stated to apply only for future reactor sites, they may, by implication,' raise concerns for a number of existing reactor sites. While the staff recommends issuance of the proposed rule, the staff also wishes to inform the Commission of these concerns and to note that the Federal Register notice is explicitly requesting comments in this area.
The proposed regulation would add or modify existing requirements for obtaining information to characterize meteorological and hydrological factors at a site. This information will then be reviewed by the staff for evaluating plant design features in matching a proposed design to the site.
Site meteorological characteristics are proposed to be eliminated as a factor in determining site suitability.
Staff experience, as well as contractor studies regarding site meteorology, have shown that while meteorological conditions at a given site vary significantly over time, there is much less variation from site to site. The differences in site meteorology should be reflected in the
, design requirements for certain plant features. However based on the above studies, the staff concludes that the average meteorological characteristics between one site and another are sufficiently similiar that characterization of individual site meteorology is not a good discriminator with regard to site suitability. To obtain additional views on this matter, the proposed
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1 The Commissioners 7
. Federal Register Notice has included a question on the inclusion of meteorological criteria in Part 100.
4 The proposed regulation reflects the requirement currently in 10 CFR Part 52.17 for review of emergency planning i considerations for early site permits. The rule would
{ require that important site factors such as population
! distribution, topography, and transportation routes be considered and examined in order to determine whether
- there are any site characteristics that could pose a
- significant impediment to the development of an emergency j plan. Limitations of access or egress in the immediate i vicinity of a nuclear power plant should be identified at j the site approval phase.
1 j A proposed revision to Regulatory Guide 4.7, for j consistency with the proposed regulation, is also included p in the package.
B. Seismic Sitina and Earthouake Enaineerina Criteria The staff proposes to amend the regulations to update the seismic siting and engineering criteria for new nuclear power plants. The proposed regulatory action is applicable only to applicants that apply for a construction permit, operating license, early site permit, i design certification, or combined license (construction permit and operating license) on or after the effective date of the regulation.
The proposed regulation would allow NRC to benefit from J experience gained in the application of the procedures and .
methods set forth in the current-regulation, the .
difficulties encountered, and the rapid advancement in the state of the art of earth sciences. Detailed guidance that has created difficulty for applicants and the staff in terms of inhibiting the use of needed judgment, latitude, and the use of evolving methods of analyses has ;
been deleted from the proposed regulations and placed into i a proposed regulatory guide. Also, the proposed ;
regulation would require the use of probabilistic as well as deterministic evaluations to determine the vibratory ground motion at the site. Probabilistic analyses will provide an explicit expression of the overall uncertainty in the derived ground motion.
The proposed regulations would better reflect industry design practices and the associated staff review procedures that have evolved since the regulation was issued. For example, the proposed regulation would move
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The Commissioners 8 ,
the location of the seismic input motion control point from the foundation level to the ground surface.
Criteria not-associated with the selection of the site or establishment of the Safe Shutdown Earthquake Ground Motion (SSE) have been placed into Part 50. This action i is consistent with the location of other design !
requirements in Part 50.
The specification that the Operating Basis Earthquake Ground Motion (0BE), the vibratory ground motion that will .
t , assure safe continued operation, is one-half the SSE has .
been deleted from the proposed regulation and replaced !
with two options: applicant selection of an OBE that is l 'either one-third of the SSE or greater. With the OBE !
! level set at one-third or less of the SSE, only the SSE is l used for design; the OBE only serves the function of an inspection and shutdown' level. If the OBE is greater than one-third of the SSE, the current practice of using both the OBE and SSE for design continues; and in addition, the j OBE serves the function of an inspection and shutdown l level. This change responds to one of the major criticisms with the existing regulations, that the OBE controls the design of some parts of the plant.
l The proposed regulation (for new applications) would treat plant shutdown associated with vibratory ground motion exceeding the OBE (or significant plant damage) as a '
condition in every operating license. Section 50.54 is proposed to be revised accordingly. Related plant' shutdown and OBE exceedance guidelines for operating plants are being developed separately by NRR.
Because the revised criteria presented in the proposed regulation will not be applied to existing plants, the licensing bases for existing nuclear power plants must ,
remain part of the regulations. Therefore, the proposed criteria on. seismic and geologic siting would be '
designated as a new Appendix B to 10 CFR Part 100 and r would be added to the existing body of regulations. In addition, earthquake engineering criteria will be located in 10 CFR Part 50, in a new Appendix S. Since Appendix S is not self executing, applicable sections of Part 50 (550.8 and 550.34) are revised to reference Appendix S.
The proposed regulation would also make conforming amendments to 10 CFR Parts 52 and 100. Sections 52.17(a)(1), 52.17(a)(1)(vi), 100.8, and 100.20(c)(1) and (3) would be amended to note Appendix B to Part 100 or Appendix S to Part 50.
The staff has developed the following draft regulatory guides and standard review plan section to provide
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prospective licensees with the necessary guidance for implementing the proposed regulations:
DG-1015, " Identification and Characterization of Seismic Sources, Deterministic Source Earthquakes and Ground Motion," 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.
DG-1016, Second Proposed Revision 2 to Regulatory Guide 1.12, " Nuclear Power Plant Instrumentation for Earthquakes," ' describes seismic instrumentation type and location, operability, characteristics, installation, actuation, and maintenance that are acceptable to the NRC staff.
DG-1017, " Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Post-Earthquake Actions," provides guidelines that are acceptable to the NRC staff for a timely evaluation of the recorded seismic instrumentation data and for determining whether or not plant shutdown is required.
DG-1018, " 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 restart of a plant that has been shut down by a seismic event.
Draft Standard Review Plan Section 2.5.2, Proposed Revision 3, " Vibratory Ground Motion," describes procedures to assess the ground motion potential of seismic sources at the site and to assess the adequacy of the SSE.
General The draft guides and standard review plan section should be issued simultaneously with the proposed revision to the regulations. Additional minor and conforming changes to other Regulatory Guides and standard review plan sections are also package. planned but are not included as part of this These are discussed in the statement of considerations (Enclosure 1).
During the development of this proposed rule the staff benefitted from four public meetings with industry groups.
Principal attendees included staff from the Nuclear Management and Resources Council (NUMARC), Electric Power
A.
- The Commissioners 10 Research Institute (EPRI), Department of Energy (DOE) and industry. During the first meeting (March 6, 1991), the staff discussed schedule and technical topics for potential inclusion in the revision of Appen:'lx A to Part 100. The other meetings (April 17, 1991, Fe ruary 4, 1992 and April 23,1992) provided industry and other interested members of the public with an opportunity to express their-views on the Appendix A revision.
The enclosed Federal Register Notice contains information on the scope' of this rulemaking and requests public input.
The Federal Register Notice also addresses actions related to new and revised regulatory guides and standard review plan sections.
The ACRS subcommittees were briefed on the staff's approach on December 10,' 1991 (seismic), January 7,1992 (nonseismic), and February 5, 1992 (seismic). The ACRS full committee was briefed on January 10, 1992 (nonseismic) and on February 7, 1992 (seismic). The ACRS provided comments to the Comission in . letters dated January 15, 1992 (Enclosure 14) and February 14, 1992 (Enclosure 15).
In the letter of January 15, 1992, the ACRS stated that they believed that the staff's proposed revision to Part 100 and the proposed interim revision to Part-50 were reasonable and should proceed. However, they recommended further work with regard to both Part 100 and Part 50 as part of the staff's longer term efforts to revise Part 50.
Regarding Part 100, the ACRS recommended further work to reexamine or justify the basis for key requirements such as the exclusion zone, emeEgency planning zone (EFZ), and the maximum population density in light of the large amount of experience and information that has been accumulated since 1962. Further,_the ACRS recommended that the relation of these requirements to the Safety Goal Policy should be established. Finally, the ACRS recommended that meteorological requirements be incorporated into Part 100 to exclude " unacceptable" sites.
In the letter of February 14, 1992, dealing with the seismic portions of the proposed regulation, the ACRS stated that they have no reservations or concerns at this time that would argue against publication far comment of the several proposed revisions considered in their review.
The staff considered the issues raised by the ACRS in the development of the proposed Part 100 regulation. A single
' < revision of Part 100 was proposed in SECY-90-341 as well as in a memorandum to the Comission dated December 13, l
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The Commissioners ^i 11 '
i 1990. The purpose was to complete the Part 100 update :
prior. to the expected submittal date of. an application for i an early site permit, as part of a Department of Energy ;
sponsored initiative. This proposal was approved by the ;
Commission in its SRM dated January 25, 1991. The staif i continues to believe this approach appropriate and is working to have all Part 100 revisions completed in one revision. In this regard, the staff is requesting
', comments on those issues raised by the ACRS in order to resolve these issues in a single rulemaking effort. .
The staff believes that codifying the guidance of !
Regulatory Guide 4.7 is appropriate and reflects the large amount of experience gained in licensing reviews. . As noted earlier, the basis for the exclusion area radius is {
that staff experience has shown that the dose values of Part 100 will likely be met at this distance for a typical plant having available engineered safety features. In addition, the staff.has evaluated the proposed radius in l relation to the Safety Goals and has confirmed that the L proposed value will meet the quantitative health objectives for a 3800 MWt light water reactor.
Application of the proposed population density values are l.
expected to keep large population centers away from the plant and in practice would accomplish what the LPZ is intended to accomplish, while still allowing for a i
' reasonable selection of sites in all regions of the nation. The staff also confirmed that for a plant similar to those evaluated in NUREG-1150, the quantitative health 1
I objectives (QHO) would be met at the recommended population density. However, because the QH0's are based l- on individual risk, the QH0s do not provide a measure of the appropriateness of'any specific population density.
The staff also reexamined the ten mile EPZ in SECY-90-341, in response to the Commission's SRM of February 13, 1990, and noted that today's methodologies tended to indicate that radiation doses and consequences would generally be lower at a given distance than previously predicted.
However, the staff recommended that the present EPZ be maintained.in order to provide assurance that an adequate planning base be maintained.
As noted earlier, studies.have indicated that individual site meteorological characteristics are not a good discriminator of site suitability. !
l Finally, the ACRS raised several concerns regarding the staff's long term effort to update Part 50 and the i j development of a replacement for the TID-14844 source 2
i term. These concerns are being considered by the staff '
and will be addressed in these longer term efforts. !
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i The Commissioners 12 The Office of the General Counsel has reviewed this paper i and has no legal objections. l Recommendations: That the Commission. .,
- 1. Acorove the issuance'of the enclosed draft documents for a 90-day public comment period. ;
- 2. Certify that this rule, if promulgated, will not have I a significant economic effect on a substantial number i of small entities pursuant to the Regulatory ,
Flexibility Act of 1980 (5 U.S.C. 605 (b)).
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- 3. '
If211:
- a. The proposed regulation and draft Federal I Register notice (Enclosure 1) and notice of availability of draft regulatory guides and draft standard review plan section (Enclosure 5) will be published in the Federal Reaister for a i 90-day public ccament period. !
i
- b. A notice of availability of a regulatory I 4
analysis (Enclosure 2) and an environmental !
assessment and finding of no significant i environmental impact (Enclosure 3) will be ;
- upplied concurrently to the Public Document i Room. !
s c.
Because Appendix S to Part 50 and Appendix B to Part 100 are new, an "information collection :
requirement" is being submitted to OMB for review (Enclosure 4). It is noted that the :
overall estimated burden on the staff and industry remains essentially the same; the ;
proposed revisions have added requirements to l use probabilistic evaluations in seismic and j geologic siting while potentially reducing the t required earthquake engineering analyses. ;
- d. A public annour.coment (Enclosure 12) will be ;
issued when the notice of proposed rulemaking '
and notice of availability of the draft !
regulatory guides and draft standard review plan i section are filed with the Office of the Federal ;
Reaister. -
- e. The appropriate Congressional committees will be I informed (Enclosure 13).. j
- f. Copies of the Federal Reaister notices will be distributed to all power reactor permittees and ,
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l0 1 j L The Commissioners 13 licensees. The notices will be sent to other '
interested parties upon request.
- g. The Chief Counsel for Advocacy of the Small
-Business Administration will be notified of the Commission's determination, pursuant to the Regulatory Flexibility Act of 1980 (5 U.S.C. 605 (b)), that these proposed regulations, draft ,
regul: tory guides, and draft standard review l plan section will not have a significant economic effect on a substantial number of small i entities.
i
- h. A Backfit Analysis is not required for this proposed rule, because these amendments do not '
l involve any provisions that would impose !
backfits as defined in 10 CFR 50.109(a)(1). l I
l
, / i i f5 mes M. ylor l
! xecutive Director l for Operations L
Enclosures:
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- 1. . Federal Register Notice of Rulemaking
- 2. Regulatory Analysis !
- 3. Environmental Assessment !
- 4. OMB Reporting Review Package.
- 5. Federal Register Notice of Regulatory Guide and Standard Review Plan Section Availability .
- 6. Proposed Kevised Regulatory Guide 4.7, (General Site i Suitability Criteria) l 7. Proposed Draft Regulatory Guide DG-1015, (Seismic Sources)
- j. 8. Proposed Draft Regulatory Guide DG-1016, Second Proposed i l Revision 2 to Regulatory Guide 1.12, (Seismic Instrumentation)
- l. 9. Proposed Draft Regulatory Guide DG-1017, (Plant Shutdown)
- 10. Proposed Draft Regulatory Guide DG-1018, (Plant Restart) !
- 11. Proposed Revision 3 to Standard Review Plan ,
Section 2.5.2 (Vibratory Ground Motion) i l 12. Draft Public Announcement
! 13. Draft Congressional Letters
! 14. ACRS January-15, 1992 Letter 3
- 15. ACRS February 14, 1992 Letter i
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Commissioners' comments or consent should be provided directly ,
to the Office of the Secretary by COB Wednesday, July 1, 1992. '
l Commission Staff Office comments, if any, should be submitted l to the Commissioners NLT Wednesday, June 24, 1992, with an information copy to the Office of the Secretary. If the paper is of such a nature that it requires. additional review and l comment, the Commissioners and the Secretariat should be '
apprised of when comments may be expected. ,
DISTRIBUTION-Commissioners !
OGC I OCAA i OIG IP OCA OPA f REGIONAL OFFICES EDO j ACRS ASLBP :
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1 DRAFT FEDERAL REGISTER NOTICE l PROPOSED REVISION OF 10.CFR PART 50 1
AND 10 CFR PART 100 PROPOSED ISSUANCE OF !
APPENDIX S TO 10 CFR PART 50 4
-AND APPENDIX B TO 10 CFR PART 100
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[7590 - 01]
NUCLEAR REGULATORY COMMIaSION
). 10 CFR Parts 50, 52 and 100 i ,
j RIN 3150-AD93 l l
Reactor Site Criteria !
Including Seismic and Earthquake Engineering Criteria for ;
- Nuclear Power Plants 4
i k , AGENCY: Nuclear Regulatory Commission.
l ACTION: Proposed rule. I i
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SUMMARY
- The Nuclear Regulatory Commission is proposing to amend its regulations I j
i to update the criteria used in decisions regarding power reactor siting, including geologic, seismic, and earthquake engineering considerations for future
! nuclear power plants. The proposed rule would allow NRC .to . benefit from ;
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' experience gained in the application of the procedures and methods set forth in j .the current regulation and to incorporate the rapid advancements in the earth i i
sciences and earthquake engineering. The proposed rule primarily consists of two ;
separate changes, namely, the source term and dose considerations, and the I
seismic and earthquake engineering considerations of reactor siting. l
- DATE: Coment period expires 90 days after date of publication in the Federal i Register. !
Coments received after this date will be considered if it is l practical to do so, but the Comission is able to assure consideration only for coments received on or before this date.
ADDRESSES: Mail written comments to: Secretary, U.S. Nuclear Regulatory Commission, Washington,.DC 20555, Attention: Docketing and Service Branch.
Deliver coments to 11555 Rockville Pike, Rockville, Maryland, between 7:45 am and 4:15 pm, Federal workdays.
Copies of the regulatory analysis, the environmental assessment and finding of no'significant impact, and comments received may be examined at the NRC Public Document Room at 2120 L Street NW. (Lower Level), Washington, DC.
FOR FURTHER INFORMATION CONTACT: Dr. Andrew J. Murphy, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Comission, Washington, DC 20555, telephone -(301) 492-3860, concerning the seismic and earthquake engineering aspects and Mr. Leonard Soffer, Office of Nuclear Regulatory Research, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 492-3916, concerning other siting aspects. i FRN - 1
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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. Siting Policy Task Force Recommendations. ,
VII. Related Regulatory Guides and Standard Review Plan Section.
VIII. Future Regulatory Action.
IX. Referenced Documents ,
X. Electronic Format.
XI. Questions.
XII. Finding of No Significant Environmental Impact: Availability.
XIII. Paperwork Reduction Act Statement.
-XIV. Regulatory Analysis.
XV. Regulatory Flexibility Certification.
XVI. Backfit Analysis.
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.o I. Background -
I The present regulation regarding reactor site criteria (10 CFR Part 100) was promulgated April 12, 1962 (27 FR 3509). 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 I
(PIRG) filedminimum incorporate a petition for rulemaking (PRM-100-2) requesting that the NRC exclusion area and low population zone distances and population density limits into the regulations.
On April 28, .1977, Free Envirunment, 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. 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. On July 29,1980(45FR50350),theNRC
' 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.
December .On 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 and in a substantial increase in the overall protection of the public health safety. Because of possible renewed interest in power reactor siting, the NRC is proceeding with a rulemaking in this area. Because the proposed regulations.would include population density criteria for future nuclear power reactor sites, the Commission concludes that the remaining issue in PRM-50-20 has been addressed as part of this rulemaking action.
Appendix A to 10 CFR Part 100, " Seismic and Geologic Siting Criteria for Nuclear Power Plants," was originally issued as a proposed regulation on November i 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 i 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 l 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).
II. Objectives The objectives of this proposed regulatory action are to --
- 1. State the criteria for future sites that, based upon experience and i importance to risk, have been shown as key to protecting public health and FRN - 3 a
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i 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 tecnnical understandings; and
- 3. Relocate the requirements that apply to plant design into 10 CFR Part 50 thereby effectively decoupling siting from plant design.
III. Genesis The proposed regulatory action reflects changes that are intended to (1) benefit from the experience gained in applying the existing regulation and from research; (2) resolve interpretive questions; (3) provide needed regulatory flexibility to incorporate state-of-the-art improvements in the geosciences and earthquake engineering; and (4) simplify the language to a more " plain English" text.
The prop Led regulatory action would apply to applicants who apply for a construction permit, operating license, preliminary design approval, final design approval, manufacturing license, early site permit design certification, or combined license on or after the effective data of we final regulations.
Criteria not associated with the selection of the site or establishment of the Safe Shutdown Earthquake Ground Motion (SSE) have been placed into 10 CFR Part 50. This action is consistent with the location of other design requirements in 10 CFR Part 50.
Because the revised criteria presented in the proposed regulation would not be applied to existing plants, the licensing bases for existing nuclear power plants must remain part of the regulations. Therefore, the proposed revised reactor siting criteria would be added as Subpart B in 10 CFR Part 100 and would apply to site applications received on or after the effective data of the final regulations. The criteria on seismic and geologic siting would be added as a new Appendix B to 10 CFR Part 100. The dose calculations and the earthquake engineering criteria will be located in 10 CFR Part 50 (950.34(a) and Appendix S,respectively). Because Appendix S is not self executing, applicable sections of Part 50 (550.34 and 550.54) are revised to reference Appendix S. The proposed regulation would also make conforming amendments to 10 CFR Parts 52 and 100. Sections 52.17(a)(1)(vi), and 100.20(c)(1) would be amended tu note Appendix B to 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 influence plant design requirements rather than siting. It is desirable to state directly those siting criteria which, through importance to risk, have been shown to be key to assuring public health and safety. Further, significant advances 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 FRN - 4
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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
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are in various stages of obtaining operating licenses. Therefore, these provisions must remain in force and effect. i
'The approach of establishing the revised requirements in new sections and an appendix to 10 CFR Part 100 and relocating plant design requirements to 10 CFR Part 50 while retaining the existing regulation was chosen as the best alternative. The public will benefit from a clearer, more uniform, and more consistent licensing process that incorporates updated information and is subject to fewer interpretations. The NRC staff will benefit from improved regulatory '
implementation (both technical and legal), fewer interpretive debates, and increased regulatory flexibility. Applicants will derive the same benefits in addition to avoiding licensing delays caused by unclear regulatory requirements.
V. Major Changes A Reactor Siting Criteria (Nonseismic). '
Since promulgation of the reactor site criteria in 1962, the Commission has approved more than 75 sites for nuclear power reactors and has had an opportunity ,
to review a number of others. As a result of these reviews, a great deal of experience has been gained regarding the site factors that influence risk and .
their range of accsptability. Much of the experience gained by the NRC staff in these reviews has been reflected in the issuance of Regulatory Guide 4.7, ,
" General Site Suitability Criteria for Nuclear Power Stations," which was issued for comment in 1974, and revised in 1975. It also reflects the Commission's policy of keeping reactors away from densely populated centers. A review of the Regulatory Guidelines implementation has shown that its application is expected to result in low risk to the public while allowing a good selection of potential l
reactor sites in all regions of the nation.
The site criteria presented in the proposed regulation are based on those contained primarily in Regulatory Guide 4.7, and represent current NRC practice.
In addition, numerous risk studies on radioactive material releases to the environment under severe accident conditions have all confirmed that the present ;
siting practice is expected to effectively limit risk to the public. These l 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. -
The proposed criteria basically decouple siting from accident source term ,
and dose calculations. Experience has shown that these factors have tended to influence plant design aspects rather than siting. Accident source term and dose considerations are proposed to be applied to plant design aspects and would be '
, relocated to Part 50. The Commission considers it appropriate, based on the extensive experience and confirmatory studies noted above, to state directly those site criteria that have been shown to be key to protecting public health :
-and safety. These reactor site criteria are expected to be independent of plant design and, as such, are independent of the plant type to be built at the site.
The Commission considers this appropriate because it expects that future reactors licensed under Part 50 or Part 52 will reflect, through their design, construction, and operation, risk characteristics that are equal to or better than existing plants. Therefore, there would be an extremely low probability for accidents that could result in release of significant quantities of radioactive '
fission products. In addition, the recommendations of the Siting Policy Task Force were considered in making these changes as discussed in Section XII of this FRN - 5
! LI proposed rule. - '
Rationale for Individual Criteria l
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E 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 afrom potential man-related hazards.
The present regulation has no numerical size requirement,- in terms of '
distance, for the exclusion area. The present regulations assesses the !
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' consequences of a postulated radioactive fission product release within i containment, coupled with assumptions regarding containment leakage, performance of certain fission product mitigation systems, and atmospheric dispersion factors i
for a hypothetical individual located at any point on the exclusion area l boundary. The plant and site combination is considered to be acceptable if the l
' calculated regulation. consequences do not exceed the dose values given in the present Regulatory Guide 4.7 suggests an exclusion area distance of 0.4 miles (640 meters). This distance has been found, in conjunction with typical l
engineered safety features, to meet the dose values in the existing regulation.
Future reactors would be expected to be as good or better in meeting the dose l
-criteria at this distance.
The Commission considers an exclusion area to be an essential feature of a reactor site and is retaining this requirement for future reactors. However, L
' in keeping with the recommendation of the Siting Policy Task Force to decouple site requirements from reactor design, the proposed regulation would eliminate the use of a postulated source term, assumptions regarding mitigation systems and !
dispersion factors, and the calculation of radiological consequences to determine l l
L the sizes of the exclusion area and low population zone. It would instead require !
l a minimum exclusion area distance of 0.4 miles (640 meters) for power reactors. i This distance, together with typical engineered safety features previously :
' reviewed by. the staff, has been found to satisfy the dose guidelines in the ;
present regulation. An exclusion area of this size or larger is fairly common i for most-power reactors in the U.S. It has not been unduly difficult for most i prospective. applicants to find and obtain a suitable site. 1 Finally, this distance has also been found to readily satisfy the prompt fatality quantitative health objective of the Commission's Safety Goals Policy,
'when coupled with plant designs as reflected by those in NUREG-1150, and for a 4
reactor power level of . 3800 Megawatts (thermal Therefore, the minimum exclusion area distance proposed would assure a).very low level of risk to individuals, even for those located very close to the plant.
Although an exclusion area size of about 0.4 miles is considered appropriate for reactor power levels of current designs, the Commission is also
.considering whether or not this size unduly penalizes potential reactors that have significantly lower power levels and is therefore requesting comments on this. subject.
B. Low Ponulation 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 i
j 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 j be located no closer than one and one-third times the outer radius of the LPZ.
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Finally, the dose to a hypothetical individual located at the outer radius of the '
LPZ over,the entire course of the accident must not be in excess of the dose values given in the regulation. _ Regulatory Guide 4.7 suggests that an. outer <
radius of about 3 miles (4.8 km) for the LPZ has been found to satisfy the dose i values in the present regulation.
Several practical problems have arisen in connection with the LPZ. Before 1980, the LPZ generally defined the distance over which public protective actions were contemplated in the event of a serious accident. The _ regulations in 10 CFR
- 50.47 now requires plume exposure Emergency Planning Zones (EPZ) of about 10 miles for each plant.
The LPZ also places restrictions on the proximity of the nearest densely populated center of 25,000 or more residents. However, without numerical
, requirements for the outer radius of the LPZ, this requirement. has little practical effect. Typical LPZs for existing power reactors have several thousand ,
residents. If Regulatory Guide 4.7 were followed and a distance of 3 miles were selected as the LPZ outer radius, a maximum population within the LPZ at the time of site approval would be about 14,000 residents. Finally, the staff has sometimes experienced difficulty in defining a " densely populated center." ,
The Commission considers that the functions intended for the LPZ, namely, i a low density of residents and the feasibility of taking protective actions, have been accomplished by other regulations or can be ' accomplished by other means.
Protective action requirements are defined via the use of the EPZ, while restrictions on population close to the plant can be assured via proposed population density criteria. For these reasons, the Commission is proposing to ,
eliminate the requirement of an LPZ for future power reactor sites for purposes '
of determining site suitability. '
C. Pooulation Density Criteria. The present regulation .contains no population density requirements other than the requirement, noted above, that the distance to the nearest population center containing more than about 25,000 residents must be no closer than one and one-third times the outer radius of the LPZ. This was recognized as a potential concern when the present regulation was l promulgated. As the Commission noted in its Statement of Considerations on :
April 12,1962 (27 FR 3509), accompanying the issuance of the regulation, "...in some cases where very large cities are involved, the population center distance may have to be greater than those suggested by these guides."
As a result of the significant experience gained in the siting of power reactors, the staff issued Regulatory Guide 4.7 in 1974. With respect to population density this guide states as follows:
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" Areas of low population density are preferred for nuclear power i station sites. High population densities projected for any tioe during the lifetime of a station are considered during both the NRC staff review and the public hearing phases of the licensing process. If the population density at the proposed site is not acceptably low, then the applicant will be required to give special attention to alternative sites with lower population densities. '
If the population density, including weighted transient population, projected at the time of initial operation of a nuclear power station '
exceeds 500 persons per square mile averaged over any radial distance out to 30 miles (cumulative population at a distance divided by the area at -
that distance), or the projected population density over the lifetime of the facility exceeds 1000 persons per square mile averaged over any radial distance out to 30 miles, special attention should be given to the FRN - 7 l
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consideration of alternative sites with lower population densities."
The basis for this guide was that it provided for reasonable separation of. +
reactor sites from large population centers while also assuring an adequate .
selection of sites in all-regions of the nation. However, no comparisons with explicit risk criteria were provided at that time.
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 i
sparsely-populated areas ...has been endorsed in non-binding NRC guidelines.for. '
reactor siting." However, 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 provides population density criteria out to' a distance of 30 miles from the reactor, not 40 miles. 1 i
-An. illustration of the degree of separation distance provided for in this I guide from population centers of various sizes may be useful. 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 l within this distance would yield a total population of about 25,000 persons.
Similarly, a city of 100,000 or more residents should be'no closer than about 10 miles (16 km); a city of 500,000 or more persons should be no closer than about 20 miles (32 km)., and a city of 1,000,000 or more persons should be no closer than about 30 miles (50 km) from the reactor.
The Commission has examined these guidelines with regard to the Safety Goal. The Safety Goal, quantitative health objective in regard to latent cancer fatality states that, within a distance of ten miles (16 km) from the reactor, the risk to the population of latent cancer fatality from nuclear power plant operation, including accidents, should not exceed one-tenth of one percent of the likelihood of latent cancer fatalities from all other causes. In addition to the risks of latent cancer fatalities, the Commission has also investigated the likelihood and extent of land contamination arising from the release of long-lived radioactive species, such as cesium-137, in the event of a severe reactor accident.
.- The results of these analyses indicate that the. latent cancer fatality quantitative health objective noted above is met for current plant designs. From analysis done in support of this proposed change.in regulation, the likelihood ;
of land contamination from a severe accident sufficient to require long term '
condemnation of land beyond 30 miles (50 km) is very low. Other analyses !
indicate that population density restrictions out to 40 miles could make it !
difficult to obtain suitable reactor sites in some regions of the nation.
Because the population density values of Regulatory Guide 4.7 have been in
'use since 1975, and these values afford an adequate supply of potential reactor sites in every region of the nation while providing assurance of low risk of latent cancer fatality as well as land contamination, the Commission considers it prudent to maintain these population density values for future power reactor sites. The Commission wishes to emphasize., however, that nuclear power plants meeting current safety standards could be safely located at sites significantly more dense than 500 people per square mile.
For these reasons, the Commission is proposing that, at the time of initial site approval or early site permit renewal, population density values of no more than 500 people per square mile averaged over any radial distance out to 30 miles .
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be used for judging the acceptability of future nuclear power plant sites. i Similarly, in keeping with Regulatory Guide 4.7, the projected population density 40 years after initial site approval should not exceed 1000 people per square mile.
With regard to the petition by Free Environment, Inc. (PRM-50-20), the Comission concludes that the criteria 'in Regulatory Guide 4.7 provide a reasonable degree of separation for a range of population centers, including
" major" population centers, depending upon their size. Further, codifying the population density criteria of this guide is expected to ensure a low level of risk, including the risk of latent cancer fatality as well as long-term land contamination. Finally, the Comission concludes that granting of the petitioner's request to specify population criteria out to 40 milps rather than 30 miles would not substantially reduce the risks to the public, but could !
significantly increase the difficulty of obtaining suitable reactor sites in some i regions of the nation. For these reasons, the Commission has decided not to adopt the proposal by Free Environment, Incorporated.
An important point regarding population projections and their application !
should be made. Because the validity and reliability of population projections, particularly for relatively small regions, decreases markedly as the projection time period increases, population projections for the purpose of assessing site suitability are to be limited to 40 years. Population projections beyond this time period become unreliable and speculative. The 40 year period for population projections is to be distinguished from the 60 year or more piant lifetime.
Because analyses have shown that current plant designs can meet the Comission's Safety Goals and that other risks can be kept at a very low level at sites that have. significantly higher population densities than those being i proposed, the Comission wishes to emphasize that these population density levels do not 1ndicate the upper limits of acceptability. These levels represent preferred values, that, if exceeded, require that an applicant provide justification for not locating a reactor at an alternative site having a lower ,
l population density. Therefore, the population density limits proposed in the <
j regulation are intended to be used only in the siting decision process to be J i applied at the time of initial site approval or early site permit renewal to l determine whether alternative sites that have lower population densities should be considered. The Comi.ssion does not intend to consider license conditions or ,
operating restrictions upon an operating reactor solely upon the basis that the population density around it may reach or exceed the proposed siting decision ;
values given above during the plant lifetime. Because of the possibility for ~
confusion resulting from numerical values being cited in the regulation, the Comission is also requesting comments on whether numerical population density values should be cited in the regulation or whether these should be stated in a regulatory guide only. The Comission is also requesting coments on'whether the values of 500 and 1000 persons per square mile are appropriate, and whether '
population density criteria need be specified out to 30 miles, or whether another distance is more appropriate.
D. Meteoroloaical Factors. Radiological doses that incorporate site !
meteorological data need no longer be calculated for the purpose of determining site suitability. Meteorological data will still be needed for safety analysis and for assessing the adequacy of certain plant features, as well as to determine plant adequacy in regard to meteorological extremes, such as tornados and maximum l probable precipitation. Therefore, the proposed regulation maintains the l requirement to collect and characterize meteorological data representative of the L site. <
l The Comission has examined the variations in site meteorology that have !
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< o influenced dose - calculations in past licensing reviews. Individual ' site meteorology characteristics have been used primarily to. determine atmospheric
. dispersion or dilution factors in order to evaluate doses to hypothetical individuals at the exclusion area and LPZ outer radius. The degree of dilution increases with ' increasing distance between the release point and any hypothetically exposed individual, but it also is affected by other factors, including the time of day. In this regard, the dispersion factor could vary significantly at a given site and show a pronounced diurnal variation. However, when the time-averaged dispersion factor of a given site is compared with that of other sites, the variation between one site and another is much less.
Analyses reported in NUREG/CR-2239, " Technical Guidance- for Siting Criteria Development," dated December 1982, for example, show that calculated average individual consequences for an identical postulated release of radioactivity to the environment using data from yeather stations throughout the United States yielded results that varied only by about a factor of two. Based upon these considerations, the Comission has determined that the average meteorological characteristics between one site and another 'are sufficiently similar that characterization of individual site meteorology is not a significant discriminator in determining site suitability when compared to the uncertainties in other areas of the determination of risk to the health and safety to the public.
However, site meteorological characteristics are needed in safety analysis and for assessing the adequacy of certain plant design features.
E.. Hydroloaical Factors. These factors are important in establishing the magnitude of external hazards from ground-water contamination, such as by containment affect large basemat melt through, which could contaminate aquifers and thereby populations. The proposed regulation adds or modifies existing requirements for obtaining information a site important to risk. This information.will to characterize hydrological factors at be reviewed by the staff and used as interface criteria in matching a proposed. design to the site.
F. Nearby Industrial and Transoortation Facilities. This area of review would be incorporated into the regulations for determining site suitability.
This area of review has, in fact, been a part of the NRC review for many years.
The proposed regulation' involves no substantive changes in this area and merely codifies what has been NRC practice for a number of years.
G. Feasibility of Carrvina out Protective Actions. The proposed regulation would require that important site factors such as population distribution, topography, and transportation routes be considered and examined in order to determine whether there are any site characteristics that could pose a significant impediment to the development of an emergency plan.
approach. Planning for emergencies is part of the Commission's defense-in-depth The Commission has concluded that site characteristics that may represent an impediment to the development of adequate emergency plans, such as limitations of access or egress in the imediate vicinity of a nuclear power plant, should be identified at the site approval phase. This is consistent with the approach the Commission has taken in early site reviews under 10 CFR Part 52.
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H. Periodic Reportina of Man-Related Activities. Conditions around a site may change and significant changes in the nature of the industrial, military, and transportation facilities may occur. Early identification of activities or facilities that are potentially hazardous could permit timely changes in the procedures or plant features to minimize the change in the risk to the health and FRN - 10
I safety of the public. Man-related activities potentially hazardous to a plant are typically large major pipelines, industrial major or transport airports, facilities such as major highways, etc. Relativel i activity have been shown to be of little concern.y minor changes in industrial The Commission is considering whether periodic reporting of significant offsite activities should be required and is requesting comments on whether significant offsite facilities within five miles of the reactor should be periodically updated every five years.
Interim Change to 10..CFR Part 50 1
The proposed change to 10 CFR Part 50 would simply relocate from 10 CFR Part 100 the requirements for each applicant to calculate a whole body and a thyroid dose at specified distances. Because these requirements affect reactor design rather than siting, they are more appropriately located in 10 CFR Part 50.
For this proposed revision, the source term and methodolor,y for performing the dose calculations.would remain unchanged from the current requirements.
These requirements would continue to apply to future applicants for a -i 1
construction permit, design certification, or an operating license, but are ,
intended to be interim requirements until such time as more specific requirements l
are insights. developed regarding revised accident source terms and severe accident
! B Seismic and Earthquake Engineering Criteria.
The folloafng major changes in the proposed revision to Appendix A, !
" Seismic and Geologic Siting Criteria for Nuclear Power Plants," to Part 100, are associated with the proposed seismic and earthquake engineering criteria i rulemaking: )
! 1. Seoarate Sitina from Desian.
!' 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 l
l revised criteria presented in the proposed regulation will not be applied to existing plants, the licensing basis for existing nuclear power plants must remain part of the regulations. The criteria on seismic and geologic siting would be designated as a new Appendix B, " Criteria for the Seismic and Geologic Siting of Nuclear Power Plants After (Effective Date of the i Regulation]," to 10 CFR Part 100. Criteria on earthquake engineering would be l
designated as a new Appendix S, " Earthquake Engineering Criteria for Nuclear p Power Plants," to 10 CFR Part 50.
! 2. Remove Detailed Guidance from the Reaulation. The current regulation contains both requirements and guidance on how to satisfy the requirements. For example,Section IV, " Required Investigations," of Appendix A,-states that investigations are required for vibratory ground motion, l surface faulting, and seismically induced floods and water waves. Appendix A then provides detailed guidance on what constitutes an acceptable investigation. A similar situation exists in Section V, " Seismic and Geologic Design Bases," cf Appendix A.
Geoscience assessments require considerable latitude in judgment. This ,
l latitude in judgment is needed because of limitations in data and the state-i 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 FRN - 11
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existing regulation was developed. The existing regulation states that it is ;
i based on limited geophysical and geological information and will be revhed 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 i 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 level of detail presented in the proposed regulation would be !
reduced considerably. The proposed regulation would. identify and establish -
basic requirements. Detailed guidance, that is, the procedures acceptable to the NRC for meeting the requirements, would be contained in a draft regulatory '
guide to be issued for public comment as Draft Regulatory Guide, DG-1015,
" Identification and Characterization of Seismic Sources, Deterministic Source Earthquakes, and Ground Motion."
- 3. Use of Both Deterministic and Probabilistic Evaluations. The proposed regulation would require-the use of both probabilistic and.
deterministic evaluations. 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 j 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 l 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 uncertainty in geoscience parameter. Because so little is known about - ;
earthqutke phenomena (especially in the eastern United States), there have '
always been differences of opinion among experts as to how the prescribed process in Appendix A is to be carried out. Experts often delineate.very i 1
different estimates of the largest earthquakes to be considered and different i ground-motion models. i Over the past decade, analysis methods for encompassing these '
differences 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 i of using these probabilistic methods is their ability to not only incorporate i different models and different data sets, but also to weight them using i judgments as to.the validity of the different models and data sets, and ;
thereby to provide an explicit expression for the overall uncertainty in the ground motion estimates and a means of assessing sensitivity to various input parameters, _
i Probabilistic methods have been used by many groups, not only in the seismic-hazard area but in many other areas. In the seismic-hazard area, many i of the practitioners participated in either the NRC-Lawrence Livermore '
National Laboratory (LLNL) or the Electric Power Research Institute (EPRI) seismic-hazard projects over the past decade.
The advantages of these probabilistic methods are manifest. However, their limitations are important too. In the seismic-hazard area, the most important limitation is that the " bottom-line" results from these analyses .
tend to be dominated by the tails rather than the central tendencies of the distributions of knowledge and expert opinion.
For these reasons, the proposed revision of Appendix A to 10 CFR Part ;
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100 has adopted an approach using both probabilistic and deterministic -
evaluations. The staff proposes to use both the deterministic (currently being used) and the probabilistic evaluations together and compare the results of each to provide insights unavailable if either method were used alone. The principal limitations of the deterministic evaluation --- its ability' to incorporate only one model and-one data. set at a time and its inability to allow weighted incorporation of numerous models --- can be assessed by
- comparing its results with the results of a probabilistic evaluation l accomplished in parallel. Similarly, the principal limitation of the probabilistic evaluation --- its tendency to allow its results to be dominated by the tails rather.than the central tendency of distributions of uncertain knowledge or expert opinion --- can be assessed by comparing its results with t the results of one or more deterministic evaluation. ;
l The NRC believes that taken together,.this approach can allow more '
l informed judgments as to what the appropriate Safe Shutdown Earthquake Ground l
Motion.should be for a given site. .Both the applicant's judgments and those ;
of the NRC will be improved. Therefore, the NRC believes that this approach j is the best way to accomplish the objective of this aspect of the revised regulation and arrive, through analysis, at a site-specific ground motion that l
. appropriately captures what is known about _the seismic regime. Using both '
probabilistic and deterministic evaluations will lead to a more stable and predictable licensing process than in the past.
In order to implement this approach, the NRC has proposed a requirement l that the probability of exceeding the Safe Shutdown Earthquake Ground Motion
! at a site be lower than the median probability of exceedance computed for the i 1
current population of the operating plants. This requirement assures that the ;
design level _s at new sites will be comparable to those at many existing sites, )
particularly more recently licensed sites. This criterion is also used to '
identify significant seismic sources, in terms of magnitude and distance, affecting the estimates of ground motions at a site.
I
- 4. Safe Shutdown Earthauake. The existing regulation (10 CFR Part 100, Appendix A, Section V(a)(1)(v)) states that when the maximum vibratory accelerations of the Safe Shutdown Earthquake at the foundations of the nuclear power plant structures are determined to be less than one tenth the acceleration of gravity (0.1 g) ..... it shall be assumed that the maximum vibratory accelerations of the Safe Shutdown Earthquake at these foundations
! are. at least 0.1 g (Also,Section V(a)(1)(iv) contains the phrase "at each of the various foundation locations") The location of the seismic input motion control point as stated in the existing regulation has led to confrontations with many applicants that believe this stipulation is inconsistent with good engineering fundamentals.
The proposed regulation would move the location of the seismic input motion control point from the foundation-level to free-field, at the free ground surface or hypothetical rock outcrop, as appropriate. The 1975 version of.the Standard Review Plan placed the control motion in the free-field. The proposed regulation is also consistent with the resolution of Unresolved Safety Issue (USI) A-40, " Seismic Design Criteria" (August 1989), that resulted in.the revision of Standard Review Plan Sections 2.5.2, 3.7.1, 3.7.2, i and 3.7.3. '
- 5. Value of the Operatina Basis Earthauake Ground Motion (OBE) and Reauired OBE Analyses. The existing regulation (10 CFR, Appendix A,Section V(a)(2)) states that the maximum vibratory ground motion of the OBE is one-
' FRN - 13 4
half th'e maximum vibratory ground motion of the Safe Shutdown Earthquake ground motion. Also, the existing regulation (10 CFR, Appendix A,Section VI(a)(2)) states ~that the engineering method used to' insure that structures,
. systems, and components are capable of withstanding the effects of the OBE shall involve the.use of either a suitable dynamic analysis or a suitable qualification test. In some cases, for instance piping, these multi-facets of the OBE in the existing regulation made it possible for the OBE to have more design significance than the SSE. A decoupling of the OBE-and SSE has been suggested in several documents. For instance, the NRC staff, SECY-79-300, suggested that design for a single limiting event and inspection and '
evaluation for earthquakes in excess of some specified limit may be the most sound regulatory approach. NUREG-1061, " Report-of the U.S. Nuclear Regulatory Commission Piping Review Committee," Vol.5, April 1985, (Table 10.1) ranked a decoupling of'the OBE and SSE as third out of six high priority changes. In SECY-90-016, "Evolut;onary 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.
For the evolutionary reactors, the NRC will consider requests to decouple the OBE from the SSE on a design-specific basis.
Activities equivalent to OBE-SSE decoupling are also being done in foreign countries. For instance, in Germany their new design standard-requires only one design basis earthquake (equivalent to the SSE). They require an inspection-level earthquake (for shutdown) of 0.4 SSE. This level was set so that the vibratory ground motion should not induce stresses .
exceeding the allowable stress limits originally required for the OBE design.
The proposed regulation would allow the value of the OBE to be set at (i) one-third or.less of the SSE, where OBE requirements are satisfied without an explicit response or design analyses being performed, or (ii) a value greater than one-third of the SSE, where analysis and design are required.
There are two issues the applicant should consider in selecting the value of the OBE: first, plant shutdown is required if vibratory ground motion exceeding that of the OBE occurs (discussed below in Item 6, Required Plant Shutdown), 'and second, the amount of analyses associated with the OBE. An applicant may determine that at one-third of the SSE level, the probability of exceeding the OBE vibratory ground motion is too high, and the cost associated with plant shutdown for inspections and testing of equipment and structures prior to restarting the plant is unacceptable. Therefore, the applicant may voluntarily select an OBE value at some higher fraction of the SSE to avoid plant shutdowns. However, if an applicant selects an OBE value at a fraction
- of the SSE higher than one-third, a suitable analysis shall be performed to demonstrate that the requirements associated with the OBE are satisfied. The design shall take into account soil-structure interaction effects and the expected duration of the vibratory ground motion. The requirement associated with the OBE is that all structures, systems, and components of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the public shall remain functional and within applicable stress and deformation limits when subjected to the effects of the OBE in combination with normal operating loads. As stated above, subject to further confirmation,-it is determined that if an OBE of one-third of the SSE is used, the requirements of the OBE can be satisfied without the applicant performing any explicit response analyses. However, some minimal design checks (additional discussion below) must be performed. 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. In'this case, the OBE serves the function of an inspection and FRN - 14
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e s shutdown earthquake. There are situations associated with current analyses i where only OBE is associated with the design requirements, for example, the I ultimate heat sink (see Regulatory Guide 1.27, " Ultimate Heat Sink for Nuclear Power Plants"). . In these situations, a value expressed as a fraction of the SSE response would be used in the analyses..Section VIII of this Proposed rule identifies existing guides that would be revised, technically to maintain the existing design philosophy. With regard to piping analyses, positions on fatigue ratcheting and seismic anchor motion are being developed and will be issued for public comment'in a draft regulatory guide separate from this rulemaking.
- 6. Reauired' 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 is required. The supplementary information to the final regulation published (November 13, 1973, 38 FR 31279, Item 6e) includes the following statement: "A footnote has been added to 150.36(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 l
risk to the health and safety of. the public." At that time, it was the intention of the Commission to treat the Operating Basis Earthquake as a limiting condition of operation. From the statement in the Supplementary l 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 j expected to be implemented by being included among the technical ;
specifications submitted by applicants after the adoption of Appendix A. In fact, applicants did not include OBE shutdown requirements in their technical specifications.
The proposed regulation would treat plant shutdown associated with ,
vibratory ground motion exceeding the OBE or significant plant damage as a !
condition in every operating license. The shutdown requirement would be a !
condition of the license (10 CFR 50.54) rather than a limiting condition of '
operation (10 CFR 50.36), because the necessary judgments associated with exceedance of the vibratory ground motion or significant plant damage can not }
be adequately characterized in a technical specification. A new paragraph :
650.54(ee) would be added to the regulations to require plant shut down for l licensees of nuclear power plants that comply with the earthquake engineering i criteria in Paragraph IV(a)(3) of Proposed Appendix S, " Earthquake Engineering ;
l Criteria for Nuclear Power. Plants," to 10 CFR Part 50. Draft Regulatory Guide '
DG-1017, " Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator ;
Post-Earthquake' Actions," is being developed to provide guidance acceptable to '
the NRC staff for determining whether or not vibratory ground motion exceeding the OBE ground motion.or significant plant damage had occurred and nuclear i power plant shut down is required. The guidance is based on criteria !
developed by the Electric Power Research Institute (EPRI). Draft Regulatory :
Guide DG-1018, " Restart of a Nuclear Power Plant Shut Down by a Seismic '
l Event," is being developed to provide guidelines that are acceptable to the :
i NRC staff for performing inspections and tests of nuclear power plant i equipment and structures prior to plant restart. This guidance is also based :
on EPRI reports. ,
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t i
-7 3 Clarify Interoretations. In Appendix B to 10 CFR Part 100, changes have been made to resolve questions of interpretation. As an example, j definitions and required investigations stated in the proposed regulation ;
would be significantly changed to eliminate or modify phrases that were more applicable to only the western part of the United States. !
i VI. Siting Policy Task Force Recommendations
'i The Siting Policy Task Force made nine recommendations with regard to !
revision of the reactor siting criteria in NUREG-0625, " Report of the Siting :
Policy Task Force," August 1979. The individual recommendations and the disposition and actions being taken in regard to each of these are discussed below.
Recommendation 1.
Revise Part 100 to change the way protection is provided for accidents )
by incorporating a fixed exclusion area and protection action distance and population density and distribution criteria.
)
- 1. Specify a fixed minimum exclusion distance based on limiting the individual risk from design basis accidents. Furthermore, the regulations should clarify the required control by the utility over activities taking place in land and water portions of the exclusion area.
- 2. .Specify a fixed minimum emergency planning distance of 10 miles.
The physical characteristics of the emergency planning zone should provide reasonable assurance that evacuation of persons, including transients, would be feasible if needed to mitigate the consequences of accidents.
- 3. Incorporate specific population density and distribution limits outside the-exclusion area that are dependent on the average population of the region.
- 4. Remove the requirement to calculate radiation doses as a means of establishing minimum exclusion distances and low population zones.
Disposition and Action.
Recommendation I has been or is largely being adopted by the Commission.
With regard to item 1, a fixed minimum exclusion area distance of 0.4 mile, commensurate with past NRC experience in the review of design basis accidents, is being proposed. The Commission believes that the existing requirements regarding control over any land portion of the exclusion area together with ,
current emergency planning requirements make any new requirements on exclusion area control unnecessary. The recommendations in item 2 were adopted by the Commission shortly after the Three Mile Island accident and are contained in .
10 CFR 50.47. The recommendations in item 3 are being adopted except that the !
population density and distribution limits are proposed to be applicable '
nationwide. The recommendation of Item 4 is being adopted.
Recommendation 2. .
Revise 10 CFR Part 100 to require consideration of the potential hazards 3 posed by man-made activities and natural characteristics of sites by !
establishing minimum standoff distances for: l
- 1. Major or commercial airports,
- 2. Liquid Natural Gas (LNG) terminals,
- 3. Large propane pipelines,
- 4. Large natural gas pipelines, ,
i
- 5. Large quantities of explosive or toxic materials, FRN - 16 i
-. .- .= _- -_ . . - .
e .
- 6. Major dams, and l
- 7. Capable faults.
Disoosition and Action. l Recommendation 2 is being adopted in part and rejected in part. 10 CFR Part 100 is to be revised to include consideration of man--related hazards.
However, establishing minimum standoff distances by regulation for the hazards cited is not feasible. NRC review has found that acceptable separation distances are not readily quantified and can depend upon many other factors such as the topography, size, and operational aspects of the facilities, in addition to the distance from the reactor. Accordingly, the proposed regulation will require that the hazards be identified and evaluated so that they can be adequately considered in the design of the reactor to be located on the site. Present NRC review criteria, as given in the Standard Review Plan (SRP), Section 2.2.3, are considered adequate.
Recommendation 3.
Revise 10 CFR Part 100 by requiring a reasonable assurance that interdictive measures are possible to limit groundwater contamination resulting from Class 9 accidents within the immediate vicinity of the site.
Disposition and Action.
The Commission is not adopting this recommendation. However, requirements on future reactor designs will address the need to consider and minimize containment failure under severe accident conditions. Future reactor designs will need to address the potential for ground water contamination as part of their environmental review under 10 CFR Part 51.
Recommendation 4.
Revise Appendix A to 10 CFR Part 100 to better reflect the evolving technology in assessing seismic hazards.
Disposition and Action.
The Comission is proposing to adopt this recommendation in this rulemaking.
Recommendation 5.
Revise 10 CFR Part 100 to include consideration of post-licensing changes in offsite activities.
- 1. The NRC staff shall inform local authorities (planning commission, county commissions, etc.) that control activities within the emergency planning zone (EPZ) of the basis for determining the acceptability of a site.
- 2. The NRC staff shall notify those Federal agencies as in item 1 above that may reasonably initiate a future Federal action that may influence the
, nuclear power plant.
- 3. The NRC staff shall require applicants to monitor and report potentially adverse offsite developments.
- 4. If, in spite of the actions described in items 1 through 3, there are offsite developments that have the potential for significantly increasing the risk to the public, the NRC staff will consider restrictions on a case-by-case basis.
Disoosition and Actign.
This recommendation is already in effect or being adopted. Item I is already covered by existing emergency planning requirements. Item 2 is being accomplished by issuance of a Significant Hazard Consideration statement by the NRC staff. The Commission is requesting comments on Item 3. With regard to item 4, the Commission retains the right to order restrictions on a l
l FRN - 17 1
, e case--by--case basis.
Recommendation 6.
Continue the current approach relative to site selection from a safety viewpoint, but select sites so that there are no unfavorable characteristics requiring unique or unusual design to compensate for site inadequacies.
Disposition and Action.
The Commission is not adopting this recommendation. In the current and proposed Part 100 regulations, applicants may provide specific plant design features to compensate for site inadequacies. As long as these design features adequately account for the conditions at the site, public health and safety will be protected. These specific design features may involve added costs. However, the Commission has concluded that any economic consideration should be left to the applicant.
Recommendation 7. -
Revise Part 100 to specify that site approval be established at the earliest decision point in the review and to provide criteria that would have to be satisfied for this approach to be subsequently reopened in the licensing process.
Disposition and Action.
The Commission considers that the early site permit provisions of 10 CFR Part 52 accomplish this recommendation.
Recommendation 8.
Revise 10 CFR Part 51 to provide that a final decision disapproving a proposed site by a state agency whose approval is fundamental to the project would be a sufficient basis for NRC to terminate review. The termination of a review would then be reviewed by the Commission.
Disposition and Action.
The Commission is not adopting this recommendation because it is considered inappropriate. This recommendation would give a State the authority to grant issuance of a construction permit for a nuclear facility.
Only the Federal government has this authority. States do have an independent right to deny site approval as long as it is not a radiological health and safety, common defense, or security concern.
Recommendation 9.
Develop common bases for comparing the risks for all external events.
Disoosition and Action.
The Siting Policy Task Force's primary recommendation in this area was that an interdisciplinary effort should be undertaken with the objective of developing quantitative risk comparisons of all external events and natural Phenomena. The Commission considers this to be a desirable objective but notes that the Siting Policy Task Force made no specific recommendations with regard to siting criteria or rulemaking. The Commission therefore considers this recommendation inapplicable in the present context of examination of
. siting criteria, but notes that recent developments in probabilistic risk analysis (PRA) have considered examination of the risk from external events in detail.
VII. Related Regulatory Guides and Standard Review Plan Section The NRC is developing the following draft regulatory guides and standard FRN - 18
e s '
review plan section to provide prospective licensees with the necessary guidance for implementing the proposed regulation. The notice of availability for these materials is published elsewhere in this issue of the Federal Register.
- 1. DG-1015, " Identification and Characterization of Seismic Sources, Deterministic Source Earthquakes, and Ground Motion." The draft guide provides general guidance and recommendations, describes acceptable procedures j and provides a list of references that present acceptable methodologies to l identify and characterize capable tectonic sources and seismogenic' sources.
- 2. DG-1016, Second Proposed Revision 2 to Regulatory Guide 1.12,
" Nuclear Power Plant Instrumentation for Earthquakes." The draft guide describes seismic instrumentation type and location, operability, I characteristics, installation, actuation, and maintenance that are acceptable l to the NRC staff.
- 3. DG-1017, " Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Post-Earthquake Actions." The draft guide provides guidelines that are acceptable to the NRC staff for a timely evaluation of the recorded seismic instrumentation data and to determine whether or not plant shutdown is required.
- 4. DG-1018, " Restart of a Nuclear Power Plant Shut Down by a Seismic Event." The draft guide provides guidelines that are acceptable to the NRC staff for performing inspections and tests of nuclear power plant equipment and structures prior to restart of a plant that has been shut down because of a seismic event.
- 5. Draft Standard Review Plan Section 2.5.2, Proposed Revision 3
" Vibratory Ground Motion." The draft describes procedures to assess the ground motion potential of seismic sources at the site and to assess the ,
adequacy of the SSE. l
- 6. Draft Regulatory Guide 4.7, Revision 2, dated December 1991, " 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.
VIII. 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 to coincide with the publication of the final regulations that would implement this proposed action.
The following regulatory guides will be revised to incorporate editorial changes, for example to reference new paragraphs in Appendix B 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."
FRN - 19
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- 6. 1.102, " Flood Protection for Nuclear Power Plants." -
- 7. 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes."
- 8. 1.122, " Development of Floor Design-Response Spectra for Seismic Design of Floor-Supported Equipment or Components."
The following regulatory guides will be revised to update the design or analysis philosophy, for example, to change OBE to a fraction of the SSE:
- 1. 1.27, " Ultimate Heat Sink for Nuclear Power Plants"
- 2. 1.100, " Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power Plants" ,
- 3. 1.124, " Service Limits and Loading Combinations for Class 1 Linear-Type Component Supports"
- 4. 1.130, " Service Limits and Loading Combinations for Class 1 Plate- and-Shell-Type Component Supports" 1
- 5. 1.132, " Site Investigations for Foundations of Nuclear Power l Plants"
- 6. 1.138, " Laboratory Investigations of Soils for Engineering Analysis and Design of Nuclear Power Plants" 7, 1.142,. " Safety-Related Concrete Structures for Nuclear Power 8.
Plants (Other than Reactor Yessels and Containments)"
1.143, " Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants" Minor and confonning changes to other Regulatory Guides and standard ruiew plan sections as a result of proposed changes in the nonseismic criteria are also planned. If substantive changes are made during the revisions, the applicable guides will be issued for public comment as draft guides.
IX. Referenced Documents An i'terested n person may examine or obtain.co referenced in this proposed rule as set out below. pies of the documents Copies of NUREG-0625, NUREG-1150, and NUREG/CR-2239 may be purchased from the Superintendent of Documents, U.S. Government Printing Office, P.O.
Box 37082, Washington, DC 20013-7082. Copies are also'available from the National Technical Information Service, 5285 Port Royal Road, Springfield, VA 22161. A copy is also available for inspection and copying for a fee in the
.NRC Public Document Room, 2120 L Street, NW. (Lower Level), Washington, DC.
Regulatory Guides 1.27 and 4.7 are available for inspection and copying for a fee at the Commission's Public Document Room, 2120 L Street, NW. (Lower level), Washington, DC. Copies of-issued guides may be purchased from the Government Printing Office (GP0) at the current GPO price. Information on current GP0 prices may be obtained by contacting the Superintendent of Documents, U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20013-2171. Issued guides may also be purchased from the National Technical Information Service on a standing order basis. Details on this service may be obtained by writing NTIS, 5826 Port Royal Road, Springfield, VA ;22161.
SECY 79-300, SECY 90-016, and WASH-1400 are available for. inspection and copying for a fee at the Commission's Public Document Room, 2120 L Street, FRN - 20
- y.
- I i NW. (Lower Level), Washington, DC. -
1 X. Electronic Format for Submittal of Public Comments i
The comment resolution process will be improved if each comment is J identified with the document title, section heading, and paragraph number '
addressed. Commenters may submit, in addition to the original paper copy, a l
' copy of the letter in an electronic format on IBM PC DOS compatible 3.5 or 5.25 inch double sided double density (DS/DD) diskettes. Data files should be .
l provided in Wordperfect 5.1 format. ASCII code is also acceptable or if
' formatted text is required, data files should be provided in IBM Revisable - a Form Text Document Content Architecture (RFT/DCA) format. i XI. Questions In addition to soliciting comments on all aspects of this rulemaking, the Commission specifically requests comments on the following questions.
A Nonseismic Criteria.
1.
Should the exclusion area distance be smaller than 0.4 mile (640 meters) for plants having reactor power levels significantly less than i 3800 Megawatts (thermal) and should the exclusion rea distance be allowed to i vary according to power level with a minimum value (for ample,' O.25 miles or l E400 meters)?
, 2. The Commission proposes to codify the population density i
guidelines in Regulatory Guide 4.7 which states that the population density should not exceed 500 people per square mile out to a distance .of 30 miles at the time of site approval and 1000 people per square mile 40 years thereafter.
l Comments are specifically requested on questions 2A, 28, and 2C given below.
A. Should numerical values of population density appear in the regulation or should the regulation provide merely general guidance, with numerical values provided in a regulatory guide?
B. Assuming numerical values are to be codified, are the values of l
500 persons per square mile at the time of site approval and 1000-persons per square mile 40 years thereafter appropriate? If not, what other numerical l values should be codified and what is the basis for these values?
C. Should population density criteria be specified out to a distance other than 30 miles (50 km), for example, 20 miles (32 km)? If a different distance is recommended, what is its basis?
- 3. Should the Commission approve sites that exceed the proposed population values of 10'CFR Part 100.21, and if so, under what conditions?
- 4. Should holders of early site permits, construction permits, and operating license permits be required to periodically report changes in potential offsite hazards (for example, every 5 years within 5 miles)? If so, '
what regulatory purpose would such reporting requirements serve?
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- 5. What continuing regulatory significance should the safety j requirements in 10 CFR Part 100 have after granting the initial operating license or combined operating license under 10 CFR Part 527 l
- 6. Are there certain site meteorological conditions _that should i preclude the siting of a nuclear power plant? If so, what are the conditions :
that can not be adequately compensated for by design features?
- 7. In the description of the disposition of the recommendations of j the Siting Policy Trsk Force report-(NUREG-0625), it was noted that the Commission was not adopting every element of each recommendation. Are there compelling reasons to reconsider any recommendation not adopted and, if so, ,
what are the bases for reconsideration? !
1 B Seismic and Earthquake Engineering Criteria.
l The proposed guide, DG-1015, outlines concepts and procedures to be used in conjunction with the probabilistic/ deterministic seismic hazard ;
evaluations. Rationale for the approach is discussed in Section V.B(3) of l this Proposed rule.
The NRC is currently performing confirmatory studies to evaluate and refine these proposed procedures. A limited study has been completed j demonstrating the feasibility of procedures and the validity of the concepts. ;
However, the staff would like to solicit comments on the concepts outlined in the proposed guide at this time. To facilitate the review, results of the i
application of the proposed procedure to four test sites are published separately (Letter report from D. Bernreuter of LLNL to A. Murphy of NRC).
There are divergent views on the role probabilistic seismic hazard analysis should play in the licensing arena. There is a general consensus within the NRC staff that the revised seismic and geological siti% criteria i should allow considerations for a probabilistic hazard analysis. There is l also a general belief that the probabilistic analysis should be calibrated ;
against the past practices for siting and licensing the current generation of ;
nuclear power plants. There is a general consensus that ground motions should ;
be calculated using deterministic methods once the controlling earthquakes are !
determined. With regard to the role c7 the probabilistic analysis, views range from an advocacy of a predomintntly probabilistic analysis to the probabilistic/ deterministic approa6 proposed here to the currently used "
predominantly deterministic approrch. Given these divergent views, the NRC would like to invite comments regarding the use of probabilistic seismic ,
hazard analysis and the balance between the deterministic and probabilistic evahationn. This and other associated issues are itemized below. (As the ~
detailed technical studies are completed some of the staff positions may be confirmed, but specific comments would be helpful at this time.) j
- 1. In making use of both deterministic and probabilistic evaluations, '
l how should they be combined or weighted, that is, should one dominate over the other? (The NRC staff believes that deterministic investigation and their use in the development and evaluation of the Safe Shutdown Earthquake Ground Motion will remain an important aspect of the siting regulations for nuclear power plants for the foreseeable future. The NRC staff also believes that probabilistic seismic hazard assessment methodologies have reached a level of maturity to warrant a specific role in siting regulations.)
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-., - - - . - - - . . - . - . . - - - - - - - . - . - - - - - . =
- 2. .In making use of the probabilistic and deterministic evaluations as proposed in Draft Regulatory Guide DG-1015, is the proposed procedure in Appendix C to DG-1015 adequate to determine controlling earthquakes from the probabilistic analysis? ,
- 3. In determining the controlling earthquakes, should the median !
values of the seismic hazard analysis, as described in Appendix C to Draft l Regulatory Guide DG-1015, be used to the exclusion of other statistical ;
measures,such as, mean or 85th percentile? (The staff has selected t probability of exceedance levels associated with the median hazard analysis .
estimates as they provide more stable estimates of controlling earthquakes.) '
be raised or lowered, that is, should the next generation of nuclear power plants have design levels for seismic events approximately equal to, greater !
than, or less than the current nuclear power plants?
- 5. The proposed Appendix B has included a criterion that states: "the probability of exceeding the Safe Shutdown Earthquake Ground Motion-is l considered acceptably low if it is less than the median probability computed !
I from the current [ EFFECTIVE DATE OF THE REGULATION] population of nuclear ;
i power _ plants". This is a relative criterion without any specific numerical l
value of the probability of exceedance. Because of the current- status of the )
l j probabilistic seismic hazard analysis, method dependent probabilities or
' i target levels are identified in the proposed regulatory guide. Comments are l solicited as to whether the above criterion, as stated, needs to be included in the regulation and, if not, should it be included in the regulation in a different form (e.g., a specific numerical value).
- 6. For .he probabilistic analysis, how many controlling earthquakes should be generated to cover the frequency band of concern for nuclear power l
i plants? (For the four trial plants used to develop the criteria presented in Draft Regulatory Guide DG-1015, the average of results for the 5 Hz and 10 Hz spectral velocities was used to establish the probability of exceedance level.
Controlling earthquakes were evaluated for this frequency band, for the average of I and 2.5 Hz spectral responses, and for peak ground acceleration.)
.XII. Finding of No Significant Environmental Impact: Availability l The Commission has determined under the National Environmental Policy
' Act of 1969, as amended, and the Commission's regulations in Subpart A of 10 CFR Part 51, that this proposed regulation, if adopted, would not be a major Federal action significantly affecting the quality of the human environment and therefore an environmental impact statement is not required.
i 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 l NRC has concluded that relocating the requirement for a dose calculation to ;
Part.50 and adding more specific site criteria to Part 100 does not decrease ;
the protection of the public health and safety over the current regulations. '
The proposed amendments do not affect nonradiological plant effluents and have j no other environmental impact.
, The addition of Appendix B to 10 CFR Part 100, and the addition of l Appendix S to 10 CFR Part 50, will not change the radiological environmental l impact offsite. Onsite occupational radiation exposure associated with 1 j FRN - 23
-.,m ,. ..n. , - , .., _ . _ . , . . . _ , . ,
. c, 1
l inspection and maintenance will not change. These activities are principally associated with base line inspections of structures, equipment, and piping, and with maintenance of seismic instrumentation. Base line inspections are needed to differentiate between pre-existing conditions at the nuclear power plant and earthquake related damage. The structures, equipment and piping selected for these inspections are those routinely examined by plant operators during normal plant walkdowns and inspections. Routine maintenance of seismic instrumentation ensures its operability during earthquakes. The location of the seismic instrumentation is similar to that in the existing nuclear power plante. The proposed amendments do not affect nonradiological plant effluents and have no other environmental impact.
The environmental assessment and finding of no significant impact on which this determination is based are available for inspection at the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC. Single copies of the environmental assessment and finding of no significant impact are available from Mr. Leonard Soffer, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 492-2916, or Dr. Andrew Murphy, Office of Nuclear Regulatory Research, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 492-3860.
XIII. Paperwork Reduction Act Statement This proposed regulation amends information collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.).
This proposed regulation has been submitted to the Office of Management and Budget for review and approval of the paperwork requirements.
There is no public reporting burden related to the nonseismic siting criteria. Public reporting burden for the collection of information related to the seismic and earthquake engineering criteria is estimated to average 800,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> per response, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information.
Send comments regarding this burden estimate or any other aspect of this collection of information, including suggestions for reducing this burden, to the Information and Records Management Branch (MNBB 7714), U.S. Nuclear Regulatory Commission, Washington, DC 20555; and to the Desk Officer, Office of Information and Regulatory Affairs, NE0B-3019, (3150-0011 and 3150 - 0093), Office of Management and Budget, Washington, DC 20503.
XIV. Regulatory Analysis The Commission has prepared a draft regulatory analysis on this proposed regulation. The analysis examines the costs and benefits of the alternatives considered by the Commission. The draft analysis is available for inspection in the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC. Single copies of the analysis are available from Mr. Leonard Soffer, l Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 492-3916, or Dr. Andrew J. Murphy, Office of Nuclear Regulatory Research, U.S. Hud ear Regulatory Commission, Washington, DC 20555, telephone (301) 492-3860.
. The Commission requests public comment on the draft regulatory analysis.
Comments on the draft analysis may be submitted to the NRC as indicated under the ADDRESSES heading.
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< w l
Regulatory Flexibility Certification l
XV. 1 In accordance with the Regulatory Flexibility Act of 1980 (5 U.S.C.
605(b)), the Commission certifies that this proposed regulation will not, if 7
promulgated, have a significant economic impact on a substantial number of ,
- small entities. This proposed regulation affects only the licensing and j 4 operation of nuclear power plants. Nuclear power. plant site applicants do not i fall within the definition of small businesses as defined in Section 3 of the !
Small Business Act (15 U.S.C. 632), the Small Business Size Standards of the i Small Business Administrator (13 CFR Part 121), or the Commission's Size l Standards (56 FR 56671; November 6, 1991).
XVI. Backfit Analysis The NRC has determined that the backfit rule, 10 CFR 50.109, does not I apply to this proposed regulation, and therefore, a backfit analysis is not required for this proposed regulation because these amendments do not involve any provisions that would impose backfits as defined in 10 CFR 50.109(a)(1).
The proposed regulation would apply only to applicants for future nuclear power plant construction permits, preliminary design approval, final design approval, manufacturing' licenses, early site reviews, operating licenses, and combined operating licenses.
List of Subjects 10 CFR Part 50 - Antitrust, Classified information, Criminal penalty, Fire protection, Incorporation by reference, Intergovernmental relations, 1 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, l 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 i design certification.
10 CFR Part 100 - Nuclear power plants and reactors, Reactor siting criteria.
For the reasons set out in the preamble and under the authority of the !
Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 553, the NRC is proposing to adopt the following <
amendments to 10 CFR Parts 50, 52 and 100. )
PART 50 - DOMESTIC LICENSING OF l' PRODUCTION AND UTILIZATION FACILITIES i
- 1. The authority citation for Part 50 continues to read as follows:
AUTHORITY: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68 Stat. !
936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 1244, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201, 2232, 2233, 2236, 2239, ;
2282); secs. 201, as amended, 202, 206, 88 Stat. 1242, as amended, 1244, 1246, ;
FRN - 25
=
4 9 I
i (42 U.S.C. 5841, 5842, 5846). '
Section 50.7 also issued under Pub. L.95-501, .sec.10, 92 Stat. 2951 (42 U.S.C. 5851). Section 50.10 also issued under secs. 101, 185, 68 Stat. ,
936, 955 Stat. 853 as amended (42 U.S.C. 2131, 2235), sec.102, Pub. L.91-190, 83 I under sec.(42 U.S.C. Sections 4332). 50.13, 50.54(dd) and 50.103 also issued l
108, 68 Stat. 939, as amended (42 U.S.C. 2138). Sections 50.23, '
l 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, l
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).
l Section 50.78 also issued under sec. 122, 68 Stat. 939 (42 U.S.C. 2152). l 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 1 U.S.C. 2237).
< l For the purposes of sec. 223, 68 Stat. 958, as amended (42 U.S.C. 2273), l il 50.5, 50.46(a) and (b), and 50.54(c) are issued under sec. 161b, 68 Stat.
948, as amended (42 U.S.C. 2201(b)); il 50.5, 50.7(a), 50.10(a)-(c),
! 50.34(a) and (e), 50.44(a)-(c), 50.46(a) and (b), 50.47(b), 50.48(a), (c),
(d), and (e), 50.49(a), 50.54(a)(1), (1)(1), (1)-(n), (p), (q), (t), (v), and (y),
i 50.55(f), 50.55a(a), (c)-(e), (g), and (h), 50.59(c), 50.60(a ,
l Stat. 949, as amen)ded (42 U.S.C. 2201(1)); and fl50.49(
i 50.54(w),(z),(bb),
l 50.70(a), 50.71(a)-(c(cc), and (dd), 50.55(e), 50.59(b), 50.61(b), 50.62(b),
! ) and (e), 50.72(a), 50.73(a) and (b), 50.74, 50.78, and 50.90 are issued under sec. 1610, 68 Stat. 950, as amended (42 U.S.C. 1 2201(o)).
- 2. In 550.2, add the definitions for exclusion area, low population zone, and population center distance to read as follows:
6 50.2 Definitions.
As used in this part, 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 i result.
Low conulation zone means the area immediately surrounding the exclusion area which contain 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 of total
- population within this zone because the situation may vary from case to case.
1 l
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- 1. ._ .
o a Whether a specific number of people can, for example, be evacuated from a specific area, or instructed to take shelter, on a timely basis will depend on many factors such as location, number and size of highways, scope and extent of advance planning, and actual distribution of residents within the area.
Population center distance means the distance from the reactor to
, residents. boundary of a densely populated center containing more then 25,000 the nearest 3.
In 150.8, paragraph (b) is revised to read as follows:
650.8 Information collection requirements: OMB approval.
(b) part appear The in approved information collection requirements contained in this 50.30, 50.33, 50.33a, 50.34, 50.34, 50.34a, 50.35, 50.36, 50.36a, 50.48, 50.49, 50.b4, 50.55, 50.55a, 50.59, 50.60, 50.61, 50.63, 50.64, 50.65, I, J, K, 50.71, M, N, 0, 50.72, Q, R,50.80, and S. 50.82, 50.90, 50.91, and Appendices A, B, E, G, H, 4.
In 650.34, footnotes 6, 7, and 8 are redesignated as footnotes 8, l 9 and 10, paragraph (a)(1) is revised and paragraphs (a)(12) and (b)(10) are added
- to *read as follows:
(1) assessmentAofdescription the facility.andSite safety assessment of the site and a safety characteristics must comply with Part 100 cf this chapter. Special attention must be directed to plant design features j this assessment, an applicant shall assume a fissionfrom product relea the core into the containment assuming that the facility is operated at the ultimate power level contemplated. The applicant shall perform an evaluation and analysis of the postulated fission product release, using the expected ;
demonstrable containment leak rate and any fission product cleanup systems intended to mitigate the consequences of the accidents, together with ,
applicable site characteristics t offsite radiological consequence,s. including site meteorology, to evaluate the i The evaluation must determine that: '
(i) An individual located at any point on the boundary of the exclusion area for two hours immediately following the onset of the postulated fission product release would not receive a total radiation dose to the whole body in excess of 25 rem' or a total radiation dose in excess of 300 rem' to i
i
- The fission product release assumed for this evaluation should be based upon a major accident, I hypothesized or determined from considerations of possible accidental events, that would result in potential hazards not exceeded by those from any accident considered credible. Such accidents have generally been assumed j to result in substantial meltdown of the core with subsequent release into the containment of appreciable quantitles of fission products. !
' The whole body dose of 25 rem referred to above has been stated to correspond numerically to the once in a lifetime accidental or emergency dose for radiation workers which, according to NCRP reconsnendations may be disregarded in the deternination of their radiation exposure status (see NBS Handbook 69 dated June 5,1959). )
[
More recently, this whole body dose value has also been provided as guidance for radiation workers performing emergency services involving life saving activities or protectica of large populations where lomer doses are j not practicable (see EPA, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents, ;
Draft, September 1990). However, neither its use nor that of the 300 rem value for thyroid exposure as set j forth in this section are intended to imply that these numbers constitute acceptable limits for emergency doses i to the public under accident conditions. Rather, this 25 rem whole body value and the 300 rem thyroid value j have been set forth in this section as reference values, which can be used in the evaluation of plant design FRN - 27 i
)
l
4 o the thyroid from iodine exposure.
(ii) An individual located at any point on the outer radius of a 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 total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure. For purposes of this evaluation, a low population zone boundary of 3.0 miles is assumed.
(iii) 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.
A NOTE: Reference is made to Technical Information Document (TID) 14844, dated March 23, 1962, which contains a fission product release into containment which has been used in past evaluations. The fission product release given in TID--14844 may be used as a point of departure upon consideration of severe accident research insights available since its issuance, upon consideration of plant design features intended to mitigate the consequences of accidents, or upon characteristics of a particular reactor. Copies of Technical Information Document 14844 may be obtained from the Commission's Public Document Room, 2120 L Street, NW. (Lower Level), Washington, DC., or by writing the Director of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC. 20555.
(12) On or after [ EFFECTIVE DATE OF THE REGULATION), applicants who apply for a construction permit pursuant to this part, or a design certification or combined license pursuant to Part 52 of this chapter, as partial conformance to General Design Criterion 2 of Appendix A to this part, shall comply with the earthquake engineering criteria in Appendix S of this part.
(b)
(10) On or after (EFFECTIVE DATE OF THE REGULATION], applicants who apply for an operating license pursuant to this part, or a design certification or combined license pursuant to Part 52 of this chapter, as partial conformance to General Design Criterion 2 of Appendix A to this part, shall comply with the earthquake engineering criteria of Appendix S to this part. However, if the construction permit was issued prior to [ EFFECTIVE DATE OF THE REGULATION], the applicant shall comply with the earthquake engineering criteria in Section VI of Appendix A to Part 100 of this chapter.
- 5. In 550.54, paragraph (ee) is added to read as follows:
650.54 Conditions of licenses.
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.
I FRN - 28
(ee) For licensees of. nuclear power plants that have implemented the earthquake engineering criteria in Appendix S of this part, plant shutdown is required if the criteria in, Paragraph IV(a)(3) of Appendix S are exceeded.
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.
- 6. Appendix S to Part 50 is added to read as follows:
APPENDIX S TO PART 50 - EARTHQUAKE ENGINEERING CRITERIA FOR NUCLEAR POWER PLANTS General Information This appendix applies to applicants who apply for a design certification or combined license pursuant to Part 52 of this chapter or a construction permit or operating license pursuant to Part 50 of this chapter on or after
[ EFFECTIVE DATE OF THIS REGULATION]. However, if the construction permit was issued prior to (EFFECTIVE DATE OF THIS REGULATION], the operating license applicant shall comply with the earthquake engineering criteria in Section VI of Appendix A to 10 CFR Part 100.
I. Introduction Each applicant for a construction permit, operating license, design certification, or combined license is required by 150.34(a)(12),
p 150.34(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, a condition of all operating licenses for nuclear power plants, as specified in 150.54(ee), is plant shutdown if the criteria in Paragraph IV(a)(3) of this appendix are exceeded.
These criteria implement General Design Criterion 2 insofar as it requires structures, systems, and components important to safety to withstand the effects of earthquakes.
II. Scope The evaluations described in this appendix are within the scope of investigations permitted by 150.10(c)(1) of this chapter.
III. Definitions As used in these criteria:
The Safe Shutdown Earthauake Groynd 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 reouired to withstand the effects of the Safe Shutdown Earthouake Ground Motion or surface deformation are those necessary to assure: **g (1) The integrity of the reactor dBelant pressure boundary, (2) The capability to shut down the reactor and maintain it in a safe FRN - 29
. --. - ~- - - - -
e ,
shutdown condition, or '
(3) The capability to prevent or mitigate the consequences of accidents l that could result in potential offsite exposures comparable to the guideline '
exposures of $50.34(a)(1) of this chapter. l The Operatina Basis Earthauake Ground Motion (OBE) is the vibratory !
ground motion for which those features of the nuclear power plant necessary :
for continued public operation will remain without undue risk to the health and safety of the functional.
l 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 resoonse spectrum is a plot of the maximum responses (acceleration, velocity, or displacement) of a family 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.
Surface deformation is distortion of soils or rocks at or near the ground surface by the processes of folding, faulting, compression, or !
I extension is associatedas with a result of various earthquake earth forces. Tectonic surface deformation processes, !
i t
Combined license means a combined construction permit and operating l license with conditions for a nuclear power facility issued pursuant to l Subpart C of Part 52 of this chapter.
l Desian 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."
IV. Application To Engineering Design I
The following are pursuant to the seismic and geologic design basis requirements of Paragraphs V(a) through (f) of Appadix B to Part 100 of this chapter:
(a) Vibratory Ground Motion.
(1) Safe Shutdown Earthquake Ground Motion. The Safe Shutdown Earthquake Ground Motion must be characterized by free-field ground motion response spectra at the free ground surface or hypothetical rock outcrop, as appropriate. 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 developed from an ensemble of response spectra related to the vibratory motions ctused by more than one earthquake.
At a minimum, the horizontal Safe Shutdmn Earthquake Ground Motion at the foundation level of the structures must be an appropriate response spectrum with a peak ground acceleration of at least 0.1g.
The nuclear power plant must be designed so that, if the Safe Shutdown Earthquake Ground Motion occurs, certain structures, systems, and components will remain functional and within applicable stress 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 l safety-related structures, systems, and components. The design of the nuclear i
FRN - 30
, w 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 Paragraph V(f) of Appendix B to.
. Part 100 of this chapter.
The required safety functions of. structures, systems, and components must be assured during and after the vibratory ground motion associated with the Safe Shutdown Earthquake Ground Motion through design, testing, or qualification methods.
The evaluation must take into account soil-structure interaction effects and the expected duration of vibratory motion. It is permissible to design for strain limits in excess of yield strain in some of these safety-related structures, systems, and components during the Safe Shutdown Earthquake Ground Motion and under the postulated concurrent loads, provided the necessary safety functions are maintained.
(2) Operating Basis Earthquake Ground Motion.
(1) 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.
The requirements associated with this Operating Basis Earthquake Ground Motion in Paragraph (a)(2)(1)(b)(1) 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. Analysis and design must be performed to demonstrate that the requirements associated with this Operating Basis Earthquake Ground Motion in (1) are satisfied. The design must take into account soil-structure interaction effects and the expected duration of vibratory ground motion.
(1) 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 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. Prior to resuming operations, the licensee must demonstrate to the Connission that no functional damage has occurred to those features necessary for continued operation without undue risk to the health and safety of the public.
(4) Required Seismic Instrumentation. Suitable instrumentation must be provided so that the seismic response of nuclear power plant features important to safety can be evaluated promptly after an earthquake.
-(b) Surface Deformation. The potential for surface deformation must be taken into account in the design of the nuclear power plant by providing reasonable assurance that in the event of deformation, certain structures, systems, and components will remain functional. In addition to surface deformation induced loads, the design of safety features must take into account seismic loads, including aftershocks, and applicable concurrent functional: and accident-induced loads. The design provisions for surface deformation must be based on its postulated occurrence in any direction and Guidance is being developed in Draft Regulatory Guide 06-1017. " Pre-Earthquake Planning and heediate Nuclear Power Plant Operator Post-Earthquake Actions."
FRN - 31 4
4 .
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.
distantly generated seismic activity and other design conditions determi pursuant to Paragraphs V(e) and (f) of Appendix B to Part 100 of this chapter must be taken into account in the design of the nuclear power plant so as to prevent undue risk to the health and safety of the public.
PART 52 -- EARLY SITE PERMITS; STANDARD DESIGN CERTIFICATIONS; AND COMBINED LICENSES FOR NUCLEAR POWER PLANTS 7.
The authority citation for Part 52 continues to read as follows:
AUTHORITY:
953, 954, 955, 956,Secs. 103, 104, 161, 182, 183, 186, 189, 68 Stat. 936, 948, as amended, sec. 234, 83 Stat. 1244, as amended (42 U.S.C.
2133,1246, 1244, 2201, 2232, 2233, 2236, 2239, 2282); secs. 201, 202, 206, 88 Stat. 1242, >
as amended (42 U.S.C. 5841, 5842, 5846).
- 8. In 552.17, the introductory text of para paragraph (a)(1)(vi) are revised to read as follows: graph (a)(1) and 552.17 Contents of applications.
(a)(1) The application must contain the information required by 50.33(a)--(d), the information required by 150.34(a)(12) and (b)(10), and, to this section, the information required by 950.33(g) and (j), andthe e 650.34(b)(6)(v). The application must also contain a description and safety assessment of the site on which the facility is to be located, with appropriate attention to features affecting facility design. 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 theofradiological 650.34(a)(1) this chapter. consequence evaluation factors identified in of this chapter. Site characteristics must comply with Part 100 In addition,
- the
- application should describe the following:
(vi) characteristics of the proposed site;The seismic, meteorological, hydrologic, and 9.
follows: In 10 CFR Part 52, Appendix Q, paragraph 8 is added to read as Suitability Issue. Appendix Q to Part 52 - Pre-Application Early Review of the Site 8.
early site permit is subject to a full early site permit review.Notwithstandin t
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(
e e -
PART 100 - REACTOR SITE CRITERIA
- 10. The authority citation for Part 100 continues to read as follows:
AUTHORITY: Secs. 103, 104, 161, 182, 68 Stat. 936, 937, 948, 953, as amended (42 U.S.C. 2133, 2134, 2201, 2232); sec. 201, as amended, 202, 88 Stat. 1242, as amended, 1244 (42 U.S.C. 5841, 5842).
- 11. The table of contents for Part 100 is revised to read as follows:
- PART 100 - REACTOR SITE CRITERIA Sec.
4 100.1 Purpose.
100.2 Scope.
100.3 Definitions.
100.8 Information collection requirements: OMB approval.
- Subpart A -
Evaluation Factors for Stationary Power Reactor Site Applications before [ EFFECTIVE DATE OF THIS REGULATION] and for Test 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 (EFFECTIVE DATE OF THIS REGULATION).
100.20 Factors to be considered when evaluating sites.
100.21 Determination of exclusion area and population distribution.
100.22 Evaluation of potential man-related hazards.
APPENDIX A - Seismic and Geologic Siting Criteria for Nuclear Power Pla.its.
APPENDIX B - Criteria for the Seismic and Geologic Siting of Nuclear Power Plants
- 12. Section 100.2 is revised to read as follows:
6100.1 Purpose.
(a) This part sets forth standards for evaluation of the suitability of proposed sites for stationary power and testing reactors subject to Part 50 or ,
Part 52 of this chapter. l (b) This part identifies the factors considered by the Commission in the evaluation of reactor sites and the standards used in approving or disapproving proposed sites.
1
- 13. Section 100.2 is revised to read as follows:
1100.2 Scope. '
l (a) This part applies to applications filed under Part 50 or Part 52 of this chapter for early site permit, construction permit, operating license, or
{
FRN - 33 I i
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combined license (construction permit and operating license) for power' and testing reactors.
l (b) The site criteria contained in this part apply primarily to reactors I l for which there is significant operating experience. These site criteria can also be applied to other reactor types, such as for reactors that are novel in l design and unproven as prototypes or pilot plants. For plants without significant operating experience, it is expected that these basic criteria will )
. be applied in a manner that takes into account the lack of experience. In the l
application of these criteria which are deliberately flexible, the safeguards ,
provided , either site isolation or engineered features, should reflect the lack i of certainty that only experience can provide, i
- 14. Section 100.3 is revised to read as follows:
6100.3 Definitions.
As used in this part:
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 l in an exclusion area under appropriate limitations, provided that no significant !
hazards to the public health and safety will result. '
Low conulation zone means the area immediately surrounding the exclusion area which contains residents, the total number and density of which are such that there is a reasonable probability that appropriate protective measures could be taken in their behalf in the event of a serious accident. These guides do not specify a permissible population density or total population within this zone because the situation may vary from case to case. Whether a specific number of people can, for example, be evacuated from a specific area, or instructed to take shelter, on a timely basis will depend on many factors such as location, number and size of highways, scope and extent of advance planning, and actual ;
distribution of residents within the area.
Population center distance means the distance from the reactor to the nearest boundary of a densely populated center containing more than about 25,000 residents.
Power reactor means a nuclear reactor of a type described in 9950.21(b) or 50.22 of this chapter designed to produce electrical or heat energy.
Testina reactor means a testina facility as defined in 650.2 of this chapter.
, 15. Section 100.8 is revised to read as follows:
l 9100.8 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 l Budget (OMB) for approval as required by the Paperwork Reduction Act of 1980 (44 FRN - 34
e e U.S.C. 3501 et seq.)'.
contained in this part under OMB control has approved number the information collection requirements 3150 - 0093.
(b) The approved information collection requirements contained in this part appear in Appendix A and Appendix B.
16.
follows: A heading for Subpart A is added directly before $100.10 to read as Subpart A -
i Evaluation Factors for Stationary Power Reactor Site j Applications Reactors. before (EFFECTIVE DATE OF THIS REGULATION] and for Test 17.
Section 100.10 is added to read as follows:
1100.10 Factors to be' considered when evaluating sites.
Factars considered in the evaluation of sites include those relating both to the proposed reactor design and the characteristics peculiar to the site. It is expected that reactors will reflect through their design, construction and operation an extremely low probability for accidents that could result in release of significant quantities of radioactive fission products. In addition, the site location and the engineered features included as safeguards against the hazardous consequences exposure. of an accident, should one occur, should insure a low risk of public In particular, the Commission will take the following factors into consideration in determining the acceptability of a site for a power or testing reactor:
(a) Characteristics of reactor design and proposed operation including--
(1) Intended use of the reactor including the proposed maximum power level and the natura and inventory of contained radioactive materials; (2) The extent to which generally accepted engineering standards are ,
applied to the design of the reactor;.
I (3) The extent to which the reactor incorporates unique or unusual features having a significant bearing on the probability or consequences of accidental release of radioactive materials; (4) The safety features.that are to be engineered into the facility and those barriers that must be breached as a result of an accident before a release of radioactive material to the environment can occur.
(b) Population density and use characteristics of the site environs, including the exclusion area, low population zone, and the population center distance.
(c) Physical characteristics of the site, including seismology, meteorology, geology, and hydrology.
(1) Appendix A to Part 100, " Seismic and Geologic Siting Criteria for Nuclear Power Plants," describes the nature of investigations required to obtain j the geologic and seismic data necessary to determine site suitability and to provide reasonable assurance that a nuclear power plant can be constructed and {
i operated at a proposed site without undue risk to the health and safety of the public. It describes procedures for determining the quantitative vibratory ground motion design basis at a site due to earthquakes and describes information ;
needed to determine whether and to what extent a nuclear power plant need be l
designed to withstand the effects of surface faulting.
(2) Meteorological conditions at the site and in the surrounding area I
FRN - 35 l 1 l
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.should be considered. .
(3) Geological and hydrological characteristics of the proposed site may have a bearing on the consequences of an escape of radioactive material from the facility. . Special precautions, should be planned if a reactor is to be located at a site where a significant quantity of radioactive effluent might accidentally flow into nearby streams or rivers or might find ready access to underground water tables.
(d) Where unfavorable physical characteristics of the site exist, the proposed site may nevertheless be' found to be acceptabl.e if the design of the facility includes appropriate and adequate compensating engineering safeguards.
- 18. Section 100.11 is added to read as follows:
6100.11 Determination of. exclusion area, low population zone, and population center distance. ,
(a) As an aid in evaluating a proposed site, an applicant should assume a fission product release' from the core, the expected demonstrable leak rate from i the containment and the meteorological conditions pertinent to his site to derive an exclusion area, a low population zone and population center distance. For the purpose of this analysis, which shall set fceth the basis for the numerical values used, the applicant should determine the following:
(1) An exclusion area of such size t.iat an individual located at any point on its boundary for two hours immediately fuMwing onset of the postulated fission product release would not receive a total radiation dose to the whole body in excess of 25 rem
(2) A low population zone of such size that an individual located at any point on its outer boundary who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a total radiation dose to the whole body in excess of 25 rem er a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
(3) A population center distance of at least one and one-third times the distance from the reactor to the outer boundary of the low population zone.
In applying this guide, the boundary of the population center shall be determined
'The fission product release assumed for these calculations should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events, that would result in potential hazards not exceeded by those from any accident considered credible, Such accidents have generally been assumed to result in substantial meltdown of the core with subsequent release of appreciable quantities of fission products.
- The whole body dose of 25 rem referred to above corresponds numerically to the once in a lifetime accidental or emergency dose for radiation workers
, which, according to NCRP recommendations may be disregarded in the determination of their radiation exposure status (see NBS Handbook 69 dated June 5, 1959).
However, neither its use nor that of the 300 rem value for thyroid exposure as set forth in these site criteria guides are intended to imply that these numbers constitute acceptable limits for emergency doses to the public under accident conditions. Rather, this 25 rem whole body value and the 300 rem thyroid value have been set forth in these guides as reference values, which can be used in the evaluation of reactor sites with respect to potential reactor accidents of exceedingly low probability of occurrence, and low risk of public exposure to radiation.
FRN - 36
~
, j i upon consideration of population distribution. Political boundaries are not :
controlling in the application of this guide. Where very large cities are '
. involved, a greater distance may be necessary - because of total integrated population dose consideration.
(b) For sites for multiple reactor facilities consideration should be given ,
to the following: '
(1) If the reactors are independent to the extent that an accident in one reactor would not initiate an accident in another, the size of the exclusion area, low population zone and population center distance shall be fulfilled with respect to.each reactor individually. The calculated envelopes ;
of each of the plants areas shall be overlayed of the areas such that the outermost composite boundary shall then be taken as the plant boundary.
(2) If the reactors are interconnected to the extent that an accident in one reactor could affect the safety of operation of any other, the size of the exclusion area, low population zone and population center distance shall be based .
upon the assumption that all int'erconnected reactors emit their postulated :
fission product releases -simultaneously. This requirement may be reduced in '
relation to the degree of coupling between reactors, the probability . of concomitant accidents and the probability that an individual would not be exposed i to the radiation effects from simultaneous releases. The applicant would be expected to justify to the satisfaction of the Commission the basis for such a reduction in the source. term. -
(3) The applicant is expected to show that the simultaneous operation of multiple-reactors at a site will not result in total radioactive effluent ;
releases beyond the allowable limits of applicable regulations.
For further guidance in developing the exclusion area, the low NOTE:
population zone, and the population center distance, reference is made to .
Technical: Information Document 14844, dated March 23, 1962, which contains a procedural method and a sample calculation that result in distances roughly reflecting current siting practices of the Commission. The calculations -
described in. Technical Information Document 14844 may be used as a point of departure for consideration of particular site requirements which may result from evaluation of the characteristics of a particular reactor, its purpose and method of operation. Copies of Technical Information Document 14844 may be obtained
- from the Commission's Public Document Room, 2120 L Street, NW. (Lower Level), ,
Washington, DC, or by writing the Director of Nuclear Reactor Regulation, U.S.
Nuclear Regulatory Commission, Washington, DC. 20555. -
i
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I FRN - 37 4 i
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- 19. Subpart B (5 6100.20 .100.22) is added to read as follows: 1 Subpart -B -
Evaluation Factors for Stationary Power Reactor Site l
Applications on or after [ EFFECTIVE DATE OF THE FINAL REGULATION]. ;
1100.20 Factors to be considered when evaluating sites.
The Commission will take the following factors into consideration in i
determining the acceptability of a site for a stationary power reactor:-
L (a) Population density and use characteristics of the site environs, i including the exclusion area, the population distribution, and the compatibility of the site with the development of an emergency plan.
- (b) The nature and proximity of man-related hazards (e.g. airports, dams, transportation routes, military and chemical facilities).
(c) Physical characteristics of the site, including seismology, meteorology, geology, and hydrology.
(1) Appendix B, " Criteria for the Seismic and Geologic Siting of Nuclear Power Plants After [ Effective Date)," describes the criteria and nature ,
{
of investigations required to obtain the geologic and seismic data necessary to
' determine site suitability. !
l (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) should te identified and characterized.
(3) Factors important to hydrological radionuclide transport-(such as soil, . sediment, and rock characteristics, adsorption and retention i l
l coefficients,. ground water velocity, and distances to the nearest surface body l I of water) flood along should with betheobtained from on-site measurements. The maximum probable i potential for seismic induced floods discussed in Appendix l 8 should be estimatec using historical data. l t
1100.21 Determination of exclusion area and population distribution. 4 (a) Each reactor facility shall have an exclusion area, as defined in 5100.3(a) of this part.
(1) For sites with a single reactor facility, the distance to the 1.
exclusion area boundary at any point (as measured from the reactor center point) shall be at least 0.4 miles (640 meters).
- l. (2) For' sites with multiple reactor facilities, consideration should be given to the following: If the reactors are independent to the extent that an accident in one reactor would not initiate an accident in another, the size of ' each exclusion area shall be determined with respect to each reactor i individually. The exclusion area for the site shall then be taken as the plan overlay of the sum of the exclusion areas for each reactor. If the reactors are
! interconnected to the extent that an accident in one reactor would initiate an accident in another, the size of the exclusion area for each reactor shall be determined on a. case by case basis.
(b)(1) If the offsite population density at the proposed site exceeds the 1 values given in paragraph (b)(2) of this section, the site may not be approved
( 6y the Commission unless the applicant demonstrates either:
l (1) That there are no reasonably available alternative sites with i significantly lower population densities, or (ii) That the proposed site is preferred over an alternative site with -
1 significantly lower population density on the basis of other l considerations.
i FRN - 38 l
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(2) The population density, including weighted transient populat' ion, projected at the time of initial site approval or early site permit renewal should not exceed 500 people per square mile averaged over any radial distance out to 30 miles (cumulative population at a distance divided by the total circular area at that distance). The projected population density, including weighted transient population, 40 years after the time of initial site approval or early site permit renewal should not exceed 1000 people per square mi_le averaged over any radial distance out to 30 miles.
(3) Transient population must be included for those sites where a significant number of people (other than those just passing through the area) work, reside- part-time, or engage in recreational activities and 'are not s permanent residents of the area. The transient population should be considered ,
for siting purposes by weighting the transient population according to the fraction of the time the transients are in the area.
(c) Physical characteristics of the proposed site, such as egress limitations from the area surrounding the site, that could pose a significant impediment to the development of emergency plans, shall be identified.
6100.22 fyaluation of Man-related Hazards.
k (a) Potential hazards to the plant from man-related activities associated with nearby transportation routes, military, and industrial facilities shall be identified and their potential effects evaluated. Potential hazards to the plant include such effects as explosions, fires, toxic and/or flammable chemical releases, dams (both upstream and downstream), pipeline accidents, and aircraft crashes and impacts.
(b) The effects of offsite hazards shall have a very low probability of affecting the safety of the plant. The likelihood and consequences of offsite hazards shall be estimated using data and assumptions that are as' realistic and representative of the site as is practical. The design bases for which the plant shall be designed shall be specified.
- 20. Appendix B to Part 100 is added to read as follows:
APPENDIX B TO PART 100 -- CRITERIA FOR THE SEISMIC AND GE0 LOGIC SITING OF NUCLEAR POWER PLANTS AFTER [ EFFECTIVE DATE]
General Information This appendix applies to applicants who apply for an early site permit or combined license pursuant to Part 52 of this chapter, or a construction permit or operating license pursuant to Part 50 of this chapter on or after [ EFFECTIVE DATE OF THIS REGULATION]. However, if the construction permit was issued prior to [ EFFECTIVE DATE OF THIS REGULATION), the operating license applicant shall comply with the seismic and geologic siting criteria in Appendix A to Part 100 of this chapter.
- 1. Purpose General Design Criterion 2 of Appendix A to Part 50 of this chapter requires that nuclear power plant structures, systems, and components important to safety be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of FRN - 39
I capability to perform their safety functions. It is the purpose of these criteria to set forth the principal seismic and geologic considerations that guide the Commission in its evaluation of the suitability of proposed sites for nuclear power plants and the suitability of the plant design bases established in consideration sites.' of the seismic and geologic characteristics of the proposed These criteria seismological are based on the current geophysical, geological, and information effects. They will be revised concerning faults and earthquake occurrences and as necessary when more complete information becomes available.
II. Scope These criteria, which apply to nuclear power plants, describe the nature of the investigations required to obtain the geologic and seismic data necessary to determine site suitability and provide reasonable assurance that a nuclear power plant can be constructed and operated at a proposed site without undue risk to the health and safety of the public. Geologic and seismic factors required to be taken into account in the siting and design of nuclear power plants are identified.
The investigations described in this appendix are within the scope of investigations permitted by 5 50.10(c)(1) of this chapter.
Each applicant for a construction permit, operating license, early site permit, or combined license shall investigate all seismic and geologic factors that may affect the design and operation of the proposed nuclear power plant irrespective of whether such factors are explicitly included in these criteria.
Both deterministic and probabilistic evaluations must be conducted to determine site suitability and seismic design requirements for the site. Additional investigations or more conservative determinations than those included in these criteria may be required for sites located in areas with complex geology, recent tectonic deformation, or in areas of high seismicity. If an applicant believes that the particular seismic and geologic characteristics of a site indicate that some of these criteria, or portions thereof, need not be satisfied, the applicant shall identify the specific sections of these criteria in the license application and present supporting data to clearly justify such departures. The Director, Office of Nuclear Reactor Regulation approves any deviations.
III. Definitions As used in these criteria:
The maanitude of an earthquake is a measure of the size of an earthquake and is related to the energy released in the form of seismic waves. Magnitudo means Magnitude, Homent the numerical value on a standardized scale such as, but not limited to, Magnitude scales. Surface Wave Magnitude, Body Wave Magnitude, or Richter A deterministic source earthouake (DSE) is the largest earthquake that can reasonably be expected to occur in a given seismic source in the current tectonic regime, and is to be used in a deterministic analysis. It is generally based on the maximum historical earthquake associated with that seismic source, unless recent geological evidence warrants a larger earthquake, or where the rate of Considerations presented in this regulation are general.
discussion are providad in regulatory guides and standard review plan sections. Acceptable methods and a FRN - 40
1 occurrence of earthquakes indicates the likelihood of larger than the' largest historical event.
The Safe Shutdown Earthauake Ground Motion (SSE) is the vibratory ground motion for which certain structures, systems, and components must be designed to '
remain functional. l A fault is a tectonic structure along which differential slippage of the ;
adjacent earth materials has occurred parallel to the fracture plane. A fault ;
may. have gouge or breccia between its two ' walls and includes any associated monoclinal flexure or_ other similar geologic structural feature. i Surface faultina is differential ground displacement at or.near the surface caused directly.by fault movement and is distinct from nontectonic types of ,
ground disruptions, such as landslides, fissures, and craters.
.S,urface deformation is distortion of soils or rocks at or near the ground .
surface by the processes of folding, faulting, compression, or extension as a L result of various earth forces. Tectonic surface deformation is associated with .
earthquake processes.
A seismic source is a general term referring to both seismogenic sources and capable tectonic sources.
A seismoaenic source is a portion of the earth that has uniform earthquake t
. potential (same deterministic source earthquake and frequency of recurrence)
~ distinct from the surrounding area. A seismogenic source will not cause surface displacements. Seismogenic sources cover a wide range of possibilities from a well-defined tectonic structure to simply a large region of diffuse seismicity (seismotectonic province) thought to be characterized by the same earthquake recurrence model. A seismogenic source is also characterized by its involvement '
in the current tectonic regime as reflected in the Quaternary (approximately the last 2 million years) geologic history.
A canable tectonic source is a tectonic structure that can generate both earthquakes and tectonic surface deformation such as faulting or folding at or near the surface in the present seismotectonic regime. It is characterized by at least one of the following characteristics:
(1) The presence of surface or near-surface deformation of landforms or 9eologic deposits of recurring nature within the last approximately 500,000 years or at least once in the last approximately 50,000 years. '
(2) A reasonable association with one or 'more large earthquakes or sustained earthquake activity that is usually accompanied by significant surface deformation.
(3) A structural association with a capable tectonic source having !
characteristics in Paragraph (iii) (1) of this Section so that movement on one could be reasonably expected to be accompanied by movement on the other.
In some cases, the geologic evidence of past activity at or near the ground surface along a particular capable tectonic source may be obscured at a particular site. This might occur, for example, at a site having a deep :
overburden. For these cases, evidence may exist elsewhere along the structure from which an evaluation of its characteristics in the vicinity of the site can be reasonably based. This evidence must be used in determining whether the -
structure is a capable tectonic source within this definition.
Notwithstanuing the foregoing paragraphs in III (1), (2) and (3), of this !
section, structural association of a structure with geologic structural features that rea geologically old (at least pre-Quaternary) such as many of those found in the lastern region of the United States must, in the absence of conflicting evidme, demonstrate that the structure is not a capable tectonic ' so0rce within this definition.
A response soectrum is a plot of the maximum responses (acceleration, velocity, or displacement) of a family of idealized single-degree-of-freedom -
l FRN - 41 ;
1
)
s s oscillators as a function of the natural frequencies of the oscillators for a t
given ' damping value. The response spectrum is calculated for a specified
-i vibratory motion input at the oscillators' supports. l Combined license means a combined construction permit and operating license j i
with 52 ofconditions this chapter. for a nuclear power facility issued pursuant to Subpart C of Part Desian 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.
l design." A design so approved may be referred to as a " certified standard 1 f.
IV. Required Investigations' :
r The geological, seismological, and engineering characteristics of a site
-and its environs must be investigated in sufficient scope and detail to permit i
an adequate evaluation of the proposed site, to provide sufficient information to support both probabilistic and deterministic evaluations required by these criteria, 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 by the nature of the region surrounding the proposed site. The !
investigations must be carried out by a review of the pertinent literature and field investigations as identified in paragraphs (a) through (e) of this section.
(a) Vibratory Ground Motion. *
- The purpose of these investigations is to obtain information needed to -
assess the Safe Shutdown Earthquake ground motion. The seismic sources (capable !
tectonic sources and seismogenic sources) in the site region must be identified and evaluated. The deterministic source earthquakes must be evaluated for each I seismic source.
(b) Tectonic Surface Deformation.
The purpose of these investigations is to assess the potential for tectonic surface deformation near thet .ite and, if any to what extent the nuclear power '
plant needs to be designed for these occurren,ces. '
(c) Nontactonic Deformation.
l
' The purpose of these investigationsis to assess the potential for surface deformations not directly attributable to tectonics, such as those associated
.with, subsidence or collapse as in karst terrain, glacially induced offsets, and growth faulting. Paragraph IV(b) concerns investigations required for tectonic surface deformation that can occur coseismically. Nontectonic phenomena can represent significant surface displacement hazards to a site, but can in many cases be monitored, controlled, or mitigated by engineering, or it can be
~ demonstrated that conditions that were the cause of the displacements no longer exist. Geological and geophysical investigations must be carried out to identify l and define nontectonic deformation features and, where possible, distinguish them i
from tectonic surface displacements. -If such distinction is not possible, the questionable features must be treated as tectonic deformation.
(d) Seismically Induced Floods and Water Waves.
l The purpose of these investigations is to assess the potential for nearby and distant tsunamis and other waves that could affect coastal sites. Included in this assessment is the determination of the potential for slides of earth
! material that could generate waves. Information regarding distant and locally l
l FRN - 42 l
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- generated waves or tsunamis that have affected the site, and available evidence of runup and drawdown associated with these events, shall be analyzed. Local features of. coastal or undersea topography which could modify wave runup or -
drawdown must be considered. For sites located near lakes or rivers, analyses
.must include the potential for seismically induced floods or water waves, as, for
-example, from the. failure during an earthquake of a-dam upstream or from slides of earth or debris into a nearby lake.
(e) Volcanic Activity.
The purpose of these investigations is to assess the potential volcanic hazards that would adversely affect the site.
V. Seismic and Geologic Design Bases (a) Determination of Deterministic Source Earthquakes.
For each seismogenic and capable tectonic source identified in Paragraph IV(a), the deterministic source earthquake must be evaluated. At a minimum, the deterministic source earthquake must be the largest historical earthquake in each source. The uncertainty in determining the deterministic source earthquakes must be accounted for in the probabilistic analysis.
(b) Determination of the Ground Motion at the Site.
The ground motion at the site must be estimated from all earthquakes, including the deterministic source earthquake associated with each source, which could potentially affect the site using both probabilistic and deterministic approaches. In the deterministic approach, the deterministic source earthquake associated with each source must be assumed to occur at the part of the source which is closest to the site. Appropriate models, including _ local site conditions, must be used to account for uncertainty in estimating the ground motion for the site. The ground motion is defined by both horizontal and vertical free-field ground motion response spectra at the free ground surface or hypothetical rock outcrop, as appropriate.
(c) Determination of Safe Shutdown Earthquake Ground Motion.
The Safe Shutdown Earthquake Ground Motion is characterized by response spectra. These spectra are developed from or compared to the ground motions determined in Paragraph V(b). Deterministic and probabilistic seismic hazard evaluations must be used to assess the adequacy of the Safe Shutdown Earthquake Ground Motion. The probability of exceeding the Safe Shutdown Earthquake Ground Motion is considered acceptably low if it is less than the median probability-computed from the current [ EFFECTIVE DATE OF THE REGULATION) population of nuclear power plants.
At a minimum, the horizontal Safe Shutdown Earthquake Ground Motion at the foundation level of the structures must be an appropriate response spectrum with a peak ground acceleration of at least 0.19 (d) Determination of Need To Design for Surface Tectonic and Nontactonic Deformations.
Sufficient geological, seismological, and geophysical data must be provided to clearly establish that surface deformation need not be taken into account in the design of a nuclear power plant. When surface deformation is likely, an assessment of the extent and nature of surface deformations must be characterized.
(e) 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, taking into consideration the results of the investigation required by paragraph (d) of section IV in this Appendix.
FRN - 43
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4 e (f) Determination of Other Design Conditions. '
-(1) Soil Stability. Vibratory ground motions determined in Paragraph V(b) can cause soil instability from ground disruption such as fissuring, lateral spreads, differential settlement, and liquefaction, which is not directly related to surface faulting. Geological features that could affect the foundations of the proposed nuclear power plant structures must be evaluated, taking into account ,
the information concerning the physical properties of materials underlying the site and the effects of the vibratory ground motion determined in Paragraph V(b). .
(2) Slope stability. Stability of all slopes, both natural and artificial, !
must be considered, the failure of which could adversely affect the nuclear power plant. An assessment must be made of the potential effects of erosion or '
deposition and of combinations of erosion or deposition with seismic activity, taking into account information concerning the physical properties of the i materials underlying the site and the effects of the vibratory ground motion determined in Paragraph V(b).
(3) Cooling water supply. Assurance of an adequate cooling water supply for emergency and long-term shutdown decay heat removal shall be considered in the design of the nuclear power plant, taking into account information concerning -
the physical properties of the materials underlying the site, the effects of the i Safe Shutdown Earthquake Ground Motion, and the design basis for tectonic and nontactonic surface deformation. Consideration of river blockage or diversion or other failures that may block the flow of cooling water, coastal uplift or subsidence, tsunami runup and drawdown, and the failure of dams and intake structures must be included in the evaluation where appropriate. '
(4) Distant structures. Those structures that are not located in the immediate vicinity of the site but are safety-related must be designed to !
withstand the effect of the Safe Shutdown Earthquake Ground Motion. The design basis for surface faulting must be determined on a basis comparable to that of the nuclear power plant, taking into account the material underlying the structures and the different location with respect to that of the site.
3 VI. Application To Engineering Design Pursuant to the seismic -and geologic design basis requirements of l paragraphs V(a) through (f), applications to engineering design are contained in 1 Appendix S to Part 50 of this chapter for the following areas: '
(a) Vibratory ground motion.
(1) Safe Shutdown Earthquake Ground Motion (SSE).
(2) Operating Basis Earthquake Ground Motion (OBE).
(3) Required Plant Shutdown. ;
(4) Required Seismic Instrumentation.
(b) Surface Deformation.
(c) Seismically Induced Floods and Water Waves and Other Design l Conditions.
4 1
)
i Dated at Rockville, Maryland, this ._ day of , 1992.
l For the Nuclear Regulatory Commission.
l Samuel J. Chilk, Secretary of the Commission.
FRN - 44 .
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1 DRAFT REGULATORY ANALYSIS 2 PROPOSED REVISION OF 10 CFR PART 100 3 AND 10 CFR PART 50 4
5 STATEMENT OF THE PROBLEM 6
7 This Regulatory Analysis covers two considerations. First is i.he revision of 10 CFR Part 100, " Reactor Site Criteria," , for future plants. The second 8
9 consideration is the revision of Appendix A " Seismic and Geologic Siting Criteria 10 for Nuclear Power Plants," to 10 CFR Part 100. Both considerations address the 11 relocation of plant design criteria from Part 100 to 10 CFR Part 50. This 12 regulatory analysis is presented in two parts, corresponding to these two 13 considerations.
14 15 Reactor Sitina Criteria (Nonseismic) 16 17 The NRC's regulations in 10 CFR Part 100, " Reactor Site Criteria," sets forth a 18 framework that guides the Commission in its evaluation of the suitability of 19 proposed sites for stationary power and testing reactors. The present criteria 20 regarding reactor siting were issued in April 1962. There were only a few small 21 power reactors operating at that time. The present regulation requires that 22 every reactor have an exclusion area that has no residents, although transient 23 use is permitted. A low population zone immediately beyond the exclusion area 24 is also required. The regulation recognizes the importance of accident 25 considerations in reactor siting; hence, a key element in it is the determination 26 of the size of the exclusion area via the postulation of a large accidental 27 fission product release within containment and the evaluation of the radiological 28 consequences in terms of doses. Doses are calculated for two hypothetical 29 individuals, located at any point (generally, the closest point) on the exclusion 30 area boundary and at the outer radius of the low population zone, and are 31 required to be within specified limits (25 rem to the whole body and 300 rem to 32 the thyroid gland). In addition, the nearest population center, containing about 33 25,000 or more residents, must be no closer than one and one-third times the 34 outer radius of the low population zone. The effect of these requirements is to 35 set both individual and, to some extent, societal limits on dose (and implicitly 36 on risk) without setting numerical criteria on the size of the exclusion area and 37 low population zone. In practice these site criteria contained in 10 CFR 100 do 38 more to influence reactor design than siting.
39 40 Since the issuance of Part 100 in 1962, there have been significant changes and 41 developments in reactor technology. The nuclear power industry has developed and 42 matured significantly. From the existence of a few small power plants generating 43 a very small fraction of the nation's electrical energy, the industry has grown 44 until there are presently about 110 power reactors in operation in the United 45 States. These supply about 20 percent of the nation's electricity. Reactor 46 power levels have also significantly increased. Early plants typically had 47 reactor power levels of about 150 megawatts thermal, whereas today's plan's have 48 power levels about 20 to 25 times greater.
49 50 There has been increased development of and reliance upon fission product cleanup 51 systems in modern plants to mitigate the consequences of postulated accidents.
52 As a result, it is possible for present nuclear power plants to be located at 53 sites with a very small exclusion area and still meet the dose criteria of 54 Part 100.
55 RA - 1
a .
1 2 There has potential nuclear also been accidents. an increased awareness and concern regarding the effect of l 3
4 importance in reactor siting from the very beginning, major developme 5 the issuance of the Reactor Safety Study (WASH-1400, Ref. 1) in 1975, the 6
7 the Chernobyl reactor in the Soviet Union in 1986, and the issu 1150, " Severe Accident Risks:
8 9 (Ref. 2), have greatly increased awareness, knowledge and concerns i .
10 Finally, 11 since initial promulgation of Part 100 in 1962, the Commission has 12 to review a number of others. approved more than 75 sites for nuclear power plants a 13 As a result of these reviews, much experience has 14 been gained regarding the site factors that influence risk and their range of acceptability.
15 16 17 The major impetus for the proposed rule is increased interest in new nuclear 18 power for newgeneration nuclear power andplants. the possibility that applicants will request site approval 19 The Commission believes that, in the event such 20 21 address those site factors important to risk and should reflec experience since the regulation was first issued in 1962.
22 23 24 Seismic Sitino and Earthouake Enoineerino Criteria 25 26 Appendix A, " Seismic and Geologic Siting Criteria for Nuclear Power Plants " to ,
27 28 29 staff in its evaluation of the adequacy o,f applicants' inv and earthquake phenomena and proposed plant design parameters.
30 Appendix A was tn important step in establishing a definitive The issuance of regulatory 31 framework plants. for dealing with earth science issues Appendix A contains the following statement:
in the licensing of nuclear pow 32 33 34 "These criteria are based on the limited geophysical and geological 35 information occurrence and available effect. to date concerning faults and earthquake 36 37 complete information becomes available."They will be revised as necessary when m 38 39 The bases December 13, for 1973. AppendixSince thenA were established in the late 1960s and became effective 40 41 geology, along with the occurre,nce of some licensing issues not fores 42 development the application of of Appendix this regulation. A, a number of significant difficulties have arisen in 43 following: Specific problematic areas include the 44 45 1.
46 47 latitude in judgment.In mak1ng geoscience assessments, there is a need 48 of limitations in dataThis latitude in judgment is needed because and geologic and seismic analyses, and 49 because of the rapid evolution taking place in the geosciences in 50 terms of accumulating knowledge and in modifying concepts. This 51 need was recognized when Appendix A was developed.
52 detailed geoscience assessments in Appendix A, a regulation, hasHowever, h 53 created difficulty for applicants and the staff in terms of inhibiting the use of needed judgment.
54 Also, 55 flexibility in applying basic principles to new situations and theit has inhibited use of evolving methods of analyses (for instance, probabilistic) in RA - 2
O A -
1 1
- l. the licensing process.
! 2 3 2. Various sections of Appendix A lack clarity and' are subject to
- 4 different interpretations and dispute. Also, some sections in the 5
Appendix do not provide sufficient information for implementation.
6 As a result of being both overly detailed in some areas and not
- j. 7 detailed enough in others, the Appendix has been the source of 8
- 9 licensing delays and debate and has inhibited the use of some types of analyses such as probabilistic seismic hazard analysis.
i 10 i 11 3. In other siting areas, such as hydrology, regulatory guidance has 12 been handled -effectively through use of regulatory guides. Many 13 problems encountered in implementing Appendix A could best be 14 !
15 alleviated through the use of regulatory guides and a program for !
continuous updating. l 16 17 4. The Operating Basis Earthquake (0BE) is associated with (1) the 18 functionality of those features necessary for continued operation 19 20 without undue risk to the health and safety of the public, (2) an i 21 earthquake that could reasonably be expected to affect the plant !
site during the operating life of the plant, (3) a minimum fraction )
22 of the Safe Shutdown Earthquake (SSE), and (4) plant shutdown if 23 vibratory ground motion is exceeded. These multi-aspects have (
-24 !
25 resulted in seismic criteria that have led to overly stiff piping systems and excessive use of snubbers and supports which, in fact, 26 could result in less reliable piping systems. Also, regulatory 27 guidance defining an exceedance of the OBE, and plant shutdown or 28 !
restart procedures have not been developed. Post earthquake ,
29 evaluations are handled on an ad-hoc basis.
30 31 5. The stipulation in Appendix A that the SSE response spectra be 32 defined at the foundation of the nuclear power plant structures has '
33 often led to confrontations with many in the engineering commur,ity 34 who regard this stipulation as inconsistent with sound practice.
35 l 36 OBJECTIVES 37 38 Reactor Sitina Criteria (Nonseismic) 39 i
40 The objective of the proposed regulatory action is to provide a stable regulatory '
41 basis for the siting of nuclear power plants by decoupling decisions of site !
42 suitability from those affecting plant design. '
43 44 This will be accomplished by:
45 46 a. stating directly those site criteria that experience and 47 importance to risk have demonstrated that future sites should 48 meet and 49 50 b. relocating requirements that apply to reactor design from Part '
51- 50 to Part 100.
52 53 The major changes associated with the revision of the regulation are:
54 55 1. The proposed regulatory action will apply to applicants who apply RA - 3
4 e ;
I 1 .for a construction or early site permit on or after. the effective !
2 date of the final regulations. The current regulation will remain :
3 in place and be applicable to all licensees and applicants prior to !
4 the effective date of the final regulations. '
5 6
7 2. Part 100 will state directly those criteria applicable to the site 8 (e.g. exclusion area distance, population distribution).
9 10 3. Criteria such as source term and dose calculations would be used for 1 11 - evaluating plant features and not for evaluating site suitability i 12 and will be placed into Part 50 consistent with the location of ;
13 other design requirements in the regulation. ,
14 15 Since the revision to the regulation will.not be a backfit, the licensing bases i 16 for existing nuclear power plants must remain in the regulation. Therefore, the 17 revised regulation will be designated as a new subpart to Part 100 for future ,
18 plants while the current Part 100 is maintained for existing plants. I 19-20 Finally, in support of the above changes, Regulatory Guide 4.7 has been revised.
21 22 Seismic Sitina and Earthauake Enaineerina Criteria 23 '
24 The objectives of the proposed regulatory action are to:
25 26 1. Provide a stable regulatory basis for seismic and geologic siting !
27 and applicable earthquake engineering design of future nuclear power 28 plants that will avoid licensing delays due to unclear regulatory :
29 requirements-30 t 31 2. Provide a flexible structure to permit consideration of new l 32 technical understandings; and '
33 ?
34 3. Have the revision to'the regulation completed prior to the receipt 35 of an early site application.
36 37 The major points associated with the revision of the regulation are: I 38 39 1. The proposed regulatory action will apply to applicants who apply ,
40 for an early site permit, design certification, or combined license i 41 (construction permit and operating license) pursuant to 10 CFR Part '
42 52, or a construction permit or operating license pursuant to 10 CFR 43 Part 50 on or after the effective date of the revised regulation. !
44 However, if the construction permit was issued prior to the 45 effective date of the regulation, the operating license applicant :
46 must comply with the seismic and geologic siting and earthquake :
47 engineering criteria in Appendix A to 10 CFR Part 100. .
48 49 2. Criteria not associated with the selection of the site or !
50 establishment of the safe shutdown earthquake ground motion have 1 51 been placed into Part 50. This action is consistent with the 52 location of other design requirements in Part 50. i 53 i 54 Beciuse the. revised criteria presented in the proposed regulation will not be 55 applied to existing plants, the licensing bases for existing nuclear power plants
~
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, , , , ,- ~ wn,,- _m -. -- n ----4
k b -
r 1 must remain part of the regulations. Therefore, the proposed revised criteria 2 on seismic and geologic siting would be designated as a new Appendix B to 10 CFR 3 Part 100 and would be added to the existing body of regulations.
4 5 Earthquake engineering criteria will be located in 10 CFR Part 50 in a new 6 Appendix S. Since Appendix S is not self executing, applicable sections of Part 7 . 50 (i .e. , 150.34, 550.54) will be revised to reference Appendix S.
8 ,
9 The proposed rule would also make conforming amendments to 10 CFR Parts 52 and '
10 100. '
11 12 Finally, in support of the above changes, several regulatory guides and standard 13 review plan sections will be revised or developed as appropriate.
14 15 ALTERNATIVES 16 17 Reactor Sitino Criteria (Nonseismic) 18 19 The alternatives considered included:
20 e No action (e.g., continue to use existing Part 100) 21
- Delete the existing Part 100 and replace it with an entirely new 22 Part 100 that eliminates the dose calculation and specifies site 23 criteria.
24
- Retain the existing Part 100 for current plants and add a new 25 section to Part 100 for future plants that eliminates the dose 26 calculation and specifies site criteria.
27 28 The first alternative considered by the Commission was to continue using current 29 regulations for site suitability determinations. This is not considered an 30 acceptable alternative. Accident source terms and dose calculations currently 31 influence plant design requirements rather than siting. It is considered 32 desirable to be able to state directly those siting criteria which, through 33 importance to risk, have been shown to be key to assuring public health and 34 safety. Further, significant advances in the earth sciences and in earthquake 35 engineering, that deserve to be reflected in the regulations, have taken place 36 since the promulgation of the present regulation.
37 38 Deletion of the existing regulation also is not considered an acceptable 39 alternative since it is the licensing bases for virtually all the operating 40 nuclear power plants and those in various stages of obtaining their operating 41 license.
42 43 Therefore, the last option is the preferable course of action and is the option i 44 evaluated further in this analyses.
45 .
46 Seismic Sitino and Earthouake Enoineerino Criteria 47 48 The first alternative considered by the Commission was to avoid initiating a 49 rulemaking proceeding. This is not an acceptable alternative. Although the 50 siting related issues associated with the current generation of nuclear power 51 plants are completed or nearing completion, there is a renewed sense of urgency 52 to initiate the proposed regulatory action in light of the current and future 53 staff review of advanced reactor seismic design criteria. The current regulation 54 has created difficulties for applicants and the staff in terms of inhibiting 55 flexibility in applying basic principles to new situations and using evolved RA - 5
- - -_. - . _ . - - _- - - _ - _ - - . . _ - _ _ ~
Q 4 l 1 methods- of analysis in the licensing process.
2 3
4 A second alternative considered was the deletion of the existing regulation (Appendix A to Part 100). This is not an acceptable alternative because these 5
6 provisions form part of the licensing bases for many of the operating nuclear power plants and others that are in various stages of obtaining their operating 7 license. Also, geologic and seismic siting criteria are needed for future 8 plants.
9 10 Since there are problems with implementing the existing regulation (Appendix A 11 12 to Part 100), the only satisfactory alternative is to revise the regulation. The approach of establishing the revised requirements in a new Appendix B to Part 100 13 and Appendix S to Part 50 while retaining the existing regulation was chosen as 14 the best alternative.
15 l
16 Finally, the following memoranda' or reports provide further support for a l 17 revision to Appendix A to Part 100:
18 19 1.
20 Staff Requirements Memorandum from Chilk to Taylor dated January 25, 1991,
Subject:
SECY-90-341 - Staff Study on Source Term Update and l 21 Decoupling Siting from Design (Ref. 3).
22 23 "The staff should furtner ensure that the 24 revisions to Appendix A of Part 100 are 25 available to support the time schedule 26 shown in the paper (Commission Briefing on 27 28 Source Term Update and Decoupling Siting from Design (SECY-90-341), dated December 29 13,1990] for option 2, and are technically 30 supportable with the information that will 31 32 be available at the time the draft comes forward for Commission action."
33 34 2. Memorandum from Taylor to Beckjord dated September 6,1990,
Subject:
35 36 Revision of Appendix A, 10 CFR Part 100, " Seismic and Geologic 37 Siting Criteria for Nuclear Power Plants" (Ref. 4).
38 "I approve of your plan to begin work on 39 the development of a revised regulation and 40 41 this activity should be assigned a high priority status."
42 43 3. NUREG-0625, Siting Policy Task. Force (Ref. 5).
44 45 " Revise Appendix A to 10 CFR Part 100 to 46 better reflect the evolving technology in 47 assessing seismic hazards."
48 49 4. NUREG-1061, " Report of the U.S. Nuclear Regulatory Commission Piping 50 Review Committee," Vol 5, April 1985 (Ref. 6).
51 52 "The Committee recommends that 53 54 o Rulemaking amending Appendix A to 10 55 CFR Part 100 be undertaken to permit RA - 6
u a 1 decoupling of the OBE and SSE... ."
12 3 CONSEOUENCES
'4 5- .a. Costs and Benefits 6
7 Benefits 8
9 Reactor Sitina Criteria (Nonseismic) 10 11 The revision to Part 100 will be beneficial to all. The industry and the public 12 will benefit from a clearer, more uniform and consistent licensing process.
13 14 Benefits to the industry, the public, and the NRC staff will result from the
.15 following changes:
16 17 1. Clear Statement of Site Criteria. The proposed revision to Part 100 18 provies clear criteria regarding acceptable exclusion area distances and 19 population distribution. Applicants will be able to select sites that 20 meet these criteria without having to be dependent upon a reactor design.
21 In addition, the criteria have been selected to be consistent with past 22~ experience and with the quantitative health objectives in the NRC Safety 23 Goal Policy.
24 25 2. Current Practices Will Be Reflected. The proposed regulations reflect 26 industry design practices and the associated staff review procedures that 27 have evolved since Part 100 was issued in 1962. An example of this is the t 28 review. of nearby industrial and transportation facilities which will be 29 incorporated into the regulations for the purpose of site suitability and 30 has been a part of the staff review for many years. The criteria and 31 standards are the same as those currently in staff review guidance 32 documentation (Standard Review Plan, etc.). Hence, the proposed rule 33 involves no substantive changes in this area and merely codifies what has 34 been staff practice for a number of years. Additionally, the numerical 35 population density values and the exclusion area distance outlined in 36 Regulatory Guide 4.7 will be codified in the proposed rulemaking.
37 38 3. Source Term and Dose Calculations. The proposed rule would eliminate the 39 use of a postulated source term, assumptions regarding mitigation systems 40 and dispersion factors, and the calculation of radiological consequences 41 to determine the sizes of the exclusion area and low population zone. It 42 would instead require a minimum exclusion area distance.
43 44 4. Text Clarification and Elimination of Low Pooulation Zone. The Commission 45 considers that the functions intended for the " low population zone,"
46 namely, a low density of residents and the feasibility of taking 47 protective actions, have in fact been overtaken by other regulations or 48 can be accomplished by other means. Protective action requirements are 49 defined via the use of the Emergency Planning Zones (EPZ), while 50 restrictions on population close to the plant can be assured via proposed 51 population . density criteria. For these reasons, the Commission is i 52 proposing to eliminate the requirement d a low population zone for future power reactor sites.
~
53-54 55 In addition, the proposed rule would require that important site factors RA - 7
m _ _ _ . - _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
o e !
I such as population distribution, topography, and transportation route's be 2
3 considered and examined in order to determine whether there are any site t 4 characteristics that could pose a significant impediment to the develop-ment of an emergency plan. This proposed requirement is also consistent 5 with 10 CFR Part 52.
6 7 l 8 Planning for emergencies is part of the Commission's defense-in depth approach. ;
9 The Commission concludes that site characteristic that may )
10 represent an impediment to development of adequate emergency plans, such l 11 as limitations of access or egresses in the immediate vicinity of a 12 nuclear power plant should be identified at the early stage of site approval rather than at a later date prior to operation thus avoiding 13 significant licensing delays.
14 15 5. Risk to the Public. The.NRC Staff hn: generated a reduced set of source 16 17 terms based on the NUREG-1150 (Ref. 2) analyses and the Independent Risk Assessment Plant.
18 These source terms were used in the MELCOR Accident 19 Consequences Code System (MACCS) for six reactor-containment designs. The 20 results of these . analyses indicate that the risk to the public is 21 acceptably low and the guidelines of.the Commission's Safety Goal Policy 22 are met for all plants up to 3800 MWt, the largest capacity plant considered in the analyses.
23 24 25 Seismic Sitina and Earthouake Enaineerina Criteria 26 27 The revision of Appendix A to Part 100 will be beneficial to all. The public 28' will benefit from a clearer, more uniform and consistent licensing process subject to fewer interpretations. The NRC staff will benefit from improved 29 30 regulatory implementation (both technical and legal), fewer interpretive debates, and increased regulatory flexibility. Applicants will derive the same benefits 31 32 in addition to requirements. avoiding licensing delays because of unclear regulatory 33 34 35 The proposed regulatory action reflects changes intended to (1) benefit from the experience gained in applying the existing regulation; (2) resolve interpretative l 36- ;
37 questions; (3) provide needed regulatory flexibility to incorporate state-of-the-38 art improvements in the geosciences and earthquake engineering; (4) simplify the language to a more " plain English" text; and (5) acknowledge various internal l 39 staff and industry comments.
40 41 42 Benefits to applicants or NRC staff will result from the following changes:
43 1. Define seismic sources.
44 Better definition of seismic source types will eliminate a major source of licensing delays.
45 ,
46 2. Use of both deterministic and probabilistic evaluations. i' 47 The proposed regulation would require a single approach making use of 48 both deterministic and probabilistic evaluations.
49 The staff 50 proposes to use both the deterministic (currently being used) and !
the probabilistic evaluations together and compare the results of 51 each to provide insights that would be unavailable if either were 52 used alone. The principal limitations of the deterministic 53 evaluation --- its ability to incorporate only one model and one 54 data set at a time and its inability to allow weighted incorporation 55 of numerous models --- can be assessed by comparing its results with RA - 8
I t, ' p-
- l I
t l 'I the results of a probabilistic evaluation accomplished in parallel.
! 2 Similarly, the principal limitation of the probabilistic evaluation !
3 l --- its tendency to allow its results to be dominated by the tails :
4 rather. than the central tendency of distributions of uncertain '
5 l knowledge or expert opinion --- can be assessed by comparing its !
t
-6 L 7 results with the results of one or more deterministic evaluations. '
8 The staff believes that the approach of combining both evaluations 9
would allow more informed judgments as to what the appropriate Safe l 10 Shutdown Earthquake Ground Motion (SSE) should be for a given site. '
11 Therefore, it is the staff's opinion that this approach is the best !
12' way to accomplish the objective of this . aspect of the revised +
13 l
14 regulation, which is to arrive through analysis at a site-specific ground motion that appropriately captures what-is known about the :
15 seismic regime. This approach, using both - probabilistic and
'16 1 deterministic evaluations, will thus lead to a more stable and ;
"17 predictable licensing process then in the past.
'18 19 3. Reflect current design practices. The proposed regulations would 20 reflect' industry design practices and the associated staff review' '
21 procedures (for instance, the location of the control point for the 22 j seismic input) that have evolved since the initial regulation
-23 (Appendix A to Part 100) was issued in 1973. Many of these i
24 practices and procedures were incorporated into the~ revision of 25 l Standard Review Plan Sections 2.5.2, 3.7.1, 3.7.2, and 3.7.3 that l 26 27 are associated with the resolution of Unresolved Safety Issue (USI) )
A-40, " Seismic Design Criteria." -
28
, 29 4. Clarify the multi-facets associated with the Operating -Basis
! 30- Earthquake (OBE). In the existing regulation, the OBE is associated 31 with (1) the functionality of those features necessary for continued 32 operation without undue risk to the health and safety of the public, i 33 (2) an earthquake that could reasonably be expected to affect the j 34 plant site during the operating life of the plant, (3) a minimum
( 35 fraction of the- Safe Shutdown Earthquake (SSE), and (4) plant
- 36 shutdown if the vibratory ground motion is exceeded. In some cases, l
37 for instance, piping, the multi-facets of the OBE made it possible l 38 for the OBE to have more design significance than the SSE. The 39 seismological basis, that is, the association of the OBE with a
( 40 likelihood of occurrence has been removed from the proposed' 41 regulation. Other facets of the OBE, for instance, its value 42- (percent of the SSE) - and relationship with plant shutdown are 43 discussed below. The functionality aspect of the OBE remains 44- unchanged.
! 45 46 5. Value of the Operating Basis Earthquake Ground Motion (OBE) and
'47 required (OBE) analysis. ~ The proposed regulation would allow the 48 value of the OBE to be set at (i) one-third or less of the SSE, 49 where OBE requirements are satisified without an explicit response 50 or design being performed, or'(ii) a value greater than one-third of i- 51 the.SSE, where analysis and design are required. There are two 52 issues the applicant should consider in selecting the value of the j 53' .OBE: first, plant shutdown is required if vibratory ground motion
- 54 exceeding that of .the OBE occurs (discussed below in Item 6, 55 Required Plant Shutdown), and second, the amount of analyses RA - 9 i
, . 4 d I associated with the OBE. An applicant may determine that at i 2 one-third of the SSE level, the probability of exceeding the OBE
, 3 vibratory ground motion is too high, and the cost associated with 4 plant shutdown for inspections and tests of equipment and structures 5 prior to restarting the plant is unacceptable. Therefore, the 1 6 applicant may voluntarily select an OBE value at some higher j 7 fraction of the SSE to avoid plant shutdowns. However, if an i 8- applicant selects an OBE value at a fraction of the SSE higher than.
9 one-third, a suitable analysis shall be performed to demonstrate
.10 that the requirements associated with the OBE are satisfied. The 11 design shall take into account soil-structure ' interaction effects ;
12 and . the expected duration of the vibratory ground motion. The ,
13 requirement associated with the OBE is that all structures, systems, l 14' and components of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the public 15 '
'16 shall remain functional and within applicable stress and deformation '
17 limits when subjected to the effects of the OBE in combination with '
- 18. normal operating loads. As stated above, . subject to further 19 confirmation, it is determined that if an OBE of one-third of the :
20 SSE is used, the requirements of the OBE can be satisfied without 21 the applicant performing any explicit response analyses. However, 22 some minimal design checks (additional discussion below) must be 23 performed. There is high confidence that, at this ground-motion 24 level, with other postulated concurrent loads, most critical 25 structures, systems, and components will not exceed currently used 26 design limits. In this case, the OBE serves the function of an 27 inspection and shutdown earthquake. There are situations associated 28 with current analyses where only OBE is associated with the design s 29 requirements, for example, the ultimate heat sink (see Regulatory >
30 Guide 1.27, " Ultimate Heat Sink for Nuclear Power- Plants"). In 31 these situations, a value expressed as a fraction of the SSE
-32 response would - be used in the analyses. The section " Future 33 Regulatory Action" of this Regulatory Analysis identifies existing '
-34 guides that would be revised to maintain the existing design 35 philosophy. With regard to piping analyses, positions on fatigue 36 ratcheting and seismic anchor motion are being developed and will be 37 issued for public comment in a draft regulatory guide separate from 38 this rulemaking.
39 40 Activities equivalent to OBE-SSE decoupling are also being done in 41 foreign countries. For instance, in Germany their new design 42 standard requires only one design basis earthquake (equivalent to '
43 the SSE). They require an inspection level earthquake (for
-44 shutdown) of 0.4 SSE. This level was set so that the vibratory 45 ground motion should not induce stresses exceeding the allowable 46 stress limits originally required for the OBE design.
47 48 6. Guidance'for required plant shutdown. The proposed regulation would 49 treat plant shutdown associated with vibratory ground motion 50 exceeding .the OBE or significant plant damage as a condition in 51 every operating license. The shutdown requirement would be a
- 52 53 condition of the license (10 CFR 50.54) rather than a limiting 54 condition of operation (10 CFR 50.36), because the necessary judgements associated with exceedance of the vibratory ground motion '
55 or significant plant damage can not be adequately characterized in '
RA - 10 ,
w a '
I a technical specification. A new paragraph, 650.54(ee) would be 2 added to the regulations to require plant shut down for licensees of 3 nuclear power plants that comply with the earthquake engineering 4 criteria in Paragraph IV(a)(3) of Proposed Appendix S, " Earthquake Engineering Criteria for Nuclear Power Plants," to 10 CFR Part 50.
5 6 Draft Regulatory Guide DG-1017, " Pre-Earthquake Planning and 7 Immediate Nuclear Power Plant Operator Post-Earthquake Actions," is 8 being developed to provide guidance acceptable to the NRC staff for 9 determining whether or not vibratory ground motion exceeding the OBE 10 or significant plant damage had occurred and nuclear power plant ,
11 shut down is required. The guidance is based on criteria developed 12- by the Electric Power Research Institute (EPRI). Draft Regulatory 13 Guide DG-1018, " Restart of a Nuclear Power Plant Shut Down by a 14 Seismic Event," is being developed to provide guidelines that are 15 acceptable to the NRC staff for_ performing inspections and tests of 16 a nuclear power plant equipment and structures prior to plant 17' restart. This guidance is also based on EPRI reports.
18 19 7. Reduced level of detail. The level of detail presented in the 20 proposed regulations has been limited to general guidance. The 21 proposed regulations would identify and establish basic require-22 ments. Detailed guidance, that is, the procedures acceptable to the 23 NRC for meeting the requirements, has been removed and placed in 24 Draft Regulatory Guide, DG-1015, " Identification and Characteriza-25 tion of Seismic Sources, Determinisite Source Earthquakes, and 26 Ground Motion."
27 28 8. Provide greater flexibility. The proposed regulations would provide 29 ' a flexible structure that will permit the consideration of new l
30 technical understandings and state-of-the-art advancements since ;
31 the detailed guidance has been removed from the proposed regulation 32 and placed into regulatory guides.
33 34 9. Clarify interpretations. Changes have been made to the seismic and 35 geologic siting criteria to resolve past questions of interpreta-36 tion. As an example, the definitions and required investigations .
37 sections of the proposed regulations havre been significantly changed 38 to eliminate or modify phrases that were more applicable to only the 39 western United States.
40 l 41 10. Clarify text. The proposed regulations would use more explicit 42 terminology. For instance, the Safe Shu'.down Earthquake (SSE) and 43 Operating Basis Earthquake are now refereaced as the Safe Shutdown 44 Earthquake Ground Motion (SSE) and the Operating Basis Earthquake 45 Ground Motion (OBE). In addition, appropriate changes within the ,
46 text highlight that the SSE used as the design basis is not 47 associated with a single earthquake but may be a composite of 48 several expected earthquakes.
- 49. ,
50 '
51 [giti 52 53 Reactor Sitina Criteria (Nonseismic) l- 54 1
55 The costs associated with the revised regulations are subdivided into two '
RA - 11
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_-_m __m__._____.._-..._____-._m__a___
s c i
1 categories; the first is associated with siting criteria modifications (Part 2 100), the second is associated with (Part 50) modifications.
3 4 Part 100 5
6 The overall cost impact associated with revising the siting criteria aspects of 7 the regulation are neutral. Important factors in this regard are:
8 9 1. Defining a Minimum Exclusion Area Distance _and Eliminating Dose 10 Calculations.
11 The present regulation has no numerical size .
12 requirement for the exclusion area, in terms of distance, and :
instead assesses the consequences of a postulated radioactive 13 14 fission product release within containment, coupled with assumptions 15 regarding containment leakage, performance of certain fission 16-product mitigation systems and site meteorology for a hypothetical 17 individual located at any point on the exclusion area boundary as well as hydrological information. The plant and site combination is 18 19 considered to be acceptable if the calculated consequences do not 20 exceed the values given in the present rule. Regulatory Guide 4.7 <
21 suggests an exclusion area distance of 0.4 miles, since this, in conjunction with typical engineered safety features, has been found 22 to meet the dose values in the existing rule.
23 24 25 The Commission considers an exclusion area to be an essential 26 feature of a reactor site and is retaining this requirement for future reactors. However, in keeping with the recommendation of the 27 Siting Policy Task Force to decouple site requirements from reactor ,
28 !
29 design, the prorosed rule would eliminate the use of a postulated source tar::;, assumptions regarding mitigation systems and 30 meteorology, and the calculation of radiological consequences to 31 1 determine the sizes of the exclusion area and low population zone.
32 It would instead require a minimum exclusion area distance of 0.4 ;
33 miles for reactors.
34 ,
35 36 The proposed approach of eliminating the use of a postulated i
37 accident source term and the use of dose calculations for 38 determining the acceptability of a site and replacing these with population criteria and a minimum size of the exclusion area is 39 expected to reduce time and costs associated with obtaining site 1 40 approval.
41 42 2. Nearby Industrial and Transportation Facilities. This area of i 43 review is proposed to be incorporated into the regulations for the 44 !
purpose of site suitability and has been a part of the staff review l 45 for many years. The criteria and standards are the same as those 1 46 currently in staff review guidance documentation (Standard Review 47 Plan, etc.). Hence, the proposed rule involves no substantive 48 changes in this area and merely codifies what has been staff 49 practice for a number of years.
50 51 3. Feasibility of Carrying out Protective Actions. The proposed rule 52 would require that important site factors, such as population 53- distribution, topography, and transportation routes, be considered 54 and examined in order to determine whether there are any site 55 characteristics that would pose a significant impediment to the RA - 12
6 3.
I development of an emergency plan. ~
2' 3 The cost impact associated with this revision is neutral. The 4 revision is expected to increase time and costs for site approval 5 but should significantly reduce time and costs at the OL or COL '
6- stage by avoiding licensing delays.
7-8 Part 50 9
10 The overall cost impact associated with revising the reactor licensing aspects !
11 .of the regulation are neutral because the source term and dose calculations have ,
12 always been required under Part 100 for site suitability but will now be required i 13 under Part 50 and used in evaluating plant features, therefore, there is no >
14 change in cost. l 15 l 16 Seismic Sitina and Earthouake Enoineerina Criteria l 17 :
18 The costs associated with the proposed regulations are subdivided into two l 19 categories; the first is associated with the geosciences and site investigations ,
20 (Appendix B to Part 100), the second is associated with earthquake engineering 21 (Appendix S to Part 50).
22 23 Anoendix B to Part 100 l 24 25 The overall cost impact associated with the geosciences and site investigation .
26 aspects of the proposed regulation as compared to Appendix A of Part 100 are >
- 27. slightly increased in some areas but reduced overall because of anticipated 28 improvement in the licensing process. Specific examples include:
29 30 1. Reduced Licensing Delays. The licensing process will be enhanced i 31 because information needed for the staff review can be incorporated l 32 in the safety analysis reports at the time of docketing instead of 33 later through staff questions and applicant responses.
34 35 2. Probabilistic Evaluations. Probabilistic evaluations to determine 36 vibratory ground motion, surface tectonic deformation, and 37 seismically induced floods and water waves reflect to some extent 38 what is already current staff practice. In particular, probabi-39 listic hazard analyses have been used to determine the probability 40 of exceeding the Safe Shutdown Earthquake Ground Motion at the plant 41 site. However, the overall use of probabilistic evaluations as 42 suggested in Draft Regulatory Guide DG-1015, " Identification and 43 Characterization of Seismic Sources, Deterministic Source Earth-44 quakes, and Ground Motion," is new but should not have a significant 45 cost impact. Computer codes to perform the probabilistic analyses 46 are available. An applicant would input the site coordinates and 47- local site effects (current requirement) to obtain the probabilistic 48- hazard data. It is estimated that these analyses can be performed 49- within a few days.
50 51 The comparison between the deterministic (current requirement) and 52' probabilistic evaluation is new. In cases where it is judged that 53 the deterministic and probabilistic evaluations provide equivalent 54 results the process is complete. In cases where the results differ, 55 spectra are developed to make additional comparisons. Evaluations !
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1 . associated with these comparisons would be handled on an ad' hoc ,
2 basis. However, as stated above, licensing delays would be reduced j
'3 because the required data are defined and available to the applicant !
4 and staff for evaluation. i 5 1 6 As part of the Federal Reaister notice, public comments on specific 7- questions associated with the use of this approach, both 8 probabilistic and deterministic evaluations and the comparison 9 l procedure recommended by the NRC staff, are requested. '
10 11 3. Seismic Sources. .The new approach towards seismic sources (using 12 seismogenic sources instead of tectonic provinces) and other '
13 clarifications of the licensing approach are expected to reduce time f 14 and costs required for obtaining site approval.
15 l 16 Appendix S to Part 50 1 17 '
18_ The overall cost impact associated with the earthquake engineering aspects of the 19 proposed regulation are neutral or reduced. Specific examples include:
20 21 1. Reduced OBE Analysis. The response analyses -associated with 'the 22 Operating Basis Earthquake Ground Motion (0BE) may be eliminated if 23 the applicant sets the OBE at one-third of the Safe Shutdown 24 Earthquake Ground Motion'(SSE). Selecting an OBE value greater than 25 one-third of the SSE does not increase the analytical effort above 26 current requirements.
27 28 2. Control Point Location. Changing the location of the control point 29 (the point at which the vibratory ground motion is ar? lied) from the .
30 31 foundation level to the free-field does not affect costs. The following discussion from Section 2.1.1.4 of NUREG-1233 (pages 13 32 and 14) is applicable: 3
-33 34 "A number of recent plants were designed to 35 the 1975 Standard Review Plan reouirements 36 which specified the free-field nation at ,
37 the free-surface for soil-st ructure 38 interaction analysis. During the operating 39 license (OL) review, the implementatimi of 40 the current position of input motion at the 41 foundation level in the free field resulted 42 in a modification of some structural floor 43 beams of seismic Category I structures at "44' one plant. No hardware changes resulted at 45 other plants. (Note that the staff's 46 investigation was limited to the Safe 47 shutdown systems and structures that housed 48 them, and allowance was made for tested 49 strength values in some cases.)"
50 51J 3. Seismic Instrumentation.. Although the seismic instrumentation 52 requirements are different (only time-history accelerographs instead 53 of time-history accelerographs, response spectrum recorders and peak 54- accelerographs), the cost is essentially the same as that associated 55 with operating plants; there are fewer instruments required. The RA - 14 t _ _ . _ _ . . _ --
(= 0 , 1 l
[
1 2 maintenance and calibration costs with the new solid-state seismic 3 instrumentation are less than that associated with the current instrumentation.
4 The processing of instrumentation data will be l-5 done at the site, thereby reducing the potential for prolonged plant
- 6. shutdown while data are being evaluated. In general, the ability to 7 expeditiously assess the effects of the earthquake on the plant will 8
save both staff and licensee resources.
9 4. Post-Earthquake Activities. In preparation of post-earthquake 10 activities, l
11 it is recommended that the 11cen ee inspect and 12 base-line certain structures, equipment and piping. Base line i 13 inspections would differentiate between pre-existing conditions at
! 14 the nuclear power plant and earthquake related damage. The struc-
! 15. tures, . equipment and piping selected for these inspections are
! 16 comprised of those routinely examined by plant operators during i 17 normal plant walkdowns' and inspections. After an earthquake, plant 18 operators familiar with the. plant would walkdown and visually l inspect accessible areas of the plant. Unnecessary plant shutdowns 19 20 would be avoided since the pre-earthquake condition of equipment 21 and structures (for example, physical appearance, leak rates,
'22 vibration levels) would be known. This approach has been submitted l
23 to the NRC staff for approval by the Nuclear Management and !
i 24 Resources Council (NUMARC) and is documented in an Electric Power !
! 25 Research Report, EPRI NP-6695, " Guidelines for Nuclear Power Plant- i 26 Response to an Earthquake." The associated cost impact is minimal and recommended by industry. '
27 28 IMPACTS L 29 a. Other NRC Proarams
! 30 l 31 None for the Nonseismic siting criteria.
! 32
! 33 34 Although Appendix A to 10 CFR Part 100 is titled " Seismic and Geologic 35 Siting Criteria for Nuclear Power Plants," it is also referenced in two 36 other parts of the regulation. They are (1) Part 40, " Domestic Licensing 37 of Source Material," Appendix A, " Criteria Relating to the Operation of <
38 Uranium Mills and the Disposition of Tailings or Waste Produced by the 39 Extraction or Concentration of Source Material from Ores Processed 40 Primarily for Their Source Material Content,"Section I, Criterion 4(e),
G and (2) Part 72, " Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-level Radioactive Waste," Paragraphs (a)(2)(b) 42 and (a)(2)(f)(1) of 572.102. The proposed regulation, Appendix B to Part 43 100, is still applicable only to nuclear power plants. The revision of 44 Part 72 and Appendix A to Part 40, subject to the implementation of 45' Appendix B to Part-100, should be a separate rulemaking initiative.
46
, 47 .b. Other Government Acencies
, 48
! 49 Since the siting and licensing of nuclear power plants is carried out 50 solely by NRC staff, no impact is projected for other government agencies.
j :
51
! 52 c. Constraints
. 53 54 None.
i 55 RA - 15 L
1
/
- e a 1 DECISION RATIONALE-2
- 3 Reactor Sitina Criteria (Nonseismic) 4 5 The major considerations that have guided the Commission in this proposed 6 revision to.the reactor site criteria are as follows:
7 8 1. -The criteria will assure a low risk for individuals as well as for 9 society in general, even in the event of severe but unlikely reactor 10 accidents. The proposed criteria are consistent with the Commission 11- Safety Goal Policy with respect to the risk of both prompt and 12 latent cancer fatalities. In addition, the Commission has examined 13 the risks associated with land contamination or property damage in 14 the event of significant releases of long-lived radioactive species, 15 such as cesium. The proposed criteria are expected to result in a 16- low likelihood of any significant offsite contamination' of densely 17- populated areas.
28 19 2.- The criteria will assure that man-made activities as well as natural 20 events associated with the site location are identified and used in 21 matching a design with the site.
22 23 3. The criteria will assure that a range .of protective actions can 24 feasibly be carried out to protect the public in the event of ,
l 25 emergency.
l- 26 The proposed revisions reflect current staff practice. The revised l
27 28 regulations will not reduce risk, bet would improve the description in the 29 regulations of current staff practice in licensing.
l 30 l 31 Seismic Sitina and Earthauake Enaineerina Criteria 32 33 The recommendations to revise the existing regulation (Appendix A to 10 CFR Part 34 100) and replace it with the regulations pertaining to the geosciences and site 35 investigations (Appendix B to Part 100) and earthquake engineering (Appendix S 36 to Part 50) are based primarily on qualitative rather than quantitative or :
37 probabilistic (i.e., core damage frequency reduction) arguments. The staff's '
38 evaluation augments the regulatory analysis associated with the implementation '
39 of Unresolved Safety Issue (USI) A-40, " Seismic Design Criteria" (NUREG-1233, 40 Ref. 7). USI A-40 was implemented in August 1989 through the revision of 41 Standard Review Plan Sections 3.7.1, " Seismic Design Parameters," 3.7.2, " Seismic 42 System Analysis," 3.7.3, " Seismic Subsystem Analysis," and 2.5.2, " Vibratory 43 Ground Motion." ;
44-45- The staff's conclusion is that for' operating reactor and operating license -
46 applicants, the proposed regulations would have little effect on risk. Operating '
~47 plants generally have been, and will be, seismically upgraded by plant-specific -
48 actions such as implementation of the Systematic Evaluation Program (SEP), the i l 49 implementation of Generic Letter 88-20, Supplement 4 " Individual Plant 50 Examinations of External Events (IPEEE) for Severe Accident Vulnerabilities," the ;
51 . proposed implementation of USI A-46, " Verification of Seismic Adequacy of 52 Equipment in Operating Plants," and NRC Bulletin programs. Therefore, this ,
53 regulatory action will be applicable only to applicants who apply for an early 54 site permit, design certification, combined license, construction permit or 55 operating license on or after the effective date of the final regulations.
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1 2
No overall increases in costs are expected to impicmnt the proposed regulat' ions 3 for applicants for early site permits, design certifhations, combined licenses, 4
construction permits or operating license. In additi.n, the proposed regulations 5
will reduce delays in the licensing proces. because information needed for the 6
staff review can be incorporated in the safot, analysis reports at the time of
'7 docketing instead of later through staff .1uestions and applicant responses.
8 Therefore, the proposed theregulations.
staff proposes that_ all new applicants be required to comply with l
9 i 10 Current Reaulatory Action 11 12 13 The current-regulatory action consists of the following:
14 1.
i 15' Revisions to 550.2, 650.8, 650.34, 550.54, and 652.17.
l l 16 2. Revisions to 5100.1, 6100.2, 6100.3, and 1100.8.
! 17 18 3. Add Subpart B, 1100.20, 6100.21, and 1100.22.
19.
20 4.
21 Add a new Appendix B to Part 100, Criteria for the Seismic and l 22 Geologic Siting of Nuclear Power Plants After [ EFFECTIVE DATE OF THIS REGULATION]
23 24 5.
25 Add a new Appendix S to Part 50, Earthquake Engineering Criteria for Nuclear Power Plants 26 27 6. Issue new Regulatory Guides for public coment:
1- 28 i 29 a. DG-1015, " Identification and Characterization of Seismic 30 31 Sources, Deterministic Source Earthquake, and Ground Motion"
! 32 b.
33 DG-1017, " Pre-Earthquake Planning and Imediate Nuclear
! Power Plant Operator Post-Earthquake Actions" 34 35 -- c. DG-1018, " Restart of a Nuclear Power Plant Shut Down by a 36 Seismic Event" 37 38 39 7. Issue Revised Regulatory Guides for public coment:
40
- 41 a. Proposed Revision 2 to Regulatory Guide 4.7, " Genera! Site 42 Suitability Criteria for Nuclear Power Stations" 43 44 b. DG-1016, Second Proposed Revision 2 to Regulatory Guide 1.12,
'45
" Nuclear Power Plant Instrumentation for Earthquakes" 46 47 8. Issue Revised Standard Review Plan Section for public coment:
48 49 2.5.2, Vibratory Ground Motion.
. 50-
[ 51 4 52 Future Reaulatory Action
- 53
, 54 Several existing regulatory guides will be revised to incorporate editorial 55 changes or maintain the existing design or analysis philosophy. These guides RA - 17
i e I will be issued subsequent to the publication of the final regulations that w~ould 2 implement this proposed action.
3 4 The following regulatory guides will be revised to incorporate editorial changes 5 The type of changes contemplated would be to reference new paragraphs in Appendix 6 B to Part 100 or Appendix S to Part 50:
7 8 1. 1.57, " Design Limits and Loading Combinations for Metal Primary 9 Reactor Containment System Components" 10 11 2. 1.59, " Design Basis Floods for Nuclear Power Plants" 12 13 3. 1.60, " Design Response Spectra for Seismic Design of Nuclear Power 14 Pl ants" 15 16 4. 1.83, " Inservice Inspection of Pressurized Water Reactor Steam 17 Generator Tubes" 18 19 5. 1.92, " Combining Modal Responses and Spatial Components in Seismic 20 Response Analysis" 21 22 6. 1.102, " Flood Protection for Nuclear Power Plants" 23 24 7. 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes" 25 26 8. 1.122, " Development of Floor Design Response Spectra for Seismic 27 Design of Floor-Supported Equipment or Components" 28 29 The following regulatory guides will be revised to maintain existing design 30 or analysis philosophy. For example, the types of changes contemplated would be 31 to change OBE to a fraction of the SSE.
32 l
33 1. 1.27, " Ultimate Heat Sink for Nuclear Power Plants" 34 35 2. 1.100, " Seismic Qualification of Electric and Mechanical Equipment i 36 for Nuclear Power Plants" l 37
! 38 3. 1.124, " Service Limits and Loading Combinations for Class 1 Linear-
- 39 Type Component Supports" 40 41 4. 1.130, " Service Limits and Loading Combinations for Class 1 Plate-
[ 42 and-Shell-Type Component Supports" l 43
! 44 5. 1.132, " Site Investigations for Foundations of Nuclear Power Plants" l 45
- 46 6. 1.138, " Laboratory Investigations of Soils for Engineering Analysis l 47 and Design of Nuclear Power Plants'
- 48 l 49 7. 1.142, " Safety-Related Concrete Structures for Nuclear Power Plants 50 l
51 (Other than Reactor Vessels and Containments)"
52 8. 1.143, " Design Guidance for Radioactive Waste Management Systems, 53 Structures, and Components Installed in Light-Water-Cooled Nuclear 54 Power Plants" l 55 RA - 18
1 If substantive changes are made during the revisions, the applicable guides Will i
2 be issued for public comment as draft guides.
3 4 IMPLEMENTATION 5
6 7
This regulatory action is applicable only to applicants that apply for an early 8
site permit, design certification, combined license, construction permit, or operating license on or after the effective date of the final regulations. If 9
10 the construction permit was issued prior to the effective date of the proposed regulation, the operating license applicant must comply with the seismic and 11 geologic siting and earthquake engineering criteria in Appendix A to Part 100.
12 13 14 l
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( e I REFERENCES 2
~
3 1. U.S. Nuclear Regulatory Commission, " Reactor Safety Study-An Assessment of I 4 Risks in U.S. Commercial Nuclear Power plants," NUREG-75/014 (WASH-1400),
5 December 1975.
6 7 2. U.S. Nuclear Regulatory Commission, " Severe Accident Risks: An Assessment for 8 Five U.S. Nuclear Power Plants," NUREG-1150, December 1990.
9 10 3. Staff Requirements Memorandum from S.J. Chilk to J.M. Taylor, Subject SECY 1 11 341, January 25, 1991.
12 13 4. Memorandum from J.M. Taylor to E.S. Beckjord, Subject Revision of Appendix A, 14 10 CFR Part 100, September 6, 1990. '
15 16 5. U.S. Nuclear Regulatory Commission, " Report of the Siting Policy Task Force,"
17 NUREG-0625, August 1979. .
18 19 6. U.S. Nuclear Regulatory Commission, " Report of the U.S. Nuclear Regulatory 20 Commission Piping Review Committee," NUREG-1061, Volume 5, April 1985. ,
21 22 7. S.K. Shaukat and N.C. Chokshi, " Regulatory Analysis for USI A-40, ' Seismic 23 Design Criteria,'" NUREG-1233, U.S. Nuclear Regulatory Commission, September 24 1989.
25 26 8. Electric Power Research Institute, " Guidelines for Nuclear Plant Response to 27 an Earthquake," NP-6695, December 1989.
28 I
l 4
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ENCLOSURE 3 l
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I DRAFT ENVIRONMENTAL ASSESSMENT AND FINDING OF NO SIGNIFICANT IMPACT l
PROPOSED REVISION OF I 10 CFR PART 100 AND APPENDIX A TO 10 CFR PART 100 l l
l l
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I
1 DRAFT ENVIRONMENTAL ASSESSMENT AND FINDING OF NO SIGNIFICANT IMPACT 2
PROPOSED REVISION OF 10 CFR PART 100. APPENDIX A T0 10 CFR PART 100.
3 AND 10 CFR PART 50 4
5 6 The Nuclear Regulatory Commission is amending its regulations to update the l 7
reactor siting criteria, seismic and geologic siting criteria, and earthquake ;
8 engineering regulations for nuclear power plants. The nonseismic and seismic 1 9 areas are discussed separately.
10 11 12 Identification of Proposed Action 13 14 Reactor Sitina Criteria (Nonseismic) i 15 16 Title 10 CFR Part 100, " Reactor Site Criteria," was originally issued in April 17 1962. The proposed amendment will apply to applicants who apply for site 18 approval on or after the effective date of the final regulation. Since the 1 19 revision to the regulation will not be a backfit, the bases for existing nuclear 20 power plants must remain in the same regulation. Therefore, the revised 21 regulation on siting will be designated Subpart B of 10 CFR Part 100.
22 23 24 Criteria not associated with site selection will be relocated into Part 50 consistent with the location of other design requirements in the regulation.
25 Hence, source term and dose calculations will be used for evaluating plant 26 features and not site suitability.
27 28 The proposed rule would eliminate the use of a postulated accident source term 29 and the use of a dose calculation in the determination of acceptability for a 30 nuclear power plant site. It would also eliminate the designation of a low 31 population zone. Instead, it would set a minimum size for the exclusion area and 32 would set population density criteria around proposed nuclear power reactor 33 sites. In addition, criteria regarding the evaluation of man-made hazards and 34 the feasibility of carrying out protective actions in the event of an emergency 35 would be incorporated.
36 37 Seismic Sitina and Earthauake Enaineerina Criteria 38 39 Appendix A, " Seismic and Geologic Siting Criteria for Nuclear Power Plants," to 40 10 CFR Part 100, " Reactor Site Criteria," was originally issued as a proposed 41 rule on November 25,1971 (36 FR 22601); published as a final rule on November
, 42 13, 1973 (38 FR 31279); and became effective on December 13, 1973. There have l 43 been two amendments to Appendix A to 10 CFR Part 100. The first amendment,
- 44 issued November 27,1973 (38 FR 32575), corrected the final rule by adding the 45 legend under the diagram. The second amendment resulted from a petition for
- 46 rulemaking (PRM 100-1) requesting that an opinion interpreting and clarifying 47 Appendix A with respect to the determination of the Safe Shutdown Earthquake be 48 issued. A notice of filing of the petition was published on May 14, 1975 (40 FR l 49 20983). The substance of the petitioner's proposal was accepted and published 50 as an immediately effective final rule on January 10, 1977 (42 FR 2052).
51 52 The proposed amendment will apply to applicants who apply for an early site 53 permit, design certification, combined license, construction permit, or operating 54 license on or after the effective date of the revised regulation. However, if 55 the construction permit was issued prior to the effective date of the regulation, l
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[
- 4 4
)
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l
, I the operating license applicant shall comply with the seismic and geologic siting i 2 and earthquake engineering criteria in Appendix A to 10 CFR Part 100. Because
'3 the revised criteria presented in the proposed regulation will not be applied to 4 existing plants, the licensing bases for existing nuclear power plants must i remain part of the regulations.- . Therefore, the proposed revised criteria on 6 seismic and geologic siting would be designated as a new Appendix B to 10 CFR 7 Part 100,'" Criteria for the Seismic and Geologic Siting of Nuclear Power Plants 8 After [ EFFECTIVE DATE OF THIS REGULATION]," and would be added to the existing 9 body of regulations.
10
, 11 Criteria not associated with site selection or establishment of the Safe Shutdown 1 l 12 Earthquake Ground Motion (SSE) have been placed into 10 CFR Part 50. This action I l 13 is. consistent with the location of other design requirements in Part 50. Hence, i 14 earthauake engineering criteria would be located in Appendix S to 10 CFR Part 50, 1- 15 " Earthquake Engineering Criteria for Nuclear Power Plants."
16 17 The proposed regulatory action incorporates changes that are intended to (1) i 18 benefit from the experience gained in applying the existing regulation, (2) i 19 resolve interpretative questions, (3) provide needed regulatory flexibility to 20 incorporate improvements in the geosciences and earthquake engineering, and (4) 21 simplify the language to a more " plain English" text.
! 22 l 23
! 24 Need for the Proposed Action
. 25 ;
l 26 Reactor Sitina Criteria (Nonseismic) 27 i 28 Since its initial promulgation in 1962, the Commission has approved more than 75 l l
29' sites for nuclear power plants and has had an opportunity to review a number of i 30- others. As a result of these reviews, much experience has been gained regarding 1
, the site factors that influence risk and their range of acceptability.
33 Additionally, there has also been an increased awareness and concern regarding 34 the effect of potential nuclear accidents. Although accident considerations have i 35. 'been of key importance in reactor siting from the very beginning, major l 36 developments such as the issuance of the Reactor Safety Study (WASH-1400) in i 37 1975, the occurrence of the Three Mile Island accident in 1979, the Chernobyl l 38 accident in the Soviet Union in 1986, and the issuance of NUREG-1150, " Severe l 39 Accident Risks: An Assessment for Five U.S. Nuclear Power Plants," in December
! 40 1990 have greatly increased awareness, knowledge, and concerns in this area.
41 42 The major impetus for the proposed rule is increased interest in new nuclear 43 power generation and the possibility that applicants will request site approval 44 for new nuclear power plants. The Commission believes that, in the event such i 45 requests materialize, the criteria for siting power reactors should directly l
46 address.those site factors important to risk and should reflect the significant 47 . experience learned since the regulation was first issued in 1962.
48 49 Seismic Sitina and Earthauake Enaineerina Criteria
- 50 l 51 The experience gained in the application of the procedures and methods set forth i 52' in the current regulation and the rapid advancement in the earth sciences have i 53 made it necessary to update the 1973 criteria.
! 54 55 EA - 2
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1 2
Environmental Imoacts of the Proposed Action 3
Reactor Sitina Criteria (Nonseismic) 4 5 ,
6 Subpart B to Part 100 contains the considerations that will guide the Commission :
7 in its evaluation of the suitability of a proposed site for nuclear power plants 8
after the effective date of the final regulation. The revision to Part 50 will 9 contain plant the engineering considerations for evaluation of the suitability of the design.
10 The amendment to 10 CFR Part 100 would reflect current licensing 11 practice and would not change the radiological environmental impact. Stated 12 differently, the proposed regulatory actions for future siting applications (10 13 CFR Part 100, Subpart B) are specifically based on maintaining the present level 14 of. risk of A) they radiological releases as in the regulation (10 CFR Part 100, Subpart replace.
15 l 16 17 Seismic Sitina and Earthouake Enaineerina Criteria l
l 18 19 Proposed Appendix B to Part 100 contains the seismic and geologic considerations 20 that would guide the Commission in its evaluation of the suitability of sites proposed for nuclear power plants. Proposed Appendix S to Part 50 contains the 21 22 earthquake engineering considerations that would guide the Commission in its l 23-evaluation of the suitability of the plant design bases. The revision of
! 24 Appendix A to 10 CFR Part 100 as stated in Appendices B and S reflect current !
! 25 licensing practice in earthquake engineering and enhanced current staff practice 26 in seismic and geologic siting through the use of probabilisitic evaluations, i
27 Therefore, the radiological environmental impact offsite will not change. Stated !
28 differently, the proposed regulatory actions (Appendix B to Part 100 and Appendix !
29 S to Part 50) are specifically based on maintaining the present level of risk of '
i 30 radiological releases, thus having zero effect compared to the regulation (Appendix A to Part 100) they replace.
31 32 Onsite occupational radiation exposure associated with inspection and maintenance 33 will not change. These activities are principally associated with baseline 34 inspections of structures, equipment, and piping and maintenance of seismic 35 -instrumentation. Baseline inspections are needed to differentiate between pre-36 37 existing conditions at the nuclear power plant and earthquake-related damage. i 38 The structures, equipment, and piping selected for these insputions are those routinely examined by plant operators during normal plant walkdowns and 39 inspections.
40 Routine maintenance of seismic instrumentation ensures its operability during earthquakes. The location of the seismic instrumentation is 41 42 similar to that in the existing nuclear power plants. In addition, the proposed 43 regulatory guide pertaining to seismic instrumentation (Second Proposed Revision
{ to Regulatory Guide 1.12, " Nuclear Power Plant Instrumentation for Earthquakes")
44 specifically cites i
occupational radiation exposure as a consideration in l
45 selecting the location of the instruments.
46 47 The proposed amendments do not affect non-radiological plant effluents and have 48 no other environmental impact. Therefore, the Commission concludes that there 49 are no significant non-radiological environmental impacts associated with the 50 proposed amendments to the regulations.
51 52 4
53 Alternatives to the Proposed Action
! 54 l 55 As required by Section 102(2)(E) of NEPA (42 U.S.C.A. 4332(2)(E)), the staff has EA - 3
O e 1 considered possible alternatives to the proposed action.
2 3 The first alternative considered by the Commission was to avoid initiating a 4 rulemaking proceeding. This is not an acceptable alternative. Present accident 5 source terms and dose calculations presently influence plant design requirements 6 rather than siting.. It is considered desirable to be able to state directly 7 those siting criteria which, through importance to risk, have been shown to be 8 key to assuring public health and safety. Further, significant advances in the 9 earth sciences and in earthquake engineering, that deserve to be reflected in the 10 regulations, have taken place since the promulgation of the present regulation.
11
- 12. A second alternative considered was deletion of the existing regulation. This 13 is not an acceptable. alternative because these provisions form the licensing 14 bases for many of the operating nuclear power plants and others that are in 15 various stages of obtaining their operating license.
16 17 For the seismic siting and earthquake engineering areas, another alternative 18 considered was replacement of the entire regulation with a regulatory guide.
19 This is not acceptable because a regulatory guide is non-mandatory. The staff.
20 believes that there could be an increase in the risk of radiation exposure to the 21 public if the siting and earthquake engineering criteria were nonmandatory.
22 )
23 1 24 The approach of establishing new sections of the regulations for revised 25 requirements while retaining the existing regulations was chosen as the best 26 alternative. The public will benefit from a clearer, more uniform and consistent 27 licensing process suoject to fewer interpretations. The NRC staff will benefit ,
28 from improved implementation (both technical and legal)of the regulations, fewer 29 interpretive debates, and increased regulatory flexibility. Applicants will 1 30 derive the same benefits in addition to avoiding licensing delays caused by '
31 unclear regulatory requirements. Adopting revised siting and engineering 32 criteria would increase the efficiency of regulatory actions associated with any 33 resurgence of licensing activity.
34 i 35 Alternative Use of Resources 36 37 No alternative use of resources was considered. :
38 39 40 Aaencies and Persons Consulted 41 42 Reactor Sitina Criteria (Nonseismic) 43 44 The NRC staff developed the enclosed rulemaking recommendations. No outside 45 agencies or consultants were used in developing this rulemaking package. ;
46 However, several public meetings were held to inform industry of the staff's '
47 efforts in revising the siting criteria.
48 ;
49 Seismic Sitina and Earthauake Enaineerina Criteria 50 During the development of the proposed regulations and supporting regulatory 51 52 guides, the NRC staff had four public meetings with interested industry groups, '
53 principally, the Nuclear Management and Resources Council (NUMARC) and the 54 Electric Power Research Institute (EPRI). The NRC staff also obtained advice !
55 from the NRC Advisory Committee on Reactor Safeguards and comments from the U.S.
EA - 4
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1 Geological Survey (USGS) staff. As a proposed rule, the regulations will be 2 released for public comment to encourage participation from the public and 3 various organizations in the development of the regulations.
4 5 Findina of No Sionificant Imoact 6
7 The Commission has determined under the National Environmental Policy Act of I 8 1969, as amended, that the proposed amendments to 10 CFR Parts 50 and 100 that j 9 would relocate dose calculation requirements, specify siting criteria '
10 (population, seismic, and geologic), and specify earthquake engineering criteria '
11 for nuclear power plants, if adopted, would not have a significant effect on the 12 quality of the human environment and that an environmental impact statement is 13 not required. '
14 '
15 This determination is based on the following: :
16 17 1. The proposed amendments to the regulations reflect current practice,
- 18 consistent with .the staff's evaluation of applicant's safety analysis 19 _ reports at the time of docketing, applicant's responses to staff initiated 20 ' questions, and the results of.research in the earth sciences and seismic 21 engineering.
.22 23 2. -The foregoing environmental assessment.
24 25 3. The qualitative, deterministic, and probabilistic assessments pertaining ;
26- to seismic events in NUREG--1070, NUREG--1233, and NUREG--1407, 27 -
28 4. The Policy Statement on Severe Reactor Accidents Regarding Future Designs 29 and Existing Plants, published August 8,1985 (50 FR 32138 30 Commission's belief that a new design for a nuclear pow), affirming er plant can the be 31 shown to be acceptable for severe accident concerns if the criteria and :
32 procedural requirements cited in 50 FR 32138 are met.
33 '
34 35 References ;
36 '
37 1. "NRC Policy on Future Reactor Designs, Decisions on Severe Accident Issues in 38 Nuclear Power Plant Regulation," NUREG-1070, July 1985.
39 40 2. " Regulatory Analysis for USI A-40, " Seismic Design Criteria" Final Report,"
41 NUREG-1233, September 1989.
42 43 44
- 3. " Procedural and Submittal Guidance for the Individual Plant Examination of <
External Events (IPEEE) for Severe Accident Vulnerabilities, Final Report," l 45 NUREG-1407, June 1991.
46 i
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ENCLOSURE 4 I
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[7590-1]
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'4 NUCLEAR REGULATORY COMMISSION 1
5 6 Documents Containing Reporting or Recordkeeping Requirements: Office of 7 Management and Budget (0MB) Review 8 '
9 AGENCY: Nuclear Regulatory Commission.
10 11 ACTION: Notice of the Office of Management and Budget review of information 12 collection.
13 14
SUMMARY
- The Nuclear Regulatory Commission (NRC) has recently submitted to 15 the Office of Management and Budget (OMB) for review the following 16 proposal. for the collection of information under the provisions of 17 the Paperwork Reduction Act (44 U.S.C. Chapter 35). There are no l 18 new or revised reporting requirements associated with the proposed 19 regulation 10 CFR Part 100, " Reactor Site Criteria," and 10 CFR Part 20 50, " Domestic Licensing of Production and Utilization Facilities."
21 22' 1. Type of submission, new, revision or extension: Revision 23 i
24 2. The title of the information collections:
25 i 26 Proposed Appendix B, " Criteria for the Seismic and Geologic
?J Siting of Nuclear Power Plants After [ EFFECTIVE DATE OF THIS OMB FRN - 1
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REGULA(ION]," to 10 CFR Part 100, and Proposed Appendix S, 2 " Earthquake Engineering Criteria for Nuclear Power Plants," to 3 10 CFR Part 50. (Revision of Appendix A, " Seismic and 4 Geologic Siting Criteria for Nuclear' Power Plants," to 10 CFR 5 Part 100.)
6 7 3. The form number if applicable: Not applicable 8
9 4. How often the collection is required:
10 11 As necessary in order for NRC to assess the adequacy of 12 proposed seismic design bases and the design bases for other 13 geological hazards for nuclear power plants constructed and 14 licensed in accordance with 10 CFR Part 50 and the Atomic 15 Energy Act of 1954, as amended (the Act).
16 17 5. Who will be required or asked to report: Applicants for a 18 construction permit, operating license, early site permit, 19 design certification, or combined license for nuclear power 20 plants.
21 22 6. An estimate of the number of responses:
23 24 1 annually.
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l l 1 7. An estimate of the number of hours needed annually to complete l 2 the requirement or request:
l 3 164,500.
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5 8. An indication of whether Section 3504(h), Pub. L.96-511 6 applies: Not applicable.
7 l 8 9. Abstract:
9 10 Proposed Appendix B to 10 CFR Part 100 contains criteria '
l 11 associated with the selection of the nuclear power plant site 12 and the establishment of the safe shutdown earthquake ground 13 motion. Proposed Appendix S to 10 CFR Part 50 contains 14 earthquake engineering criteria for nuclear power plants. In 15 combination, these appendices will replace the criteria 16 contained in Appendix A to 10 CFR Part 100, i
17 18 Copies of the submittal may be inspected or obtained for a fee from the NRC 19 Public Document Room, 2120 L Street, NW (Lower Level), Washington, DC.
20 21 Comments and questions can be directed by mail to the OMB reviewer:
l 22 l
23 Ronald Minsk i
24 Office of Information and Regulatory Affairs (3150-0014) 25 NE0B-3019 26 Office of Management and Budget 27 Washington, DC 20503 OMB FRN - 3
4 4 1 Comments can also be submitted by telephone at (202) 395-3084. -
2 3 The NRC Clearance Officer is Brenda Jo Shelton, (301) 492-8132.
.4 5 Dated at Bethesda, Maryland this day of 1991 6
7 For the Nuclear Regulatory Commission 8
9 10 Gerald F. Cranford, 11 Designated Senior Official 12 for Information Resources Management.
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OMB SUPPORTING STATEMENT FOR- -
2 3
4 PROPOSED APPENDIX B, CRITERIA FOR THE SEISMIC AND GEOLOGIC SITING !
5 0F NUCLEAR POWER PLANTS AFTER [ EFFECTIVE DATE], TO 10 CFR PART 6' 100; !
t 7 1 3 AND '
9 i 10 PROPOSED APPENDIX S, EARTHQUAKE ENGINEERING CRITERIA FOR NUCLEAR 3 11 POWER PLANTS, TO 10 CFR PART 50 12 :
13 !
(REVISION OF APPENDIX A TO 10 CFR PART 100) 14 15 16 17 Description of'the Information Collection 18 '
19 ,
20 Seismic and Earthauake Enaineerina Criteria:
21 22 Proposed Appendix B, " Criteria for the Seismic and Geologic Siting of Nuclear i 23 Power Plants After [ EFFECTIVE DATE OF THIS REGULATION]," (Criterion II, IV, and 24 V) to 10 CFR Part 100, " Reactor Site Criteria," requires applicants to provide 25 the types of- information that show evidence of the size and frequency of 26 occurrence of earthquakes, tectonic and nontectonic surface deformation, and 27 seismically induced floods and water waves. Both deterministic and probabilistic .
28 evaluations of earthquake-related phenomena are required. From these seismic and 29 geologic hazard data, applicants determine earthquake ground motion for the 30 seismic design basis, design bases for seismically induced floods and water 31 waves, the need to design for surface deformation, and other design conditions' '
32 that may be affected by earthquake ground motion, such as soil and slope 33 stability. !
34 35 Proposed Appendix S, " Earthquake Engineering Criteria for Nuclear Power Plants,"
36 (Criterion 11 and IV) to 10 CFR Part 50, " Domestic Licensing of Production and 37 Utilization Facilities," requires applicants to provide the design bases for a 38 nuclear power plant that will ensure that structures, systems, and components 39 important to safety will be able to withstand the natural phenomena specified in 40 General Design Criterion 2 of Appendix A to 10 CFR Part 50 and Proposed Appendix 41 B to 10 CFR Part 100 without loss of capability to perform their safety 42 functions.
43 44 Proposed Appendix B to 10 CFR Part 100 and Proposed Appendix S to 10 CFR Part 50, i 45 in combination, are a revision of Appendix A, " Seismic and Geologic Siting l 46 Criteria for Nuclear Power Plants," to 10 CFR Part 100. The proposed appendices 47 apply to applicants who apply for an early site permit, design certification, or 48 combined license pursuant to 10 CFR Part 52, or a construction permit or 49 operating license pursuant to 10 CFR Part 50 on or after [ EFFECTIVE DATE OF THIS 50 REGULATION). However, if the construction permit was issued prior to [ EFFECTIVE 51 DATE OF THIS REGULATION], the operating license applicant must comply with the l 52 seismic and geologic siting and earthquake engineering criteria in Appendix A to l 53 10 CFR Part 100. Appendix A to 10 CFR Part 100 will continue to serve as the i 54 criteria for the seismic and geologic siting and earthquake engineering for I
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l 1 plants licensed or granted their construction permit before [ EFFECTIVE DATE OF l
- 2 THIS REGULATION).
' l 3 1 4 It is anticipated that new plant applications could be submitted within a few 'I
, 5 years. This is based on the current and projected staff review of advanced )
! 6 reactor seismic design criteria related to 'the design certification of two i l 7 evolutionary light-water reactor designs (the Advanced Boiling Water Reactor )
, 8 (ABWR) and the System 80+ Pressurized Water Reactor)- and the Electric Power
( 9 Research Institute (EPRI) Advanced Light Water Reactor Requirements Document. !
10 Based on NRC staff experience obtained from construction permit and operating ;
11 license applications relative to Appendix A to 10 CFR Part 100, the review !
12 process for a construction permit, operating license, early site permit, design j 13 certification, or combined license, as it applies to Proposed Appendix B to 10 ;
14 CFR Part 100 and Proposed Appendix S to 10 CFR Part 50, is expected to range from !
15 one to several years. The NRC staff reviews the Safety Analysis Report for six }
16 to twenty four months and, if necessary, generates a request for additional j 17 information. The applicant usually responds within 1 to 6 months, depending on ,
18 the complexity of the issues. The average time is about 3 months. The responses !
19 are reviewed and a draft Safety Evaluation Report is written by the NRC staff. !
L 20 .This document summarizes conclusions and highlights any outstanding issues. The !
21 staff arranges for a meeting and site visit to resolve any open issues. When the i 22 open issues have been resolved, the staff writes the final Safety Evaluation 23 Report, which is published and used as a basis for the remainder of the NRC ;
1 24 licensing process (the meeting with the Advisory Committee on Reactor Safeguards !
25 (ACRS) and hearing, as necessary, before the Atomic Safety and Licensing Board) i 26 which usually takes about 1% years. .
27 l 28 A. ESTIFICATION i i 29 l
30 1. Need for the Collection of Information 31 '
32 The information required will be needed by the NRC to assess the adequacy l .33 of proposed seismic design bases (siting and engineering) and the design 34 bases for other geological hazards for nuclear power plants in support of ,
35 the agency's mission regarding adequate protection of the health and 36 safety of the public from seismic events. It is to be submitted to the ,
37 NRC. as part of the application and supporting documentation for a construction permit, operating license, 38 early site permit, design 39 certification, or combined license for a nuclear power plant. !
40 '
I 41 Moreover, Proposed Appendix B to 10 CFR Part 100 and Proposed Appendix S 42 to Part 50, supplemented by the Standard Format, regulatory guides, and :
43 the Standard Review Plan, are used by applicants as general guidance in '
44 planning investigations of nuclear power plant sites and designing nuclear ;
45 power plant structures, systems, and components important to safety to 46 withstand the effects of natural phenomena such as earthquakes.
47
, 48 2. Aaency Use of Information i l 49
[ 50 The NRC reviews the geological and seismological information to determine [
51 the suitability of the proposed site for a nuclear power plant and the 52 suitability of the plant design bases established on the proposed site.
i 53 A construction permit, early site permit, standard design certification, i OMB - 2 i i
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l 2 or combined and approved bylicense the NRC.cannot be issued until these data have been reviewe 3
4 5 New geological and seismological information that becomes known during the 6 operating life of a plant is also evaluated on the basis of these criteria.,
7 The criteria also serve as the basis for ongoing NRC research 8
in the earth sciences.
9 3.
10 Reduction of Burden Throuch Information Technoloov 11 12 There are no legal obstacles to reducing the burden associated with this l
13 collection through information technology. Moreover, NRC encourages the use of such technology. l 14 l 15 4. Effort to Identify Duolication '
l 16 17 18 This information does not duplicate other information being provided to NRC.
1 19 20 5. Effort to Use Similar Information 21 22 23 All pertinent geological and seismological information concerning the 24 nuclear site and the region around the site will be used in the analysis 25 of that site, whether it is supplied by the applicant or not. Similarly, ;
26 any available engineering and design data will be used, as applicable, in '
27 the design review of a proposed nuclear power plant whe*her it is a 28 product of the criteria requirements or not. The , <i'6bility of 29 geological, seismological, or engineering data may reduce tne applicant's 30 efforts related to site investigation or design.
31 6. Effort to Reduce Small Business Burden 32 33 34 This information collection does not affect small businesses.
35 7. Consecuences of Less Frecuent Collection 36 37 38 Less frequent collection of information will result in serious delays in 39 the licensing processes of nuclear power plants or potential additional risks to the health and safety of the public.
40 41 8. Circumstances Which Justify Variation From OMB Guidelines 42 43 There is no variation from the guidelines.
44 45 9. Consultations Outside the NRC 46 47 During the development of the proposed regulation, the NRC staff had four 48 public meetings with interested industry groups (principally, the Nuclear i
49 Management and Resources Council (NUMARC) and the Electric Power Research 50 Institute (EPRI)) related to the seismic and earthquake engineering con-51 siderations. With respect to the seismic and geological proposed 52 regulations, the NRC staff also obtained comments from the U.S. Geological 53 Survey (USGS) staff during the development of the proposed regulations.
54 As a proposed rule, the regulations will be released for public comment to i
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< a 1 encourage participation from the public and other organizations in the 2 development of the regulations.
3 4 10. Confidentielity of Information 5
6 Proprietary information is protected in accordance with the provisions 7 specified in 10 CFR Part 2 of the NRC's regulations.
8 9 11. Justification for Sensitive Ouestions 10 11 These regulations do not require sensitive information.
12 13 12. Estimated Annual Cost to the Federal Government 14 15 Current NRC staff activities that are applicable to Proposed Appendix S to 16 10 CFR Part 50 relate to standard design certification. Specifically, the 17 NRC staff is reviewing the design certification of two evolutionary light-18 water reactor designs (the Advanced Boiling Water Reactor (ABWR) and the 19 System 80+ Pressurized Water Reactor) and the Electric Power Research 20 Institute (EPRI) Advanced Light Water Reactor Requirements Document.
21 There are no site-specific construction permit, operating license, early 22 site permit, or combined license application evaluations that relate to l 23 Proposed Appendix B to 10 CFR Part 100 or Proposed Appendix S to 10 CFR 24 Part 50 being performed by the NRC staff.
25 26 Since activities related to Proposed Appendix B to 10 CFR Part 100 and 27 Proposed Appendix S to 10 CFR Part 50 are limited, the following estimates 28 also include NRC staff experience obtained from construction permit or 29 operating license application evaluations relative to Appendix A to 10 CFR l 30 Part 100.
l 31 32 a. Seismic and Geologic Evaluations 33 34 Seismic and geologic staff evaluations required for a construction 35 permit, operating license, early site permit, or combined license 36 review can range from about 1,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> for a site with 37 uncomplicated geology in a region of low seismicity to as many as 38 6,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> for very complex sites. The estimated average annual 39 effort required to review the seismology and geology of an 40 application is about 2,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> or $230,000 ($115 x 2,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />).
41 42 43 b. Earthquake Engineering Evaluations 44 45 Staff evaluations of nuclear power plant structures, systems, and 46 components, to ensure that they will perform their safety function j 47 without loss of capability, average 60,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> per plant. The
- 48 estimated annual staff burden is 12,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> per application. The staff review consists of an evaluation of several loads, one of them 49 50 being the seismic. event. Typical loadings that are considered in 51 the design and staff evaluation of the structures, systems, and 52 components include
- dead load (equipment or building weight), live 53 load (movable equipment load), earthquake, thermal effects, and 54 pressure. It is estimated that twenty five percent of the staff 55 evaluation is devoted to seismic-related issues. Therefore, the OMB - 4 l
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I annual seismic-related portion of the staff review is approxim'ately 2 3,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (25 percent of 12,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />) or $345,000 ($115 x 3,000 3 hours). !
4 5 c.. Consultants 6
7 Consultants and staff.from the U.S. Geologic Survey and Department !
8 of Energy Laboratories are employed by the NRC on a case-by-case ,
9 basis to provide advice in activities ^related to staff reviews !
10 performed in accordance with Proposed Appendix 8 to 10 CFR Part 100 :
11 or Proposed Appendix S to 10 CFR Part 50. It is anticipated that an l 12 average annual effort for these consultants would not exceed 500 t 13 hours or $57,500 ($115 x 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />). l 14 i 15 Total annual cost to the Federal Government for activities related to the l 16 proposed regulation is estimated to be $632,500 ($115 x 5,500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />). i 17-18 13. Estimate of Industry Burden ;
19 20 The estimated seismic and geological revisions burdens are as follows.
21
'22 a. Seismic and Geologic Evaluations 23 24 This estimate is based on the requirement for gathering, analyzing, 25 and synthesizing data. In order for applicants to provide the types
.of information that show evidence of the size and frequency of the 26 f 27 occurrence of earthquakes, the last time there was displacement 28 along faults at the site or in the region, or the potential for ,
29 fault offset during the life of a nuclear power plant, extensive !
30 research and analysis must be conducted. This effort involves the 31 analysis of voluminous amounts of drawings, logs, maps, seismic and 32 other geophysical records, and reports. It is estimated that the 33 industry burden will be, on the average, 24,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> per applicant.
34 The estimated annual burden is 8,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br />s- per applicant or 35 $920,000 ($115 x 8,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />).
36 37 !
38 ;
39 !
40 b. Earthquake Engineering Evaluations 41 ;
42 This estimate is based on the requirement that nuclear power plant j 43 structures, systems, and components important to safety are designed l 44 to withstand the effects of earthquakes without loss of capability 45 to perform their safety functions. In order for applicants to '
46 provide information that shows the functionality of structures, !
47 systems, and components to vibratory ground motion, suitable j 48 analysis, testing, or qualification methods are employed. i 49 50 References 1 and 2 were used to obtain an estimate of seismic- l 51 related costs in nuclear power plant design and construction. The
, 52 incremental cost estimate provided in Table 1 is based on Table 1 of i 53 Reference 1, modified as follows: (1) updated to January 1,1992, 54 costs, (2) increased the Safe Shutdown Earthquake Ground Motion OMB - 5
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1 from 0.2g to 0.3g, and (3) increased distribution system and 2 engineering costs.
3 4 It is estimated that the industry burden associated with the seismic 5- engineering (staff-related costs) of nuclear power plant structures, 6
systems, and components will average $88,850,000 per application.
7 The estimated annual burden per application will average $18,000,000 l 8 or approximately 156,500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> ($115 x 156,500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> approximately i 9 equals $18,000,000). This cost estimate may be reduced by 10 additional savings associated with standardized plant designs, and 11 more significantly, by elimination of analyses and design associated l 12 with the Operating Basis Earthquake Ground Motion (OBE) as stated in 13 Proposed Appendix S to 10 CFR Part 50.
14 15 The total annual burden on industry for activities related to the proposed 16 l
' regulations is estimated to be $18,920,000 ($115 x 164,500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />).
17 18 14. Reasons for Chance in Burden 19 '
i 20 The estimated burden on the NRC staff and industry remains the same. For 21 applicants of a construction permit, operating license, early site permit, 22 design certification, or combined license, no significant increases in 23 costs are envisioned to implement the revised regulations. In general,
- 24 the proposed revisions reflect current staff practice. Specifically, in 3 l 25 the area of geologic and seismic siting, the required probabilisitic.
26 evaluations are new but should not have a significant cost impact. Some 27 probabilistic evaluations have been used in recent licensing reviews to 28 determine the probability of exceeding the safe shutdown earthquake ground l
29 motion at the plant site. With regard to earthquake engineering, the
! 30 proposed regulation reflects or possibly reduces current staff practice.
l 31 In addition, the proposed revisions to the regulations will. reduce delays l 32 in the licensing process because information needed for the staff review 33 can be ' incorporated in the safety analysis reports at the time of 34- docketing instead of later through. staff questions and applicant 35 responses.
! 36 37 15. Publication for Statistical Use l 38 l 39 This information is not collected for statistical purposes.
40 41 42 B. COLLECTION OF INFORMATION EMPLOYING STATISTICAL METHODS 43 44 Appendix B of 10 CFR Part 100 allows for the acquisition of statistical
-45 data and the use of statistical methods, but does not require them.
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'47 48 i 49-i 50 References
! 51 52 1. NUREG/CR-1508, " Evaluation of the Cost Effects on Nuclear Power i
53 Plant Construction Resulting from the Increase in Seismic Design l 54 Level," April 1981.
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e o 1 2. Stevenson and Associates, " Differential Design and Construction Cost 2 of a Nuclear Power Plant Safety Related Piping Systems as a Function 3 of Seismic Intensity and Time Period of ' Construction for New and 4 '- Operating Plants and Current Simplified Seismic Design Initiatives,"
5 Draft, July 1990.
6 7
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Enclosures:
9 Table l', Summary of Incremental Cost Estimate 10 Table 2, OMB Supporting Statement 11 I
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2 TABLE 1 3
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SUMMARY
OF INCREMENTAL COST ESTIMATE 5
6 0.3G Safe Shutdown Earthquake Ground Motion vs 7
No Seismic Design Requirement 8 ITEM COST ESTIMATE
- 9 Foundations 5 35,425,000 10 Structures 3,675,000 11 Auxiliary Components 16,375,000 12 NSSS Components 4,425,000 13 Distribution Systems 114,875,000 14 Engineering 88,850,000 15 Turbine Hall 525,000 16 Total Cost Estimate 5 264,150,000
- 17 18 19 20 21 22 '
23 Based on Table 1 in Reference 1, modified as follows:
24 a. Updated to January 1,1992 costs. A factor of 2.2, based on an 25 26 inflation and escalation rate of 8.0 percent between January 1977 27 and 1985, and 5.0 percent between January 1985 and 1992 (from Table 28 7.2 of Reference 2) was used.
29 b.
Increased Safe Shutdown Earthquake Ground Motion from 0.2 30 9 to 0.39 .
31 Based used.
on Figures 1 and 2 of Reference 1, a cost factor of 2 was 32 33 c.
34 Increased Distribution System and Engineering costs. In addition to 35 increasing these costs based on Steps a and b, new piping costs 36 based and craft oncosts: Tables 5.10 and 5.11 of Reference 2 were used. (Material 37 5174,882,470 with seismic design and restraints, 38 $67,177,570 without seismic design and restraints. Engineering 39 costs: 563,984,090 with seismic design and restraints, 56,344,920 40 without seismic design and restraints.)
41 42 43 Ths total cost estimate does not reflect potential savings associated with 44 the use of a standardized plant designs or elimination of analyses and 45 design associated with the proposed rulemaking. Therefore, the cost estimate may be reduced.
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4 5 TABLE 2 6 DMB SUPPORTING STATEMENT 7
8 Appendix B to 10 CFR Part 100 and Appendix S to 10 CFR Part 50 9 (Revision of 10 CFR Part 100, Appendix A) 10 :
11 12 TASK HOURS OR DOLLARS
'13 14 15 ESTIMATED AVERAGE ANNUAL 164,500 16 BURDEN HOURS PER RESPONSE :
17 18 19 NUMBER OF RESPONDENTS ANNUALLY I 20 21 22 ESTIMATED TOTAL ANNUAL BURDEN 164,500 l 23 HOURS ,
24 25 26 ESTIMATED TOTAL ANNUAL COST TO $18,917,500 27 INDUSTRY 28 29' ,
30 ESTIMATED TOTAL ANNUAL STAFF 5,000 31 HOURS 32 33 !
34 ESTIMATED NRC CONSULTANT HOURS 500 35 _
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37 ESTIMATED ANNUAL COST TO THE 38 -FEDERAL GOVERNMENT 5,500 39 (STAFF + CONSULTANT HOURS) 40 41 42 43 OMB - 9
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[7590 - 01] -
f NUCLEAR REGULATORY COMMISSION i
AGENCY: Nuclear Regulatory Comission. I 9
ACTION: Notice of Availability of Draft Regulatory Guides and Standard g Review Plan Section. !
SUMMARY
- The Nuclear Regulatory Comission is proposing to issue or amend r
several Regulatory- Guides and one section of the Standard Review Plan in connection with the proposed revision of 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 i proposed guides and standard review plan revision provides prospective licensees '
with the necessary guidance for implementing the proposed revision of 10 CFR Part i 100 " Reactor Site Criteria." The notice . of availability of the proposed regulation is published'in the notices section of this Federal Register.
DATE: Coment period expires 60 days after date of publication in the Federal '
Register. Coments received after this date will be considered if it is practical to do so, but the Comission is able to assure consideration only for comments received on or before this date.
l- ADDRESSES: Mail written coments to: ' Secretary, U.S. Nuclear Regulatory l l
Comission, Washington, DC 20555, Attention: Docketing and Service Branch. >
I Deliver comments to 11555 Rockville Pike, Rockville, Maryland, between 7:45 am and 4:15 pm Federal workdays.
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t a Copies of the draft regulatory guides, standard review plan section,' and coments received may be examined at the NRC Public Document Room at 2120 L Street NW. (Lower Level),-Washington, DC.
FOR FURTHER INFPWTION CONTACT: Dr. Andrew J. Murphy, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Comission, Washington, DC 20555, telephone (301) 492-3860, concerning the seismic and earthquake engineering aspects and Mr. Leonard Soffer, Office of Nuclear Regulatory Research, U.S.
Nuclear Regulatory Comission, Washington, DC 20555, telephone (301) 492-3916, concerning other siting aspects.
(
SUPPLEMENTARY INFORMATION:
The NRC is developing the following draft regulatory guides and standard review plan section to provide prospective licensees with the necessary guidance :
for implementing the proposed revision of 10 CFR Part 100, " Reactor Site Criteria." The notice of availability for this regulation is published in the notices secti0n of this Federal Register. The 'following draft guides are available:
1
- 1. DG-1015, " Identification and Characterization of Seismic Sources, !
Deterministic Source Earthquakes, and Ground Motion." The draft guide provides general guidance and recomendations, descr'ibes acceptable procedures and provides a list of references that present acceptable methodologies to identify I 1
and characterize capable tectonic sources and seismogenic sources.
l
- 2. DG-1016, Second Proposed Revision 2 to Regulatory Guide 1.12, " Nuclear Power Plant Instrumentation for Earthquakes." The draft guide describes seismic instrumentation type and location, operability, characteristics, installation, actuation, and maintenance that are acceptable to the NRC staff.
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- 3. DG-1017, " Pre-Earthquake Planning and Immediate Nuclear Power Plant-Operator Post-Earthquake Actions." The draft guide provides guidelines that are acceptable to the NRC. staff for a timely evaluation of the recorded seismic instrumentation data and to determine whether or not plant shutdown is required.
- 4. DG-1018, " Restart of a Nuclear Power Plant Shut Down by a Seismic Event." The draft guide provides guidelines that are acceptable to the NRC staff for performing inspections and tests of nuclear power plant equipment and structures prior to restart of a plant that has been shut down due to a seismic event.
- 5. Draft Standard Review Plan Section 2.5.2, Proposed Revision 3 " Vibratory Ground Motion." The draft describes procedures to assess the ground motion potential of seismic sources at the site and to assess the adequacy of the SSE.
- 6. Draft Regulatory Guide 4.7, Revision 2, dated December 1991, " 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.
Dated at Rockville, Maryland, this _ day of , 1992.
For the Nuclear Regulatory Commission.
Eric S. Beckjord, Director, Office of Nuclear Regulatory Research.
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j J U.S. NUCLEAR REGULATORY CONNISSION Proposed Revision 2 4 March 1992
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REGELATORY GLIDE OFFICE OF NUCLEAR REGULATORY RESEARCE i 1 I I 3
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1 RECULATORY GUIDE 4.7 l CENERAL SITE SUITABILITY CRITERIA FOR NUCLEAR POWER STATIONS 1
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TABLE OF CONTENTS f.121 A. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 S. DISCUSSION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
- 1. Geology / Seismology . . . . . . . . . . , . . . . . . . . . . . 3
- 2. Meteorology . . . . . . . . . . . . . . . . . . . . . . . . . . 4
- 3. Population Considerations . . . . . . . . . . . . . . . . . . . . 6
- 4. Hydrology . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 4.1 Flooding . . . . . . . . . . . . . . . . . . . . . . . . . 6 4.2 Water Availability . . . . . . . . . . . . . . . . . . . 7 4.3 Water Quality . . . . . . . . . . . . . . . . . . . . . 7
- 5. Ecological Systems and Biota . . . . . . . . . . . . . . . . . . 8
- 6. Land Use and Aesthetics . . . . . . . . . . . . . . . . . . . . 10
- 7. Industrial, Military, and Transportation Facilities . . . . . . 11
- 8. Socioeconomics . . . . . . . . . . . . . . . . . . . . . . . . . 12
. 9. Noise . . . . . . . . .,. . . . . . . . . . . . . . . . . . . . 13 C. REGULATORY POSITION . . . . . . . . . . . . . . . . . . . . . . . . . 13
- 1. Geology / Seismology . . , . . . . . . . . . . . . . . . . . . . . 13
- 2. Meteorology . . . . . . . . . . . . . . . . . . . . . . . . . . 13
- 3. Population Consideration . . . . . . . . . . . . . . . . . . . . 14
- 4. Hydrology . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 4.1 Flooding . . . . . . . . . . . ~. . . . . . . . . . . . . 14 4.2 Water Availability . . . . . . . . . . . . . . . . . . . 14 4.3 Water Quality . . . . . . . . . . . . . . . . . . . . . . 14 4.4 71ssion Product Retention and Transport . . . . . . . . . 15
- 5. Ecological Systems and Slota . . . . . . . . . . . . . . . . . . 15 i
- 6. Land Use and Aesthetics . . . . . . . . . . . . . . . . . . . . 16
- 7. I'dostrial. Military, and Transportation Facilities . . . . . . 17
- 4. Socioeconceales . . . . . . . . . . . . . . . . . . . . . . . . . 17
- 9. Noise . . . .. . . . . . . . . . . . . . . . . . . . . . . . . 18
- 10. Emergency Planning . . . . . . . . . . . . . . . . . . . . . . 18 ;
D. IMPLEMENTATION . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 APPENDIX A SAFETY-RELATED SITE CONSIDERATIONS FOR ASSESSING SITE SWITASILITY FOR NUCLEAR POWER STATIONS . . . . . . . . . . . . . . 19 APPENDIX S ENVIRONMENTAL CONSIDERATIONS FOR ASSESSINC SITE SUITABILITY FOR NUCLEAR POWER STATIONS . . . . . . . . . . . . . . . . . .. . . . 25
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o s .
l l
1 A. INTRODUCTION 2
3 The Energy Reorganization Act of 1974 places on the Nuclear Regulatory 4 Commission (NRC) the responsibility for the licensing and regulation of pri-5 vate nuclear facilities from the standpoint of public health and safety.
6 Title 10, CFR Part 100, " Reactor Site Criteria," requires that the population 7 density, use of the site environs including proximity to man-made hazards, and 8 the physical characteristics of the site, including seismology, meteorology, 9 geology, hydrology, be taken into account in determining the acceptability of 10 a site for a nuclear power reactor. Seismic and geologie site criteria for 11 nuclear power plants are provided in Appendix A and Appendix B to 10 CFR Part ,
)
12 100. Appendix A to 10 CFR Part 50 establishes the minimum requirements for 13 tha principal design criteria for water-cooled nuclear power plants; a number 14 of these criteria arm directly related to site characteristics as well as to 15 evonts and conditions outside the nuclear power unit.
16 .
17 The National Environmental Policy Act of 1969 (NEPA) (83 Stat. 852),
18 implemented by Executive Order 11514 and the Council on Environmental Qual-19 ity's Guidelines of August 1, 1973 (38 FR 20550), requires that all agencies 20 of the Federal Government prepare detailed environmental statements on pro-21 posed major Federal actions which can significantly affect the quality of the 22 human environment. A principal objective of NEPA is to require the Federal 23 agenef to consider, in its decision-making process, the environmental impacts 24 of each proposed major action and the available alternative actions, including 25 alternative sites.
26 27 Part 51, " Licensing and Regulatory Policy and Procedures for Environ-28 mental Protection,' of Title 10, Code of Federal Regulations, sets forth the 29 Nuclear Regulatory Commission's policy and procedures for the preparation and 30 processing of environmental impact statements and related documents pursuant 31 to Section 102(2)(C) of the NEPA.
32 33 The limitations on the Commission's authority and responsibility 34 pursuant to the NEPA imposed by the Federal Water Pollution Control Act (86 35 Stat. 916) are addressed in an Interim Policy Statement published in the 36 Federal Reoister on January 29, 1973 (38 FR 2679). 1 37 38 This guide discusses the major site characteristics related to public 39 health and safety and environmental issues which the NRC staff considers in 40 determining the suitability of sites for light-water-cooled (LWR) nuclear 41 power utations.* The guidelines may be used by applicants in identifying 42 suit + 'a candidate sites for nuclear power stations. The decision that a 43 stativ. .2y be built on a specific candidate site is based on a detailed 44 evaluation of the proposed site-plant combination and a cost-benefit analysis 45 comparing it with alternative site-plant combinations as discussed in 46 Regulatory Guide 4.2. " Preparation of Enviror. mental Reports for Nuclear Power 47 stations."
48 49 Chapter 9 of Regulatory Guide 4.2 discusses the selection of a site from 50 among alternative sites. Although it is recognized that planning pethods' l 51
- For the purposes of thin cuide, nuclear power station refers to the 52 nuclear reactor unit (s), neelsar steam supply, electric generating 53 units, auxiliary systems, including the cooling system and l
54 structures such as docks that are located on a given site, and any 55 new electrical transmission towers and linas erected in connection 56 with the facilities.
57 6 site selection methodologies that have been used by the nuclear 58 power industry are described in " Nuclear Power Plant Siting, A 59 Generalized Process," Atomic Industrial Forum, August 1974, National 60 znvironmental Studies Project, R-1578.
4.7-1 1
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e e I will differ among applicants, Chapter 9 states that the applicant should pre-2 sent-its site-plant selection process as the consequence of an analysis of 3 alternatives whose environmental costs and benefits were evaluated and com-4 pared and then weighed against those of the proposed facility.
5 6 This guide is intended to assist applicants in the initial stage of 7 selecting potential sites for a nuclear power station. Each site that appears 8 to be compatible with the general criteria discussed in this guide will have 9 to be examined in greater aetail before it can be considered to be a 'candi-10 date' site, i.e., one of the group of sites that are to be considered in 11 selecting a " proposed' or " preferred' site.'
12 l 13 This guide should be used only in the initial stage of site selection 14 because it does not provide detailed guidance on the various relevant factors 15 and format for ranking the relative suitability or desirability of possible 16 sites. This guide provides a general set of safety and environa. ental criteria 17 which the NRC staff has found to be valuable in assessing candidate site 18 identification in specific licensing cases.
19 l
20 -
z mation needed to evaluate potential sites at this initial stage 21 of ..;. ee..f;t on is assumed to be limited to that information which may be 22
,23 obta'.ned from published reports, public records, public and private agencies, and individuals knowledgeable about the locality of a potential site.
24 Although in some cases the applicants may have conducted on-the-spot inves-25 tigations, it is assumed here that these investigations would be limited to 26 reconnaissance-type surveys at this stage in the site selection process.
27 28 The safety issues discussed include geologic / seismic, hydrologic, and 29 meteorological characteristics of proposed sites: potential effects on a 30 station from accidents associated with nearby industrial, transportation, and 31 military facilities; and population densities in the site environs as they 32 relate to protecting the general public from the potential radiation hazards l 33 of postulated serious accidente. 'The environmental issues discussed concern 34 potential impacts from the construction and operation of nuclear power 35 stations on ecological systems, water use, land use, the atmosphere, 36 aesthetics, and socioeconomics.
37 38 This guide does not discuss details of the engineering designs required 39 l to ensure the compatibility of the nucisar station and the site or the l 40 detailed information required for the preparation of the safety analysis and 41 environmental reports. In addition, nuclear power reactor site suitability as ,
l 42 it may be affected by the commission's materials safeguards and plant pro- i 43 tection requirements for nuclear powec plants is not addressed in this guide.
l 44 :
i 45 cuidance concerning the siting of offshore nuclear stations, high )
46 temperature gas-cooled (HTGR), liquid metal fast breeder reactors (LMFBR),
47 test reactors, and advanced siting concepts such as underground sites and 48 nuclear energy centers is not included in this guide.
49 50 A significant commitment of time and resources may be required to select 51 a suitable site for a nuclear power station, including safety and environ-l 52 mental considerations. site selection involves considerations of public i l 53 health and safety, engineering and design, economies, institutional '
54 requirements, environmental impacts, and other factore. The potential impacts l 55 of the construction and operation of nuclear power stations en the physical l
l 56
- see Chapter 3 of Regulatory Guide 4.2 for a discussion of site
- 57 selection procedures. The " proposed" site submitted by an applicant t
58 for a construction permit is that site of a number of ' candidate
- l 59 sites which the applicant prefers and on which the applicant i 60 proposee te construct a nuclear power station. ;
l 4.7-2
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o o 1 and biological environment and on social, cultural, and economic features
- are 2 usually similar to the potential impacts of any major industrial facility, but 3 nuclear power stations are unique in the degree to which potential impacts of 4 the environment on their safety must be considered. The safety requirements 5 are primary determinants of the suitability of a sit 3 for nuclear power 6 stations, but considerations of environmental impacts and public acceptance of 7 nuclear power stations are also important and need to be evaluated. ,
8 9 In the site selection process, coo-dination between applicants for 10- nuclear power stations and various red.eral, state, and local agencies will be >
11 useful in identifying potential problem areas.
12 13 Appendices A and B of this guide summar*to the important safety-related '
14 and environmental considerations for assessing 6.he site suitability of nuclear 15 power stations.
16 17 B. DIscussIow ;
18
.19 ' 1. Geology /s ismology 20 21 Nuclear power stations must be designed to prevent the loss of safety-22 related functions. Generally, the most restrictive safety-related site char-
.23 acteristics considered in determining the suitability of a site are surface i t
24 faulting, potential ground motion and foundation conditions * (including i 25 liquefaction, subsidence, and landslide potential), and seismically indue d -
26 27 floods. Criteria that describe the nature of the investigations required to i 28 obtain the geologic and seismic data necessary to determine site suitability ;
29 are provided by Appendix B, " Criteria for the seismic and Geological siting of '
l 30 Nuclear Power Plants after (EFFECTIVE DATE]* to 10 CFR Part 100. safety-l related site characteristics are identified in section 2.5 of Regulatory Guide t
31 1.70, ' standard Format and content of safety Analysis Reports for Nuclear 32 Power Plants," and Regulatory Guide 1.59, *3esign Basis Floods for Nuclear 33 Power Plants." In addition to geologic and seismic evaluation for assessing 34 seismically induced flooding potential, section 2.4 of Regulatory Guide 1.70 35 and Regulatory Guide 1.59 describe hydrologic criteria, including coincident '
36 flood events that should be considered.
37 38 '
39
- 40 Biological and physical environment includes geology, geomorphology, ,
surface and groundwater hydrology, climatology, air quality, :
41 42 limnology, water quality, fisheries, wildlife, and vegetation.
43 social and cultural features include scenic resources, recreation resources, archeological / historical resources, and community 44 45 resources including land use patterns. From " Development and the Environment: Legal Reforms to racilitate Industrial site 46 selection," final report by the Committee on Environmental Law, 47 American Bar Association, February 1974.
6 48 " Classification, Engineering Properties and Field Exploration of 49 soils, Intact Rock and In situ Masses," WASH-1301, March of 1974, 50 outlines some of the procedures used to evaluate site foundation 51 properties. '
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t
e e j
}. i
[ l 2. Atmospheric Extremes and Dispersion j 2 3
j l
4
' The potential ef fect of natural atmospheric extremes (e.g. , tornadoes' 5
and exceptional icing conditions *) on the safety-related structures of a i
6 nuclear station must be considered. However, the atmospheric extremes that i
7 may occur at a site are not normally critical'in determining the suitability 8 of a site because safety-related structures, systems, and components can be i designed to withstand most atmospheric extremes. p 9 ;
! 10 11 The atmospheric characteristics at a site are an important consideration 5 12 in evaluating the dispersion of radioactive affluents both from postulated j 13 accidents and from routine releases in gaseous effluents.' In addition to
- 14 meeting the NRC requirements for the dispersion of airborne radioactive i 15 material, the station must meet State and Federal requirements of the clean 16 Air Amendments of 1970 (PL 91-604).- This is unlikely to be an important 4
I 17 consideration for nuclear power station siting unless (1) a site is in an area
! 18 where existing air quality is near or exceeds the limits set under the clean .
19 Air Amendments,.(2) there is a potential for interaction of the cooling system 20 plume with a plume containing noxious or toxic substances from a nearby 1
1 21 facility, or (3) the auxiliary generators are operating.
j 22 23 The atmospheric data necessary for adequate assessment of the potential 4
24
' dispersion of radioactive material from design basis accidents are described
- 25 in Regulatory Guide 1.23, "Onsite Meteorological Programs." Models and 1 26 assumptions used for evaluating the potential radiological consequences of 1 certain postulated accidents are provided in Regulatory Guides 1.3, "Assump- i 1
i 27 a 28 Refer to Regulatory Guide 1.76, " Design Basis Tornado for Nuclear l Power Plants.*
l 29
- 4 j Refer to section 2.4.7 of Regulatory Guld 1.70.
', 30
- 31 Routine releases of airborne radioactive material must be kept
- as low as practicable."
32 [See 10 CFR Part 20, Sec. 20.1(c).) The 33 commission has published a proposed rule for public comment (40 FR 33029)less thatprecise substitutes term"as "aslow lowas asispracticable reasonably achievable" for the t 34 older, 35 where it appears in NRC regulations and regulatory guides.
36 section 50.34a of 10 CFR Part 50 sets forth the requirements 37 for design objectives for equipment to control releases of 5 38 radioactive material in effluents from nuclear power reactors, j 39 section 50.3sa further provides that, in order to keep power 1
40 reactor effluent releases as low as practicable, each license 41 authorizing operation of such a facility will include technical 42 specifications regarding the establishment of effluent control 43 equipment and reporting of actual releases.
44 Appendix I to 10 CFR Part 50, promulgated May 5, 1975 (40 FR 45 19439), provides numerical guidance for design objectives and 46 technical specification requirements for limiting conditions of 47 operation for light-water-cooled nuclear power plants.
48 The following regulatory guides are being prepared to assist 49 {
in application of the numerical guidance in Appendix I 50 1. calculation of Annual Average Doses to Man from Routine l 51 52 Releases of Reactor Effluents for the Purpose of Implementing Appendix I,.
53 l 54
- 2. calculations of Releases of Radioactive Materials in Liquid and Gaseous Effluents from Pressurized Water Reactors (PWRs),
55 3. calculation of Releases of Radioactive Materials in 56 Liquid and Gaseous Effluents from noiling water Reactors (awRs), and !
57 4. Methods for Estimating Atmospheric Dispersion of Gaseous 58 Effluents from Routine Releases. I 4.7-4 1
4
. , . - , . - - . - e-
e e
! l 1
2 tions Used for Evaluating the potential Radiolo 3
of-Coolant Accident for Boiling Water Reactors;gical Consequences 1.4,
- Assumptionsof a Loss-Used for l 4 Evaluating the Potential Radiological dont for Pressurized Water Reactors;' Consequences of a Loss-of-Coolant Acci-5
_6-the Potential soiling WaterRadiological Consequences of a Steam Line Break Accident forl.5, " Assumpt Reactors;' l.24, 7
- Assumptions Used for Evaluating the Potential 8 Radiological age Consequences Tank Failure;' of a Pressurized Water Reactor Radioactive Gas Stor-9 and 1.25, ' Assumptions Used for Evaluating the Potential 10 Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and i 11 storage Facility for soiling and Pressurized Water Reactors." However, the
- 12 atmospheric unusual assumptions atmospheric in the guides may not be appropriate for sites with conditions.
- 13
! 14 15 In the evaluation of potantial sites, onsite atmospheric reconnaissance 16 can determine if the atmospheric sented by the available atmosphericconditions data for atthea area.
site are adequately repre-17 Canyons or deep 18 valleys frequently have atmospheric variables that are substantially different 19 from those variables measured for the. general region. 'other topographical 20 features such as hills, mountain ranges, and lake or ocean shornlines can 21 affect the local atmospheric conditions at a site and may cause the dispersion !
22 characteristics area or region. at the site to be less favorable than thoso in the general 23 in such cases. More stringent design or effluent objectives may be required l
l 24 i 25 While it is thethe concentration of release, of radioactivity in the atmosphere at any 26 distance from point x(Ci/m ), that must l 27 3
be controlled, the 28 ratio x/Q, where g(ci/see) is the rate of release of radioactivity from the 29 source, has become a commonly evaluated term because it depends only on 30 atmospheric variables and distance from the source. :
l 31 32 33 cuidesIf1.3under assumed unfavorable atmospheric conditions (see Regulatory and 1.4) the dispersion of radioactivity released following a design basis accident is insufficient at the boundary of the exclusion area 34 35 (see the following section, ' Population Considerations") and the outer i 36 boundary of the low population zone, the plant design would not satisfy the 37 requirements of 10 CFR Part 50.34(a)(1). Thus, the design of the station 38 would be required to include appropriate and adequate compensating engineered safety features.
39 In addition, meteorological conditions are to be determined 40 for use in the environmental report required in 10 CFR Part 51 and for comparison to the meteorology assumed in the Probabilistic Risk Assessment 41 I (PRA) for a certified plant design (if such a design is to located at the 42 site) or used in the site specific PRA for a custom plant at the site. !
I 43 1
44 45 Local fogging and icing can result from plumes discharged into the i
i 46 atmosphere from cooling towers, lakes, canals, or spray ponds, but can gen-47 erally be acceptably mitigated by station design and operational practices.
48 However, some sites have the potential for severe fogging or icing due to local atmospheric conditions.
I 49 For example, areas of unusually high moisture 50 content that are protected from large-scale airflow patterns are most likely to experience these conditions. The impacts are generally of greatest poten-51 52 tial importance relative to transportation or electrical transmission corridors in the vicinity of a site.
53 54 55 A cooling system designed with special consideration for reducing drift may be required due to the sensitivity of the natural vegetation or the crops 56 57 in the vicinity of the site to damage from airborne salt particles. The vul-nerability of existing industries or other facilities in the vicinity of the t
l 58 59 site to corrosion by drift from cooling tower or spray system drift should be considered. Not only are the amount, direction, and distance of the drift 60 j
! from the cooling system important, but the salt concentration above the t
- 61 natural background salt deposition at the site is also important in assessing 1 62 drift effects. None of these considerations are critical in evaluating the 63 ;
i suitability of a site, but they could result in special cooling system design i i
4.7-5 i'
4 9
- 1. re.quirements or in the need for a larger site to confine the effects of drift 2 w! thin the site boundary. The environmental effects of salt drift are most 3 severe where saline water or water with high mineral content is used for ,
4 cordenser cooling. ,
5 t 6~ . Cooling towers produce cloudlike plumes which vary in size and altitude 7 depending on the atmospheric conditions. The plumes are often a few miles in 8 length before becoming dissipated, but the plumes themselves or their shadows A could have aesthetic' impacts. Visible plumes emitted from cooling towers in ;
10 the vicinity of airports could cause a hazard to aviation.
11 '
12 3. repulation considerations 13 .
l 14 A reactor licensee is required by 10 CFR Part 100 to designate an .
15 exclusion area and to have authority to determine all activities within that !
16 area,. including removal of personnel and property. In selecting a site for a i 17 nuclear power station, it is necessary to provide for an exclusion area in
- 18 which the applicant has such authority. Transportation corridors, such as 19 highways, railroads, and waterways, are permitted to traverse the exclusion +
20 area provided (1) these are not so close to the facility as to interfere with f
21 normal operation of the facility and (2) appropriate and effective arrange- l l 22 monts are made to control traffic on the highway, railroad, or waterway in the 23 case of emergency to protect the public health and safety. !
24 i 25 As set forth in 10 CFR Part 100, nuclear power station sites should be ,
l 26 located in areas with low population density. If the population density of a [
27 proposed site a) exceeds 500 people per square mile averaged over any radial :
! 28 distance out to 30 miles or b) is projected to exceed 1000 people per square
- l 29 mile averaged over any radial distance out to 30 miles 40 years after the time i l
30 of site approval, the applicant should give special attention to alternate !
31 sites. !
'32 l 33 WASH-1235, 'The site Population Factor, A Technique for Consideration of !
34 Population in Site Comparison," October 1974, discusses a methodology that is !
35 useful in comparing population distributions at alternative sites. !
j 36 i 37 4. urdrology I L 38 t 39 4.1 riooding 40 41 Criteria for evaluation of seismically induced floods are provided in 42 Appendix B to 10 CFR part 100. Regulatory Guide 1.59 describes an acceptable j 43 method of determining the design basis floods for sites along streams or 44 rivers and discusses the phenomena producing comparable design basis floods 45 for coastal, estuary, and creat Lakes sites. The ofreets of a probable maxi- ,
46 mum flood (as defined in Regulatory cuide 1.59), seiche, surge, or seismically i
, 47 induced flood such as might be caused by dam failures or tsunami on station ,
48 safety functions can generally be controlled by engineering design or protec- !
49 tion of the safety-related structures, systems, and components which are '
50 identified in Regulatory Guide 1.29, " Seismic Design Classification." For I 51 some river valleys, flood plains, or areas along coastlines, there may not be 52 sufficient information to make the evaluations needed to satisfy the criteria 53 for seismically induced flooding. In such cases, study of the potential for -
54 dam failure, river blockage, or diversion in the river system or distantly and I 55 locally generated sea waves may be needed to determine the suitability of a 56 site. In lieu of detailed investigations, Regulatory Guide 1.59 and Section 57 2.4 of Regulatory cuide 1.70 present acceptable analytical techniques for :
58 evaluating seismically induced flooding. ,
59 2 l
60 4.2 water availability 61 62 muelear power stations require reliable sources of water for steam 63 condensation, service water, emergency core cooling system, and other func-4.7-6
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_ . _ _ _ _ __m_ _ _ _ _ _ - _ .t - - _ _ _ _ - __ ..___u- - _ _ _ _ _ . _ _ _ - - _ _ _ _ _
o .
l 1 tions.
2 In regions where water is in short supply, the recirculation of the 3 hot cooling water through cooling towers, artificial ponds, or impoundments has been practiced.
4 5
6 Essential water requirements for nuclear power plants are that 7 sufficient shutdown, water be available for cooling during plant operation and normal 8 for the ultimate heat sink,' and for fire protection. The 9 limitations imposed by existing laws or allocation policies govern the use and 10 consumption of cooling water at potential sites' for normal operation.
11 Regulatory Guide 1.27 discusses the safety requirements. Consumptive use of 12 water may necessitate an evaluation of existing and future water uses in the 13 area to ensure adequate water supply during droughts both for station operation and other water users 14 (i.e., nuclear power station requirements versus public water supply). Regulatory 15 potential conflicts. agencies should be consulted to avoid 16 17 18 Where required by applicable law, demonstration of a request for 19 certification of the rights to withdraw or consume water and an indication 20 that the request is consistent with appropriate State and regional programa 21 and policies should be provided as part of the application for a construction permit or operating license.
22 23 24 The availability of essential water during periods of low flow or low 25 water level is an important initial consideration for identifying potential 26 sites on rivers, small shallow lakes, or along coastlines. Both the frequency 27 and duration of low flow or low level periods should be determined from the historical record and, 28 29 from projected operatingif the cooling water is to be drawn from impoundments, practices.
30 a.3 Water cuality 31 32 33 Thermal and chemical effluents discharged to navigable streams are 34 governed by the Federal Water Pollution Control Act (FWPCA, PL 92-500), 40 CFR 25 Part 122, 40 CFR Part 423, and State water quality standards. The applicant ;
36 should also determine other regulations that are current at the time sites are under consideration. Section 401(a)(1) of the FWPCA requires, in part, that 37 38 any applicant for an NRC construction permit or combined license (combined 39 construction permit and operating license) for a nuclear power station provide 40 to the NRC certification from the State that any discharge will comply with 41 applicable ments.
effluent limitations and other water pollution control require-42 In the absence of such certification, no construction permit or 43 combined license can be issued by NRC unless the requirement is waived by the 44 state or the State fails to act within a reasonable period of time. A 45 National Pollution Discharge Elimination System (NPDES) permit to discharge 46 effluents to navigable streams pursuant to Section 402 of the FWPCA may be 47 required for a nuclear power station to operate in compliance with the Act, but is not a prerequisite to an NRC construction permit or operating license.
48 49 Evaluations of the dispersion and dilution capabilities and potential 50
- Regulatory cuide 1.27, " Ultimate Heat Sink for Nuclear Power 51 Plants," provides guidance on water supply for the ultimate heat 52 aink.
53
- I 54 To the extent that site selection is dependent on water diversions 55 for consumptive use, allocation of water supply is a function of state statutory and administrative procedures.
56 A discussion of the establishment of state regulation of water 57 use is provided in " Industrial Developments and the Environment, 58 Legal Reforms to Improve the Decision-Making Process in Industrial 59 Site Selection," Special Committee on Environmental Law of the 60 American Bar Association, August 1973.
4.7-7
. m I contamination pathways.of the ground water environment under operating and 2 accident conditions.with respect to present and future users are required. !
3 > Potential radiological and nonradiological contaminants of ground water should .
4 be evaluated. The suitability of sites for a specific plant design in areas !
5 with a complex ground water hydrology or of_ sites located over aquifers that 6 are or may be used by large populations for domestic or industrial water 7 supplies or for irrigation water can only be determined after reliable !
8 assessments have been made of the potential impacts of the reactor plants on 9 the ground water. Accordingly, 10 CFR 100 Subpart B requires that site r 10 environmental characteristics,.which includes hydrological and meteorological 11 characteristics, be characterized and used in or compared to those 12 characteristics used in.the plant PRA and environmental analysis.
13 14 Although management of the. quality of surface waters is important, water ;
15-quality per se is not a determining factor in assessing the suitability cf a 16 site since adequate design alternatives can generally be developed to raeet the -
17 requirements of the Federal Water Pollution Control Act and the Commission's 18 regulations implementing NEPA. However, the environmental characteristics or 19 the complexity of the environment at a site and its vicinity may be such that ,
20 it would be difficult to obtain or develop sufficient information to estab- '
21 lish, in a timely manner, that the potential environmental impacts on water 22 quality would be acceptable. Examples of situations that could pose unusual 23 impact assessment or design problems are areas of existing marginal water 24 quality, small bays, estuaries, stratified waters, and sites that would ,
25 require intake from and discharge to waters of markedly different quality, 26 such as intake of marine water and discharge to an estuary.
27 28 The following are examples of potential environmental effects of station !
29 construction and operation that must be assessed: physical and chemical 30 environmental alterations in habitats of important species, including plant-31 induced rapid changes in environmental conditions; changes in normal current '
32 direction or velocity of the cooling water source and receiving waters '
33 scouring and siltation resulting from construction and cooling water intake 34 and discharges alterations resulting from dredging and spoil disposal; and 35 interference with shoreline processes, r 36 37 s. scological systems and slota 38 39 Areas of great importance to the local aquatic ecosystem may present ,
40 major difficulties in assessing potential impacts on populations of important 41 species'or ecological systems. such areas include those used for breeding 42 (e.g., nesting and spawning), wintering, and feeding, as well as areas where ,
43 there may be seasonally high concentrations of individuals of important 44 species.* Where the ecological sensitivity of a site under consideration
~45
- A species, whether animal or plant, e important (for the 46 purpose of this guide) if a specific causai link can be identified 47 between the nuclear power station and the species and if one or more 48 of the following criteria applies:
49 (1) If the species in commercially or recreationally valuable, 50 (2) If the species is endangered or threatened, 51 (3) If the species affects the well-being of some important 52 species within criteria (1) or (2) or if it is critical to the f 53 structure and function of a valuable ecological system or is 54 a biological indicator of radionuclides in the environment. '
55 '
56 Endangered and threatened species are defined by PL 93-205, .
57 the Endangered Species Act of 1973, as follows: "The term 58 ' endangered species' means any species which is in danger of ;
59 60 extinction throughout all or a significant portion of its range other than a species of the class Insecta determined by the 61 secretary to constitute a post whose protection under the provirions 4.7-8
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_g_
1 cannot be established from existing information, more detailed studies, as 2 discussed in Regulatory Guide'4.2., may be necessary. Impacts of station 3 construction
- and operation on the biota and ecological systems may be miti-4 gated by design and operational practices if justifiable relative to costs and 5 benefits. In general, the important considerations in the balancing of costs 6 and benefits are (a) the uniqueness of a habitat or ecological system within 7 the region under consideration and (b) the amount of habitat or ecological ,
8 system that would be destroyed or disrupted relative to the total amount of 9 the habitat or ecological system present in the region or the vulnerability of 10 the reproductive capacity of important species populations to the effects of 11 construction and operation of the plant and ancillary facilities.
12 13 The alteration of one or more of the existing environmental conditions 14 may render a habitat unsuitable as a breeding or nursery area. In some cases, 15 organisms use identical breeding and nursery areas each year; if the charac-16 teristics of the areas are changed, breeding success may be substantially 17 reduced or enhanced. Destruction of part or all of a breeding or nursery area 18 may cause pcpulation shifts that result in increased competition for the '
19 remaining suitable areas. Such population shifts cannot compensate for the 20 reduced size of the breeding or nursery areas if the remaining suitable area 21 is already occupied by the species. Some species will desert a breeding area 22 because of man's activities in the proximity to the area, even in the absence 23 of physical disturbance of the actual breeding area.
24 25 of special concern relative to site selection are those unique or 26 especially rich feeding areas.that might be destroyed, degraded, or made 27 inaccessible to important species by station construction or operation. Eval-28 untion of feeding areas in relation to potential construction or operation 29 impacts includes the following considerations: size of the feeding area 30 onsite in relation to the total feeding area offsite, food density, time of 31 use, location in relation to other habitats, topography relative to access 32 routes, and other factors (including man's activities). site modification may 33 reduce the quality of feeding areas by destruction of a portion of the food 34 base, destruction of cover, or both.
35 36 construction and operation of nuclear power stations can create barriers
- 37 to migration, occurring mainly in the aquatic environment. Narrow zones of i 38 passage for migratory animals in some rivers and estuaries may be restricted j 39 or blocked by station operation. Partial or complete blockage of a zone of
- 40 passage may result from the discharge of heat or chemicals to receiving water 41 bodies or the construction and placement of power station structures in the 42 water body. strong-swimming aquatic animals often avoid waters of adverse 43 quality, but larval and immature forms are usually moved and dispersed by 44 water currents. It is therefore important in site selection that the routes '
45 and times of movement of the immature stages be considered in relation to 46 potential effects.
47 48 A detailed assessment of potential impact on the species population 49 would be required for sites where placement of intake or discharge structures 50 would markedly disrupt normal current patterns in migration paths of important 51 of this Act would present an overwhelming and overriding risk to 52 man.' 'The term ' threatened species' means any species which is 53 likely to become an endangered species within the foreseeable futur6 54 throughout all or a significant portion of its range." Lists of 55 endangered and threatened species are published periodically in the 56 rederal neoister by the secretary of the Interior.
57 * '
A compilation of construction practices is provided in 'ceneral 58 Environmental Guidelines for Evaluating and Reporting the Ef fects of 59 Nuclear Power Plant Site Preparation, Plant and Transmission
}
60 racilities construction," Atomic Industrial Forum, February 1974. j l 4.7-9 n
a s ?
I I species. The potentials for impingement of organisms on cooling water intake 2
structures and entrainment of organisms through the cooling-system are deter- '
3 mined by a number of variables including site characteristics, intake struc-4 ture design, and placement of the structures at the site.
5 6
7 site characteristics should be considered relative to design and
- 8 placement of cooling system features and the potential of the cooling system !
9 to hold fish in an area longer than the normal period of migration or to l 10 entrap resident populations in areas where they would be adversely affected, i 11 either directly or indirectly, by limited food supply or adverse temperatures.
12 canals or areas where cooling waters are discharged may induce fish to remain 13 in an unnaturally warmed habitat. The cessation of station operation during '
14 winter can be lethal to these fish because of an abrupt drop in water !
temperature.
15 16 f.'I.and Use and AeRhetics .
I 17 i
18 19
-Many impacts on land use at the site and in the site neighborhood due to i 20 construction and operation of the plant, transmission lines, and transpor- l 21 tation corridors can be mitigated by appropriate designs and practices.
22 Aesthetic impacts can be reduced by selecting sites where existing topography 23 and forests can be utilized for screening station structures from nearby 24 scenic, historical, or recreational resources. Restoration of natural vege-25 tation, creative landscaping,' and the integration of structures with the environment can mitigate adverse visual impacts.
26 27 Preconstruction archeological excavations can usually reduce losses.
28 short-term salvage archeology may not be sufficient if extensive or valuable 29 archeological sites are found on the potential site for a nuclear station.
30 For areas of archeological concern, the chief Archeologist of the National 31 32 Park Service is an information source, as are the state Archeologist and the 33 state Liaison officer responsible for the National Historic Preservation Act activities for a particular state.
34 35 Proposed alternative land use may render a site unsuitable for a nuclear 36 power station.
37 For example, lands specified by a community (1) as planned for 38 other uses or be unsuitable. Therefore, (2) as restricted to compatible uses vis-a-vis other lands may 39 official land use plans developed by governments at 40 any level and by regional agencies should be consulted for possible conflicts 41 with power station siting. A list of rederal agencies that have jurisdiction 42 or expertise in land use planning, regulation, or management has been pub-lished by the Council on Environmental Quality."
43 44 45 Another class of impacts involves the preempting of existing land use at the site itself. For example, nuclear power station siting in areas uniquely 46 suited for growing specialty crops may be considered a type of land conversion 47 involving unacceptable economic dislocation.
48 49' sites adjacent to lands devoted to public use may be considered 50 unsuitable. In particular, the use of some sites or transmission lines or 51 transportation corridors close to special areas administered by rederal, 52 state, or local agencies for scenic or recreational use may cause unacceptable i
l 53 impacts regardless of design parameters. such cases are most apt to arise in 54 areas adjacent to nai. ural-resource oriented areas (e.g., Yellowstone National 55 Park) as opposed to recreation-oriented areas (e.g., I.ake Mead National l
56 =
station protection requirements for nuclear safeguards may influence '
l 57 landscape design and clearing of vegetation.
58
- see U.s. council on Environmental cuality,
- Preparation of
! 59 anvironmental Impact statements: cuidelines," 38 FR 20549, August i
60 1, 1973.
4.7-10 1
-. . - - -. ..- - - - _ _ - ~~..- . _~. - . - -
t 1 Recreation Area). Some historical and archeological sites may also fall into 2 this category. The acceptability of sites near special areas of public use j 3 should be determined by consulting cognizant government agencies. ;
4 s
5 6
The following rederal agencies should be consulted for the special areas listed:
I ,
8 a. National Park Service (U.S. Department of the Interior) '
9 10 National Parks; International Parks; National Memorial Parks; 11 National Battlefields, Battlefield Parks and Battlefield Sites; National .
l 12 Military Parks; Historic Areas and National Historic Sites; National Capital '
13 Parks; National Monuments and Cemeteries; National Seashores and Lakeshores; 14 National Rivers and Scenic Riverways; National Recreation Areas; National
- 15- Scenic Trails and Scientific Reserves; National Parkways 16
-17 b. National Park Service Preservation Program 18 '
19 National Landmarks Program; Historic American Buildings Survey; 20 National Register of Historic Places; National Historical Landmarks Program; ;
21 National Park Service Archeological Program 22 l 23 c. sureau of sport Fisheries and wildlife (U.s. Department of 24 25 Interior) ;
26 National wildlife Refuges 27 !
28 d. Forest service (u.s. Department of Agriculture) 29 i 30 National Forest wilderness, Primitive Areas, National Forests.
31 32 Individual states and local governments administer parks, recreation 33 areas, and other public use and benefit areas.
34 Information on these areas 35 should be obtained from cognizant state agencies such as state departments of 36 natural resources. (see publications such as the
- conservation Directory 1973: A Listing of Organizations, A 37 Natural Resource Use and Management,gencies published and byOfficials Concerned the National with wildlife 38 rederation for state-by-state references.) The Advisory Council on Historie 39 Preservation or the appropriate State historical society should be contacted 40 for information on historic areas.
41 42 It should be recognized.that some areas, as yet undesignated, may be 43 unsuitable for siting because.of public interest in future dedication to pub- ,
44 lic scenic, recreational, or cultural use. l 45 Relatively rare land types such as 46 sand dunes and wetlands are prime candidates for such future designation. ,
47 However, the acceptability of sites for nuclear power stations at some future 48 time in these areas will depena on the existing impacts from industrial, commercial, and other developments.
49 50 7. Industrial, military, and Transportation racilities 51 52 53 Potential accidents at present or projected nearby industrial, military, 54 and transportation facilities may affect the safety of a nuclear. power sta-55 tion.' . A site should not be selected if,- in the event of such an accident, it 56 is not possible to safely shut down a plant at that site or if it is not pos-57 sible to have nearby facilities alter their mode of operation or incorporate 58 features to reduce to an acceptable level the likelihood and severity of such potential accidents.
59 60
- section 2.2 of Regulatory Guide 1.70 lists these safety 61 considerations.
4.7-11
i e
- i i
i l ,
1 In the event of-an accident at a nearby industrial facility such as a 2 chemical plant, refinery, mining and quarrying operation, oil or gas well, or >
3 4
gas and petroleum product storage installetion, it is possible that missiles, !
5-shock waves, flammable vapor clouds, toxic chemicals, or incendiary frrgments l 6
may result. These may affect the station itself or the station operators in a :
way.that jeopardizes the cafety of the station. !
7 '
8 Regulatory cuide 1.7s, " Assumptions for Evaluating the Habitability of a ~
9 Nuclear Power Plant control Room During a Postulated Hazardous Chemical >
10 : Release," describes assumptions acceptable to the NRC staff for use in assess- !
-11
-12 ing the habitability of the control room during and after a postulated exter- 1 nal release of hazardous chemicals and describes criteria that are generally !
13 14 acceptable to the staff for the protection of the control room operators. ;
15 Nearby military facilities, such as munitions storage areas and ordnance 16 test ranges, may threaten station safety. The acceptability of a site depends {
17- on establishing, among other things, that the nuclear power station can be i 18 designed so its safety will not be affected by an accident at the military 19 installation. Alternatively, an otherwise unacceptable site may become
, 20 acceptable if the cognizant military organization agrees to change the instal-l 21 -lation or mode of operation to reduce the likelihood or severity of potential 22 23 accidents involving the nuclear station to an acceptable level.
24 An accident during the transport of hazardous materials (e.g., by air, 25 waterway, railroad, highway, or pipeline) near a nuclear power plant may gen-26 erste shock waves, missiles, and toxic or corrosive gases which can affect the l 27 safe operation of the station. The consequences of the accident will depend i 28 the proximity of the transportation facility to the site, the nature and )
29 30 maximum quantity of the hazardous material per shipment, and the layout of the nuclear station. Unless a station can be designed to operate safely in the 31 event of a postulated accident or an enforceable agreement can be reached to 32 limit the transport of hazardous materials or the transportation link can be 33 relocated, the proposed site may not be acceptable.
34 35 Airports are transportation facilities that pose specialized hazards to 36 37 nearby nuclear power stations. Potential threats to stations from aircraft 38 result a crash, from the aircraft itself as a missile and from the secondary effects of e.g., fire.
39 40 s. socioeconomics 41 42 social and economic issues are important determinants of siting policy.
43 It is difficult both to assess the nature of the impacts involved and to 44 determine value schemes for predicting the level or the acceptability of 45 potential impacts.
40 47 The siting, construction, and operation of a nuclear power station may 48 have significant impacts on the socioeconomic structure of a community and may 49 place severe stresses on the local labor supply, transportation facilities, 50 and community services in general. There may be changes in the tax basis and i 51 in community expenditures, and problems may occur in determining equitable 52 levels of compensation for persons relocated as a result of the station sit- !
' 53 ing. It is usually possible to resolve such difficulties by proper coordina- ;
54 tion with impacted communities; however, some impacts may be locally unaccept-55 able and too costly to avoid by any reasonable program for their mitigation. i l
56 Evaluation of the suitability of a site should therefore include consideration l 57 of purpose and probable adequacy of socioeconomic impact mitigation plans for
- 58 such economic impacts on any community where local acceptance problems can be i 59 reasonably foreseen.
t 60
! 61 Certain communities in a site neighborhood may be subject to unusual l 62 impacts that would be excessively costly to mitigate. Among such communities !
j 63 are towns that possess notably distinctive cultural character, i.e., towns i !
i 4.7-12 l ,
i e .
l that have preserved or restored numerous places of historie interest, have
- 2. specialized in an unusual industry-or avocational activity, or have otherwise 3 markedly distinguished themselves from other communities.
4 5 9. moise 6
7 8
Noise levels at nuclear stations occur during both the construction and 9
. operation phases and could have unacceptable impacts. Cooling towers, tur-10
' operation.
bines, and transformers contribute to the noise levels during station 11 12- C. REGULATORY POSITION 13-14 1. Geology / Seismology 15 16 17 16 Preferred sites are those where there is a minimum likelihood of surface L or near surface deformation, or the occurrence of earthquakes on faults in the 19 20 site vicinity (within a radius of 8 km. (5 miles)). Because of the 21 uncertainties and difficulties in mitigating the effects of permanent ground 22 displacement phenomena such as surface faulting or folding, fault creep, 23 subsidence or. collapse, the NRC staff considers it prudent to select an 24 alternate the site.
site when the potential for .oermanent ground displacement exists at
-25 26 27 sites located near geologic structures for which there is an inadequate 28 data base at the time of application to determine their potential for causing 29 surface deformation are likely to be subject to a longer licensing process in 30 view of the need for extensive and detailed geologie and seismic 31 investigations of the site and surrounding region and for the rigorous analyses of the site-plant combination.
32
.33 sites with competent bedrock for foundations generally have suitable 34 foundation conditions. In regions where there are few or no such sites, it is 35 prudent to select sites in areas with competent and stable solid soils, such-36 as dense sands and glacial tills. other materials may also provide satisfac-37 tory foundation conditions, but in any case, a detailed geologic and geotech-38 nical investigation will be required to determine static and dynamic engineer-39 ing properties of the material underlying the site in accordance with Appendix 40 B to 10 CFR Part 100.
t 41 l 42 2. Atmospheric Extremes and nispersion 43 44 As noted in section 5.2 of this guide, site atmospheric conditions are 45 46 site suitability characteristics principally with respect to the calculation 47 of radiation doses resulting from the release of fission products as a con-48 sequence of a postulated accident. _Accordingly, each applicant for site approval must collect meteorological and hydrological information for at least 49 one year that is representative of the site conditions including wind speed, 50 wind direction, precipitation, and atmospheric stability.
51 52 Nonradiological atmospheric considerations such as local fogging and
.53 icieg, cooling tower drift, cooling tower plume lengths and plume interactions
-54 between cooling tower plumes, and plumes from nearby industrial facilities 55 -should be considered in evaluating the suitability of potential sites.
l 56
- 57. 3. Population censideration
. 58
! 59 r.m _ .. yapalation density are preferred for nuclear power station
! 60 sites. High F vsus. ion densities projected for anytime during the lifetime of j 61 a station are considered during both the NRC staff review and the public hear-62 ing phases of the licensing process. If the population density at the pro-
[
63 posed site is not acceptably low, then the applicant will be required to give l
4.7-13
- , e 1 special attention to alternative sites with lower population densities.
2 3 If the population density, including weighted transient population, 4 projected at the time of site approval exceeds 500 persons per square mile 5 averaged over any radial distance out to 30 miles, (cumulative population at a 6 distance divided by the area at that distance), or the projected population 7 density for 40 years after site approval exceeds 1,000 persons per square mile 8 averaged over any radial distance out to 30 miles, special attention should be 9 given to the consideration of alternative sites with lower population 10 densities.
11 12 Transient population should be included for those sites where a 13 significant number of people (other than those just passing through the area) 14 work, reside part-time, or engage in recreational activities and are not 15 permanent residents of the area. The transient population should be taken 16 into account by weighting the transient population according to the fraction 17 of time the transients are in the area.
18 19 Based on past experience, the NRC staff has found that a minimum 20 exclusion distance of 0.4 mile, even with unfavorable design basis atmospheric 21 dispersion characteristics, usually provides assurance that engineered safety 22 features can be designed to bring the calculated dose from a postulated acci-23 dont within the guidelines of 10 CFR Part 50.34(a)(1). Also, based on past 24 experience, the staff has found that a distance of 3 miles to the outer 25 boundary of the low population zone is usually adequate. Subpart B of 10 CFR 26 loo specifies the exclusion area distance. Section 50.34 specifies an LPZ for 27 stationary power reactor applications.
28 29 4. Hydrology 30 31 4.1 riooding 32 33 To evaluate sites located in river valleys, on flood plains, or along 34 coastlines where there is a potential for flooding, the site suitability 35 studies described in Regulatory 1.59,
- Design Basis Floods for Nuclear Power 36 Plants," should be made.
37 38 4.2 water Availability 39 40 A highly dependable system of water supply sources must be shown to be 41 available under postulated occurrences of natural and site-related accidental 42 phenomena or combinations of such phenomena as discussed in Regulatory Guide 43 1.59.
44 45 To evaluate the suitability of sites, there should be reasonable 46 assurance that permits for consumptive use of water in the quantities needed 47 for a nuclear power plant of the stated approximate capacity and type of 48 cooling system can be obtair.ed by the applicant from the appropriate State, 49 local, or regional bodies.
50 51 4.3 water quality 52 53 The potential impacts of nuclear power stations on water quality are 54 likely to be acceptable if effluent limitations, water quality criteria for 55 receiving waters, and other requirements promulgated pursuant to the Federal l 56 Water Pollution Control Act are applicable and satisfied.
l 57 58 The criteria provided in 10 CFR Parts 20 and 50 will be used by the NRC l 59 staff for determining permissible concentrations of radioactive materials i
1 4.7-14 l
e , j i
f 1 discharged to surface water or to ground water.'
2 !
3 4.4 Fission Product Retention and Transport 4 ,
5 To be able to assess fission product retention and transportation via 6 groundwater, the following information should be determined for the sites l 7 ;
8 e soil, sediment, and rock characteristics (e.g., volcanic ash, ,
9 fractured limestone, etc.), 1 10 11 e absorption and retention coefficients for fission product 12 materials, !
13 i 14 e ground water velocity, and 15 16 e distance to nearest body of surface water.
17 i 18 This information should be used in.the environmental report required in 10 CFR t 19 Part 51 and compared to the hydrological information used in the PRA for a e 20 certified design (if such a design is to be located at the site) or used in !
21 the site specific PRA for a custom plant located at the site.
22 23 Aquifers that are or may be used by large populations for domestic, . .
24 municipal, industrial, or irrigation water supplies provide potential pathways 25 for the transport of radioactive material to man in the event of an accident.
.26 To evaluate the suitability of proposed sites located over such aquifers, ;
27 detailed studies of factors identified in section 2.4.13 of Regulatory cuide .
28 1.70, ' Standard Format and Content of Safety Analysis Reports for Nuclear !
29 Power Plau 1,* should be completed. l 30 31 s. scological systems and slota 32 :
33 The ecological systems and biota at potential sites and their '
l 34 environs should be sufficiently well known to allow reasonably certain ,
- 35 predictions that there would be no unacceptable or unnecessary deleterious !
36 impacts on populations of important species or on ecological systems with i l 37 which they are associated from the construction or operation of a nuclear :
l 38 power station at the site. ;
39 40 when early site inspections and evaluations indicate that critical or i 41 exceptionally complex ecological systems will have to be studied in detail to l 42 determine the appropriate plant designs, proposals to use such sites should be l 43 deferred unless sites with less complex characteristics are not available. ;
44 45 It should be determined whether any important species (as defined in 46 Section B.5 of this guide) inhabit.or use the proposed site or its environs; i 47 and the relative abundance and distribution of their populations should be :
48 considered. Potential adverse impacts on important species should be >
49 identified and assessed. The relative abundance of individuals of an impor- ;
50 tant species inhabiting a potential site should be compared to available !
51 information in the literature concerning the total estimated local population.
52 Any predicted impacts on the species should be evaluated relative to effects !
53 on the local population and the. total population of the species. The destruc-54 tion of, or sublethal ~ effects on, a number of individuals which would not '
55 adversely affect the reproductive capacity and vitality of a population or the !
56 crop of an economically important harvestable population or recreationally !
57 important population should generally be acceptable, except in the case of .
t 58 Appendix I to 10 CFR Part 50 provides numerical guidance for design
, 59 objectives and technical specification requirements for limiting l l 60 conditions of operation for light-water-cooled nuclear power l 61 stations.
l 4.7-15 l 5 i
e e I certain endangered species. If there are endangered or threatened species at 2 a site, the potential effects should be evaluated relative to the impact on 3 the local population and the total estimated population over the entire range 4 of the species as noted in the literature.
5 6 It should be determined whether there are any important ecological 7 systems at a site or in its environs. If so, determination should be made as 8 to whether the ecological systems are especially vulnerable to change or if i
9 they contain important species habitats, such as breeding areas (e.g., nesting 10 and spawning areas), nursery, feeding, resting, and wintering areas, or other 11 areas of seasonally high concentrations of individuals of important species.
12 13 The important considerations in the balancing of costs and benefits 14 include the followings the uniqueness of a habitat or ecological system 15 within the region under consideration, the amount of the habitat or ecological 16 system destroyed or disrupted relative to the total amount in the region, and 17 the vulnerability of the reproduetive capacity of important species popula-18 tions to the effects of construction and operation of the station and 19 ancillary facilities.
20 21 If sites contain, are adjacent to, or may impact on important ecological
. 22 systems or habitats that are unique, limited in extent, or necessary to the l
23 productivity of populations of important species (e.g., wetlands and estuar-24 ies), they cannot be evaluated as to suitability for a nuclear power station
! 25 until adequate assessments for the reliable prediction of impacts have been 26 completed and the facility design characteristics that would satisfactorily 27 mitigate the potential ecological impacts have been defined. In areas where 28 reliable and sufficient data are not available, the collection and evaluation 29 of appropriate seasonal data may be required.
l 30 l 31 Migrations of important species and migration routes that pass through l 32 the site or its environs should be identified. Generally, the most critical 33 migratory routes relative to nuclear power station siting are those of aquatic 34 species in water bodies associated with the cooling systems. Site conditions 35 that should be identified and evaluated in assessing potential impacts on 36 important aquatie migratory species include (1) narrow zones of passage, 37 (2) migration periods that are coincident with maximum ambient temperatures, 38 (3) potential for major modification of currents by station structures, 39 (4) potential for increased turbidity during construction, and (5) potential 40 for entrapment, entrainment, or impingement by or in the cooling water system, 41 or blocking of migration by facility structures of effluents.
42 1 43 The potential blockage of movements of inportant terrestrial animal 44 populations due to the use of the site for a nuclear power station and the 45 availability of alternative routes that would provide for maintenance of the 46 species' breeding population should be assessed.
47 48 If justifiable relative to costs and benefits, potential impacts of
- 49 plant construction and operation on the biota and ecological systems can 50 generally be mitigated by adequate engineering design and site planning and by 51 proper construction and operation practice when there is adequate information 52 about the vulnerability of the important species and ecological systems.
l 53 54 A summary of environmental considerations, parameters, and regulatory 55 positions for use in evaluating the suitability of sites for nuclear power 56 stations is provided in Appendix B to this guide. A discussion of ecological l 57 systems and habitats, the level of detail that should be addressed the site 58 selection process, and the survey, monitoring, and analytical techniques for 59 assessing impacts on important species and ecological systems will be sum-60 marized in subsequent appendices to this guide.
61 62 6. Land Use and Aesthetict 63 4.7-16 l
l
e 6 1 Land use plans adopted by Federal, State, regional, or local 2 governmental entities should be examined, and any conflict between these plans 3 and use of a potential site should be resolved by consultation with the 4 appropriate governmental entity.
5 6 For potential site on land devoted to specialty crop production where 7 changes in land use might result in market dislocations, a detailed 8 investigation should be provided to demonstrate that potential problems have 9 been identified and resolved.
10 11 The potential aesthetic impact of nuclear power stations at sites near
.12 natural-resource oriented public use areas is of particular concern, and 13 evaluation of the suitability of such sites is dependent on consideration of 14 specific station design layout. However, existing aesthetic impacts at poten-15 tial sites should be taken into account as mitigating any requirements for 16 further special cesign.
17 18 7. Industrial, military, and Transportation Facilities 19 20 Potentially hazardous facilities and activities within.5 miles of a 21 proposed site should be identified. If a preliminary evaluation of potential 22 accidents at these facilities indicates that the potential hazards from shock 23 waves and missiles approach or exceed those of the design basis tornado for 24 the region
- or potential hazards such a flammable vapor clouds, toxic chemi-25 cals, or incendiary fragments exist, the suitability of the site should be 26 determined by detailed evaluation of the degree of risk imposed by the 27 potential hazard.
28 29 The identification of design basis events resulting from the presence of 30 hazardous materials or activities in the vicinity of a nuclear power station 31 is acceptable if the design basis events include each postulated type of 32 accident for which a realistic estimate of the probability of occurrence of 33 potential exposures in excess 4
of the 10 CFR Part 50.34(a)(1) guidelines 34 exceeds approximately 10 per year. Because of the difficulty of assigning 35 precise numerical values to the probability of occurrence of the types of l 36 potential hazards generally considered in determining the acceptability of l 37 sites for nuclear stations, judgment must be used as to the acceptability of j 38 the overall risk presented by an event.
39 40 In view of the low probability events under consideration, the 41 probability of occurrence of the initiating events leading to potential con-42 sequences in excess of 10 CFR Part 50.34(a)(1) exposure guidelines should be 43 based on assumptions that are as realistic as is practicable. In addition, 44 because of the low probability events under consideration, valid statistical 45 data are often not available to permit accurate quantitative calculation of 46 probabilities. Accordingly, a conservative calculat'.on showing that the 47 probability of occurrence of potential exposures in excess of the 10 CFR Part 48 50.34(a)(1) guidelines is approximately 10d per year is acceptabis if, when 49 combined reasonable qualitative arguments, with the realistic probability can 50 be shown to be lower.
51 52 The effects of design basis events have been appropriately considered if 53 analyses of the effects of those accidents on the safety-related features of a 54 proposed nuclear. station have been performed and appropriate measures (e.g.,
55 hardening fire protection) to mitigate the consequences of such events have 56 been taken.
57 58 To evaluate the suitability of sites in detail for potential accidents 59 involving hazardous materials and activities at nearby industrial, military, 60 a The design basis tornado is described in Regulatory Guide 1.76, 61
- Design Basis Tornado for Nuclear Power Plants."
4.7-17
t e 1
2 and latorytransportation facilities, Guide 1.70 should the studies described in section 2.2 of Regu-be made.
3 4 8. Socioeconomics 5
6 7 The NRC staff considers that an evaluation of the suitability of nuclear 8 power station sites near distinctive communities should demonstrate that the 9 construction and operation of the nuclear station, including transmission and 10 transportation corridors, and potential problems relating to community ser-11 vices, such as schools, police and fire protection, water and sewage, and 12 health facilities, will not adversely affect the distinctive character of the 13 community. A preliminary investigation should be made to identify and analyze 14 problems.that proposed site.
may arise due to the proximity of a distinctive community to a 15 16 9. mois.
, 17 l 118 19 Noise levelst at proposed sites must comply with applicable Federal, state, and local noise regulations.
20 21 lo. Energency planning 22 23 24 As.a minimum, each applicant for site approval should provide a 25 description of the area within a 10 mile radius of the plant EPZ, including:
26 o population distribution (current and projected for the next 40 27 years),
29 29 o residential, industrial, public, and commercial facilities and 30 structures, 31 32 o transportation routes, including any egress limitations, and 33 34 o topography.
35 36 In addition, the applicant shall provide a description of any contacts, 37 evaluations by and assessments with local, state, and Federal government 38 . agencies with emergency planning responsibilities. An evaluation of the above 39 information with respect to its impact on the development of an emergency 40 plant that can assure adequate protective measures for the populace should be 41 provided.
42 43 n. rurt.ausuTArzou 44-45 This guide discusses the major site characteristics related to public 46 health and safety and environmental issues which the NRC staff considers in 47 determining the suitability of sites for light-water-cooled (LWR) nuclear 48- power stations. Accordingly, it.can be used after [ EFFECTIVE DATE) to 49 indicate considerations that should be addressed in the initial stage of the 50 site selection process to identify potential sites for nuclear power stations.
51 l
4.7-18
e s l
1 APPENDIX A 2
3 SAFETY-RELATED SITE CONSIDERATIONS 4 FOR ASSESSING SITE SUITABILITY 5 FOR NUCLEAR POWER STATIONS 6
7 This appendix provides a checklist of safety-related site charac-8 teristics, reletrant regulations and regulatory guides, and regulatory 9 experience and positions for assessing site suitability for nuclear power 110 , stations.
11 h
l l
4.7-19
l >
l -
l Considerations Relevant Regulations Regulatory Experlence and Regulatory Guides and Position ;
j 1 A.1 Geology / Seismology
! 2 l 3 Geologic and seismic char. 10 CFR Part 100, Appendix B, Where the potential for permanent 4 acteristics of a site, such as " Criteria for the Seismic and ground deformation such as 5 surface faulting, ground Geologic Siting of Nuclear . faulting, folding, subsidence or 6 motion, and foundation condi- Power Plants after collapse exists at a site, the NRC r 7 tions (including liquefaction, [ EFFECTIVE DATEl." staff considers it prudent to select 8 subsidence, and landslide an alternate site. ;
9 potential), may affect the Regulatory Guide 1.70, l 10- safety of a nuclear power Chapter 2 (identifies safety- Sites should be selected in areas
- 11. station. related site characteristics). for which an adequate geologic
~
12- data base exists to determine 13 Regulatory Guide 1.29 " capability." Delay in licensing 14 (discusses plant safety can result from a need for exten- )
15 features which should be sive geologic and seismic investi-i
'16 controlled by engineering gations. Conservative design of I l 17 design), safety-related structures will be
]
! 18 required when geologic, seismic,
! 19 and foundation information is 20 l i
questionable.
! 21 l 22 Sites with competent bedrock
! 23 generally have suitable foundation l 24 conditions.
25 26 If bedrock sites are not available, 27 it is prudent to select sites in 28 areas known to have a low subsi-
- 29 dence and liquefaction potential.
! 30 investigations will be required to 31 determine the static and dynamic 32~ engineering properties of the 33 material underlying the site as
, 34 stated in 10 CFR Part 100, l
35 Appendix A and Appendix B.
36 37 l
i i
i i
i i
t 4.7-20
O 6 Considerations ReIevant Regulations Regulatory Experience and Regulatory Guides and Position 1 A.2 Atmospheric Dispersion 2
3 The atmospheric conditions at 4 a site should provide sufficient 10 CFR Part 50, " Domestic Unfavorable safety-related design j 5 dispersion of radioactive Licensing of Production and basis atmospheric dispersion l 6 materials released during a Utilization Facilities." characteristics can be l 7 postulated accident to reduce compensated for by engineered i 8 the radiation exposures of Regulatory Guide 1.23, safety features. Accordingly, the 9 individuals at the exclusion "Onsite Meteorological regulatory position on atmospheric 10 area and low population zone Programs." dispersion of radiological effluents 11 boundaries to the values is incorporated into the section 12 prescribnd in 10 CFR Part Regulatory Guide 1.3 " Population Considerations" (see 13 50.34. " Assumptions Used for A.3 of this appendix).
14 Evaluating the Potential 15 Radiological Consequences of 16 a Loss of Coolant Accident for l 17 Boiling Wate& Reactors."
l 18 l 19 Regulatory Guide 1.4, 20 " Assumptions Used for 21 Evaluating the Potential 22 Radiological Consequences of 23 a Loss of Coolant Accident for 24 Pressurized Water Reactors."
25 26 Regulatory Guide 1.5, 27 " Assumptions Used for Eval-l 28 uating the Potential
, 29 Radiological Consequences of l 30 a Steam Line Break Accident l 31 for Boiling Water Reactors."
32 33 Regulatory Guide 1.24, 34 " Assumptions Used for 35 Evaluating the Potential l
'36 Radiological Consequences of l 37 a Pressurized Water Reactor 38 Radioactive Gas Storage Tank 39 Failure."
40 41 Regulatory Guide 1.25, 42 " Assumptions Used for 43 Evaluating the Potential 44 Radiological Consequences of 45 a Fuel Handling Accident in 46 the Fuel Handling and Storage 47 Facility for Boiling and 48 Pressurized Water Reactors."
49 50 4
3 4.7-21 l
a e Considerations Relevant Regulations Regulatory Experience and Regulatory Guides and Position 1
1 A.3 Population Considerations 2
3 in the event of a serious accident 10 CFR Part 100,
- Reactor Site if the population density, including 4 at a nuclear power station, Criteria," requires the following: weighted transient population, 5 effective action must be taken to projected at the time of initial site 6 minimize exposure of individuals e An " exclusion area
- approval exceeds 500 persons per 7 outside the station to any surrounding the reactor in which square mile averaged over any radial 8 radioactive materials which may the reactor licensee has the distance out to 30 miles (cumulative 9 be released during the accident. authority to determine all population at a distance divided by 10 To ensure that exposure to activities, including exclusion or the area at that distance), or the 11 populations will be minimized in removal of personnel and projected population density for 40 12 the event of an accident, the property; years after site approval exceeds 13 nuclear power station should not 1,000 persons per square mile averag-14 be located in a densely populated
- Domestic ed over any radial distance out to 30 l 15 area. Licensing of Production and miles, special attention should be j l 16 Utilization Facilities." given to the consideration of 17 altemative sites with the lower 18 e A ' low population zone" (LPZ) population densities.
l 19 which immediately surrounds l 20 the Transient population should be l 21 exclusion area, included for those sites where a 22 significant number of people (other 23 e At any point on the exclusion than those just passing through the 24 area boundary and on the outer area) work, reside part time, or
- 25 boundary of the LPZ the expo- engage in recreational activities, and 26 sure are not permanent residents of the 27 of individuals to a postulated area. The transient population should 28 release of fission products (as a be taken into account by weighing the 29 consequence of an accident) be transient population according to the 30 less than certain prescribed fraction of time the transients are in 31 values, the area.
32 33 Based on past experience the NRC 34 Regulatory Guides 1. 3, 1.4, 1. 5, staff has found that a minimum 35 1.24 and 1.25 give calculational exclusion distance of 0.4 mile, even 36 methods (see A.2 of this with the most unfavorable design 37 appendix.) basis atmospheric dispersion charac-38 teristics, provides assurance that 39 engineered safety features can be 40 added that will bring the calculated 41 doses from a postulated accident 42 within the guidelines of 10 CFR Part 43 50.34. Also based on past 44 experience, the NRC staff has found 45 that a distance of 3 miles to the outer l 46 boundary of the LPZ is usually 47 adequate.'
48 49 *The guidelines nurnbers for exclusion eres and LPZ are based on historical siting experience of light-water-cooled reactors.
4.7-22 l
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[
[
Cbnsiderations Relevant Regulations Regulatory Experience t
l
.and Regulatory GJides and Position i
j 1 A.4 Hydrology l 2 l 3 A.4.1 Flooding i
4 5 Precipitation, wind, or 10 CFR Part 100, Appendix B, To evaluate sites located in river :
6 seismically induced flooding " Criteria for the Seismic and valleys, on flood plains, or along '
7 (e.g., resulting from dam Geologic Siting of Nuclear coastlines where there is a 8 failure, from river blockage or Power Plants after potential for flooding, the studies 9 diversion, or frcm distantly (EFFECTIVE DATE)." described in Regulatory Guide 10 and locally generated sea 1.59 should be made.
l 11 waves) can affect the safety Regulatory Guide 1.59, L 12 of a nuclear power station. " Design Basis Floods for l 13 Nuclear Power Plants."
! 14 15 Regulatory Guide 1.70, l 16 " Standard Format and Content ;
i . 17 of Safety Analysis Reports for i i 18 Nuclear Power Plants,"
i 19 (Section 2.4). !
20 21 10 CFR Part 50, Appendix A, 22 " General Design Criteria for
~
23 Nuclear Power Plants;"
)
24 Criterion 2, " Design Bases for :
25 Protection Against Natural '
26 Phenomena."
27 28 A.4.2 Water Supply 29
, 30~ A safety-related water supply 10 CFR Part 100, Appendix B, - A highly dependable system of l 31 is required for normal or " Criteria for the Seismic and water supply sources should be 32 emergency shutdown and. Geologic Siting of Nuclear shown to be available under 33 cooldown. Power Plants after postulated occurrences of natural !
34 [ EFFECTIVE DATE]." phenomena and site-related !
35 accidental phenomena or ,
36 Regulatory Guide 1.59, combinations of such phenomena ;
37 " Design Basis Floods for as discussed in Regulatory Guide 38 . Nuclear Power Plants." 1.59.
l 39 40 Regulatory Guide 1.27, To evaluate the suitability of a 41 " Ultimate Hnt Sink for site, there must a reasonable 42 Nuclear Power Plants." assurance that permits for water 43 use and for water consumption in 44 the quantities needed for a 45 nuclear power plant of the stated 46 approximate capacity and type of 47 cooling system can be obtained i 48 by the applicant from the ap-
{ 49 propriate State, local, or regional
! 50 bodies.
l 51 f
4.7-23
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- ' T -
.- , _. m. . . _ . . . - _ _.,m- . . . . . . _ . , . , . .,
- . . . - - - . . . . . . - . _ ~ _ - - _ . . - . . - . . _ - - . - - . - - _ . - . - . . . .-
Considerations ReIevant Regulations Regulatory Experlence and Regulatory Guides and Position 1 A.4.3 Water Quality 2
3 contamination of ground 10 CFR Part 20, " Standards The criteria provided in 10 CFR '
4 water and surface water by For Protection Against Parts 20 and 50 will be used by 5 radioactive materials Radiation." the NRC staff for determining 6' discharged from nuclear permissible concentrations of 7 stations could cause public 10 CFR Part 50, " Licensing of radionuclides discharged to 8 health hazards. Production and Utilization surface water and ground water.
9 Facilities."
10 l
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4.7-24 l
o s Considerations Relevant Regulations Regulatory Experience and Regulatory Guides and Position 1 A.5 Industrial, Military and 2 Transportation Facilities Near 3 the Site.
4 5 Accidents at present or 10 CFR Part 100, " Reactor Potentially hazardous facilities and 6 projected nearby industrial, Site Criteria", Subpart B, activities within 5 miles of a 7 military, and transportation Section 100.22. proposed site must be identified.
8 facilities may affect the safety if a preliminary evaluation of 9 of the nuclear power station. 10 CFR Part 50, Appendix A, potential accidents of these I 10 " General Design Criteria for facilities indicates that the 11 Nuclear Power Plants," potential hazards from shock !
12 Criterion 4, " Environmental waves and missiles approach or 13 and Missile Design Bases." ' exceed those of the design basis 14~ tornado for the region (the design 15 Regulatory Guide 1.70, basis tornado is described in 16 " Standard Format and Content Regulatory Guide 1.76), or poten-17 of Safety Analysis Reports," tial hazards such as flammable .
18 Section 2.2 (lists types of vapor clouds, toxic chemicals, or l 19 facilities and potential incendiary fragments exist, the 20 accidents), suitability of the site should be 1 21 determined by detailed evaluation 22 Regulatory Guide 1.78, of the potential hazard.
23 " Assumptions for Evaluating l 24 the Habitability of a Nuclear The identification of design basis <
25 Power Plant Control Room events resulting from the i 26 During a Postulated Hazardous presence of nearby hazardous l 27 Chemical Release." materials or activities in the 28 vicinity of a nuclear power station- l 29 is acceptable if the design basis 30- events include each postulated 31 type of accident for which a 32 realistic estimate of the probability :
33 of occurrence of potential expo. '
34 sures in excess of 10 CFR Part 35 50.34 guidelines exceeds i 36 approximately 104 per year. '
37 38 To evaluate the suitability of sites 39 in detail for potential accident 40 situations involving hazardous 41 materials and activities from 42 nearby industrial, military, and i 43 transportation facilities, the i 44 studies described in Section 2.2 45 of Regulatory Guide 1.70 should 46 be made.
47 48 4.7-25 l
e o 1
2 APPENDIX B .
3 4 ENVIRONMENTAL CONSIDERATIONS FOR ASSESSING 5 SITE SUITABILITY FOR NUCLEAR POWER STATIONS i 6
7 !
8 This appendix summarizes environmental considerations related to site characteristics that 9 should be addressed in the early site selection process. The relative importance of the different 10 factors to be considered varies with the region or State in which the potential sites are located.
11 12 Site Selection processes can be facilitated by establishing limits for various parameters based 13 on the best judgment of specialists knowledgeable of the region under consideration. For example, 14 limits can be chosen for the fraction of water that can be diverted in certain situations without 15 adversely affecting the local populations of important species. Although simplistic because 16 important factors such as the distribution of important species in the water body are not taken into '
17 account, such limits can be useful in a screening process for site selection. I 18 19 A discussion of performance characteristics of light-water-cooled reactor stations which may 20 affect the environment is given in WASH 1355, " Nuclear Power Facility Performance 21 Characteristics for Making Environmental Impact Assessments," December 1974.
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i 4.7-26
. Considerations ParaTeters Regulatory Position l
l l
l' B.1 Preservation of important 2 Habitats 3
- 4. Important habitats are those The proportion of an In general, a detailed justification 5 that are essential to important habitat that would should be provided when the 6 maintaining the reproductive be destroyed or significantly destruction or significant 7 capacity and vitality of altered in relation to the total alteration of more than a few 8 important species populations
- habitat within the region in porcent of important habitat types 9 or the harvestable cup of which the proposed site is to is proposed.
10 economically or recreationally be located is a useful 11 important species. Such parameter for estimating The reproductive capacity of 12 habitats include breeding potentialimpacts of the populations of important species 13 areas (e.g., nesting and construction or operation of a and the harvestable crop of 14 spawning areas), nursery, nuclear power station. The economically or recreationally 15 feeding, resting, and wintering value of the proportion varies important populations must be
- 16 - areas or other areas of among species and among maintained unless justification for 17 seasonally high concentrations habitats. The region consid- proposed or probable changes can 18 of individuals of important ered in determining pro- be provided, i 19 species. portions is the normal l 20 geographic range of the i 21 The construction and opera- specific population in 22 tion of nuclear power stations question.
23 (including new transmission l 24 lines and access corridors con. If endangered or threatened l 25 structed in conjunction with species occur at a site, the
! 26 the station) can result in the potential effects of the 27 destruction or alteration of construction and operation of 28 habitats of important species a nuclear power station should 29 leading to changes in the be evaluated relative to the 30 abundance of a species or in potential impact on the local 31 the species composition of a population and the total 32 community. estimated population over the l 33 entire range of species.
34 35 See also Chapter 2 of 36 Regulatory Guide 4.2, 37 " Preparation of Environmental 38 Reports for Nuclear Power i 39 Stationa."
40 41 42' 'As defined for this guide in Section B.
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l Considerations ParaTeters Regulatory Position
.I 8.2 Migratory Routes of l 2 important Species 1 3 J 4 Seasonal or daily migrations The width or cross-sectional Narrow reaches of water bodies 'i
-5 are essential to maintaining area of a water body at a should be avoided as sites for i 6 the reproductive capacity of proposed site relative to the locating intake or discharge 7 some important species general width or cross- structures.
8 populations, sectional area in the portion of 9' the water used by migrating A zone of passage that will permit 10 Disruption of migratory species should be estimated, normal movement of important 11 patterns can iesult from species populations and 12 partial or complete blockage Suggested minimum zones of maintenance of the harvestable 13 of migratory routes by passage range from 1/3 to crop of economically important '
14 structures, discharge plumes, 3/4 of the width or cross- populations should be provided.
15- environmental alterations, or sectional areas of narrow ,
16 man's activities (e.g., trans- water bodies.**
-17 portation or transmission '
18 corridor clearing and site Some species migrate in 19 preparation). central, deeper areas while ,
20 others use marginal, shallow -
21 areas. Rivers, streams, and
'22 estuaries are seldom '
23 homogeneous in their lateral 24 dimension with respect to 25- - depth, current velocity, and 26 . habitat type. Thus, the use of 27 width or cross-sectional area ;
28 criteria for determining 29- adequate zones of passage 30 should be combined with a
'31 knowledge of important 32 species and their migratory 33 requirements.
34 35 36
- Water Quality Criteria.1972, National Academy of Sciences - National Academy of v
.37 Engineering, Washington, D.C.,1972.
38
- Handbook of Environmental Control. Volume Ill: Water Sunolv and Treatment. R.G.
39 Bond and C.P. Straub (Editors), CRS Press, Cleveland, Ohio,1973.
i 4.7-28
~
r O O Cbnsiderat ions Parareters Regulatory Position 1 8.3 Entrainment and 2 impingement of Aquatic 3 organisms 4
5 Plankton, including eggs, The depth of the water body The site should have cha-6 larvae, and juvenile fish, can at the point of intake relative racteristics that allow placement 7 be killed or injured by to the general depth of the of intake structures where the 8 entrainment through power water body in the vicinity of relative abundance of important 9 station cooling systems or in the site. species is small and where low 10 discharge plumes. approach velocities can be 11 The proportion of water attained. (Deep regions are 12 The reproductive capacity of withdrawn relative to the net generally less productive than 13 important species populations new available water at the shallow areas. It is not implied 14 may be impaired by lethal site is an indirect measure of that benthic intakes are 15 stresses or by subiethal the destruction of plankton necessary.)
16 stresses that affect reproduc- which in turn is indicative of 17 tion of individuals or result in possible effects on Important habitats (see B.1) 18 increased predation on the popula* ions of important should be avoided as locations for 19 affected species population, species. It has been intake structures.
20 suggested that the fraction of 21 Fish and other aquatic available new water that can 22 organisms can be killed or be diverted is in the range of 23 injured by impingement on 10% to 20% of flow.**
24 cooling water intake screens
- 25 or by entrainment in discharge The simplistic parameter 26 plumes. (proportion of water with-27 drawal) is suitable for use in a 28 screening process or site 29 selection. However, other 30 factors such as distribution of 31 important species should be 32 considered and in all cases the 33 advice of experts on the local
- 34 fisheries should be consulted l 35 to ensure that proposed
( 36 withdrawals will not be l 37 excessive.
l 38 39-40 ' Approach velocity and screen-face velocity are design criteria that may affect the impingement of 41 larger organisms, principally fish, on intake screens. Acceptable approach and screen-face velocities
, 42 are based on fish swim speeds which will vary with the species, site and season.
l 43 *The Water's Edae: Critical Problems of the Coastci Zone, B.H. Ketchum (Editor), MIT Press, 44 Cambridge, Mass.,1972.
45 *Encineerina for Resolution of the Enerav-Environment Dilemma. National Academy of Engineering, 46 Washington, D.C.1972.
l 4.7-29 l
m u 4 4 Considerations ParaTeters Regulatory Position l
1 B.4 Entrapment of Aquatic 2 Organisms 3
4 Cooling water intake and Site characteristics that will Sites where the construction of 5 discharge system features, accommodate design fer es intake or discharge canals would 6 such as canals and thermal that mitigate or prevent be necessary should be avoided 7 pumes, can attract and entrap entrapment.
8 unless the site and important organisms, principally fish, 9 species characteristics are such The resulting concentration of 10 important fish species near that entry of important species to the canal can be prevented or il the station site can result in limited by screening.
12 higher mortalities from station-13 related causes, such as im-14 pingement, cold shock, or gas 15 bubble disease, than would 16 otherwise occur.
17 18 Entrapment can also interrupt 19 normal migratory patterns.
20 21 B.5 Water Quality 22 23 Effluents discharged from Applicable EPA approved 24 nuclear power plants are Pursuant to Section 401(a)(1) of State water quality stsr:dords. the FWPCA, certification from the 25 governed under the authority 26 State that any discharge will of the Federal Water Pollution For states without EPA- comply with applicable effluent 27 Control Act (FWPCA)--(PL 92- approved water quality limitations and other water pollu-28 500). standards, thr water quality tion control requirements is 29 criteria listed in Water Quality necessary before the NRC can 30 Criteria,1972' will be used for issue a construction permit unless 31 evaluation, 32 the requirement is waived by the 33 State or the State fails to act 34 within a reasonable length of 35 time.
36 37 issuance of a permit pursuant to 38 Section 402 of the Act is not a 39 prerequisite to an NRC license or 40 permit.
41 42 Where station construction or 43 operation has the potential to 44 degrade water quality to the pos-45 sible detriment of other users,
, 46 more detailed analyses and I 47 evaluation of water quality may 48 be necessary.
49 50
- Water Quality Criteria.1972, National Academy of Sciences--National Academy of 51 Engineering, Washington, D.C.1972.
i 4.7-30
Considerations Paramters Regulatory Positlon 1 B.6 Water Availability 2
3 The consumptive use of water Applicable Federal, State, and Water use and consumption must 4 for cooling may be restricted local statutory requirements. comply with statutory 5 by statute, may be requirements and be compatible 6 inconsistent with water use Compatability with water use with water use plans of cognizant 7 planning, or may lead to an plan of cognizant water water resources planning 8 unacceptable impact to the resource planning agency. agencies.
9 water resource.
10 in the absence of a water use Consumptive use should be 11 plan, the effect on other restricted such that the supply of 12 water users is evaluated other users is not impaired and 13 considering flow or volume that applicable surface water 14 reduction and the resultant quality standards could be met, 15 ability of all users to obtain assuming normal station 16 adequate supply and to meet operational discharges and 17 applicable water quality extreme low flow conditions 18 standards (see B.5, Water defined by generally accepted 19 Quality). engineering practices.
20 21 For multipurpose impounded lakes 22 and reservoirs, consumptive use 23 should be restricted such that the 24 magnitude and frequency of 25 drawdown will not result in 26 unacceptable damage to impor-27 tant habitats (see B.1, Preser.
28 vation of Important Habitats) or 29 be inconsistent with the 30 management goals for the water 31 body.
32 33 B 7 Established Public
- 34- Amenity Areas 35 36 Areas dedicated by Fedeet Proximity to public amenity Siting in the vicinity of designated ,
37 State, or local governments to area. Viewability (see B.10, public amenity areas will generally 38 scenic, recreational, or cultural Visual Amenities). require extensive evaluation and 39 purposes are generally prohi- Justification. ;
40 bited areas for siting power 41 stations. The evaluation of the suitability of ,
42 sites in the vicinity of public_
43 Siting nuclear power stations amenity areas is dependent on 44- in the vicinity of established consideration of a specific plant 45 public amenity areas could design and station layout in 46 result in the loss or deteriora- relation to potential impacts on 47 tion of important public the public amenity area.
48 amenities.
49 4.7-31
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Cbnsiderations Par m ters. Regulatory Position 1 B.8 Prospective Designated i
.2 Amenity Areas '
3 4 Areas containing important Comparison of possible Public amenity areas that are 5 resources for scenic, recrea- amenity areas in number and distinctive, unique, or rare in a
.6 tional, or cultural use may not extent with other similar areas region should be avoided as sites 7 currently be oesignated as available on a local, regional, for nuclear power stations.
8 such by :ublic agencies but or national basis, as 9 may involve a not loss to the appropriate.
110 public if coriverted to power
, .11- generation. Thcse areas may
^12 ' include locally rare land types, 13 such as sand dunes, wet-14 lands, or coastal cliffs. l
.25 16 B.9 Public Planning 17 18 Land use for a nuclear power Officially adopted land use Land use plans adopted by 19 station should be compatible plans. Federal, State, regional, or local 20 with established land use or . government entities must be ,
21 zoning plans of governmental examined, and any conflict '
22 . entities, between these plans and use of a .I 23 proposed site must be resolved by 1 24 consultation with the appropriate 25- governmental entity, i 26 l 27- B.10 Visual Amenities l 28 l 29 The presence of power station - The solid angle subtended by The visualintrusion of nuclear
-30 structures may introduce station structures at critical power station structures as 31 adverse visualimpacts to resi- viewing points. viewed from nearby residential, 32 dential, recreational, scenic, or recreational, scenic, or cultural 33 cultural areas or other areas areas should be controlled by 34 with significant dependence selecting sites where existing 35 on desirable viewing topography and forests can be 36 characteristics, utilized for screening station 37- structures from those areas in 38 which visualimpacts would 39- otherwise be unacceptable.
40 41 B.11 Local Fogging and Icing 42 43 . Water and water vapor increase in number of hours of The hazards on transportation 44 released to the atmosphere fogging or icing caused by routes from fog or ice that result 45 from recirculating cooling operation of the station. from station operation should be 46 systems can lead to ground . evaluated. The evaluation should 47 fog and ice resulting in include estimates of frequency of 48 transportation hazards and occurrence of station induced fog-49 damage to electric ging and icing and their impact on 50 transmission systems, transportation, electrical trans-51 mission, and other activities and 52 functions.
4.7-32
m a >- m .,s a nsaan, . - , - . a - +~ - ..mw..x._ .
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Considerations ParaTuters Regulatory Position 1 B.12 Cooling Tower Drift 2
3 Concentrations of chemicals, The percent drift loss from The potential loss of important 4 dissolved solids, and . recirculating condenser terrestrial species and other 5 suspended solids in cooling cooling water, particle size resources should be considered.
6 tower drift could affect ter- distribution, salt deposition 7 restrial biota and result in rate, local atmospheric condi-8 unacceptable damage to tions, and loss of sensitive 9 vegetation and other terrestrial biota affected by ,
10 resources. salt deposition from cooling 11 tower drift.
12 !
13 B.13 Cooling Tower Plume 14 1.engths 15-16 Natural draft cooling towers The number of hours per year The visibility of cooling tower 17 produce cloud-like plumes the plume is visible as a plumes as a Nnction of direction 18 which vary in size and altitude function of direction and dis- and distance fcom cooling towers 19- depending on the atmospheric tance from the cooling should be consdered. The evalu-
- 20. conditions. The plumes are towers. ation should include estimates of 21 usually a few miles in length frequency of occurtance for 22 before becoming dissipated, plumes as well as potential 23 although plume lengths of 20 hazards to aviation in the vicinity 24 to 30 miles have been . of commercial and military
- 25. reported from cooling towers. airports.
26 Visible plumes emitted from 27 cooling towers could cause a 28 hazard to commercial and 29 military aviation in the vicinity i 30 of commercial and military 31 airports. The plumes 32 themselves or their shadows 33 could have aesthetic impacts.
34 35 B.14 Plume Interaction
'36 ,
37 Water vapor from cooling The degree to which impacts The hazards to public health, 38 tower plumes may interact may occur will vary depending structures, and other resources 39 with industrial emissions from on the distance between the from potential plume interaction 'i
-40 nearby facilities to form nuclear and fossil-fueled sites, between cooling tower plumes 41 noxious or toxic substances the hours per year of plume and plumes from fossil-fueled 42- which could cause adverse interaction, the type and sites and industrial emissions from 43 public health impacts, or concentration of chemical .
nearby facilities should be 44 result in unacceptable levels reaction products, the area of considered.
-45 of damage to biota, chemical fallout, and the local >
l 46 structures, an other resources, atmospheric conditions,
- j. 47 l
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Considerations Pa rimeters - Regulatory Position I 1 B.15 Noise i 2
3 Undesirable noise levels at Applicable Federal, State, and Noise levels at proposed sites -
4 nuclear power stations could local noise regulations. must comply with statutory 5 occur during both the requirements.
6 - construction and operation 7 phases and have unacceptable 8 impacts near the plant.
9 l 10 B.16 Economic impact of l 11 Preemptive Land Use l
12 l 13 Nuclear power stations can The level of local economic if a preliminary evaluation of net l
14 preempt large areas, dislocet!ca, such as loss of local economic impact of the use i l 15 especially when large cooling income, jobs, and production, of productive land for a nuclear i 16 lakes are constructed. The caused by preemptive use of power station indicates a potential 17 land requirement is likely to be productive land and its effect for large economic dislocation, the 18 an important issue when a on meeting foreseeable NRC staff will require a detailed 19 proposed site is on productive national demands for agricul- evaluation of the potentialimpact 20 land (e.g., agricultural land) ture products. and justification for the use of the 21 that is locally limited in avail- site based on a cost-effectiveness 22 ability and is important to the comparison of alternative station 23 local economy, or which may designs and site-station combina-l 24 be needed to meet foreseeable tions. To complete its evaluation, l
25 national demands for agri- the staff will also need informa-j 26 cultural products. tion on whether and to what i l 27 extent the land use affects 28 national requirements for agricul-29 tural products.
30 31 l
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ENCLOSURE 7 I
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DRAFT REGULATORY GUIDE DG-1015
- SEISMIC SOURCES 1
i l
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2 I l i
- j. SPECIFIC ISSUES FOR COMMENTS
! I I
e i The proposed guide, DG-1015, outlines concepts and procedures to be used
! in conjunction with the probabilistic/ deterministic seismic hazard evaluations.
Rationale for the approach is discussed in Section V.B(3) of the Federal Register notice that published the revision of Appendix A to Part 100 for public comment.
j The staff is currently performing confirmatory studies to evaluate and 2
refine these proposed procedures. A limited study has been completed demonstrating the feasibility of procedures and the validity of the concepts.
However, the staff would like to solicit comments on the concepts outlined in the l 4
proposed guide at this time. To facilitate the review, results of the ,
application of the proposed procedure to four test sites are published separately (Letter report from D. Bernreuter of LLNL to A. Murphy of NRC). l There are divergent views on the role probabilistic seismic hazard analysis '
should play in the licensing arena. There is a general consensus within the NRC l staff that the revised seismic and geological siting criteria should allow j considerations for a probabilistic hazard analysis. There is also a general ~
belief that the probabilistic analysis should be calibrated against the past practices for siting and licensing the current generation of nuclear power plants. There is a general consensus that ground motions should be calculated using deterministic methods once the controlling earthquakes are determined.
With regard to the role of the probabilistic analysis, views range from an advocacy of a predominantly probabilistic analysis to the probabilistic/ deterministic proposed here to a predominantly deterministic approach.as used currently. Given these divergent views, the NRC staff would like to invite comments regarding the use of probabilistic seismic hazard analysis and the balance between the deterministic and probabilistic evaluations.
This and other associated issues are itemized below. (As the detailed technical studies are completed some of the staff positions may be confirmed, but specific comments would be helpful at this time.)
- 1. In making use of both deterministic and probabilistic evaluations, how should they be combined or weighted, that is, should one i dominate over the other? (The NRC staff feels strongly that deterministic investigation and their use in the development and evaluation of the safe Shutdown Earthquake Ground Motion will remain l an important aspect of the siting regulations for nuclear power plants for the foreseeable future. The NRC staff also feels that probabilistic seismic hazard assessment methodologies have reached a level of maturity to warrant a specific role in siting i regulations.)
- 2. In making use of the probabilistic and deterministic evaluations as proposed in Draft Regulatory Guide DG-1015, is the proposed procedures in Appendix C to DG-1015, adequate to determine controlling earthquakes from the probabilistic analysis?
- 3. In determinig the controlling earthquakes, should the median values of the seismic hazard analysis, as described in Appendix C to Draft Regulatory Guide DG-1015, be used to the exclusion of other statistical measures,such as, mean or '85th percentile? (The staff
j . .
har selected probability of exceedance levels associated with the me11an hazard analysis estimates as they provide more stable estimates of controlling earthquakes.)
- 4. Should the median target level of IE-4 for LLNL or 3&5 for EPRI be raised or lowered, that is, should the next generation of nuclear power plants have design levels for seismic events approximately equal to, greater than, or less than the current nuclear power ;
plants? (The'NRC has a policy statement that the current nuclear .
power plants are at the appropriate level of safety.)
- 5. The proposed Appendix 8 has included a criterion that states: "the probability cf exceeding the Safe Shutdown Earthquake Ground Motion is considered acceptably low if it is less than the median probability cunputed from the current [ EFFECTIVE DATE OF THE REGULATION) population of nuclear power plants". This is a relative criterion without any specific numerical value of the pretability of exceedance. Because of the current status of th probabilistic seismic hazard analysis, method dependent prohbilit'.es or target levels are identified in the proposed requiatory guide. Comments are solicited as to whether the abr/* criterion, as stated, needs to be included in the regulation and, if not, should it be included in the regulation in a dif ferent form (e.g., a specific numerical :
value) . 1
- 6. For the probabilistic analysis, how many controlling earthquakes should be generated to cover the frequency band of concern for nuclear power plants? (For the four trial plants used to develop the criteria presented in Draft Regulatory Guide DG-1015, the average of results for the 5 Hz and 10 Hz spectral velocities was used to establish the probability of exceedance level. Controlling earthquakes were evaluated for this frequency band, for the average of I and 2.5 Hz spectral responses, and for peak ground acceleration.)
DRAFT REGULATORY GUIDE DG-1015 s IDENTIFICATION AND CHARACTERIZATION OF SEISMIC SOURCES, 4 DETERMINISTIC SOURCE EARTHQUAKES AND GROUND MOTION 5
6 7 A. INTRODUCTION 8
9 Paragraph IV (a, b and c) of proposed Appendix 8, " Criteria for the Seismic and 10 Geologic Siting of Nuclear Power Plants after (Effective Date)," to 10 CFR Part 11 100, " Reactor Site Criteria," requires investigations to assess the proposed site 12 for: (a) vibratory ground motion, (b) tectonic surface deformation and (c) non-
.13 ,
tectonic deformation. Paragraph V(a through d) of Proposed Appendix B to 10 CFR 14 Part 100 requires the determination of: (a) deterministic source earthquakes, is (b) site ground motions, (c) safe shutdown earthquake ground motion and (d) the 16 need to design for surface tectonic and non-tectonic deformations.
17 28 The purpose of this guide is to provide general guidance on acceptable procedures 19 to (1) identify and characterize seismic sources. (2) determine deterministic 20 source earthquakes (DSEs) and controlling earthquakes (CEs), and (3) compare the 21 se'essic hazard level to that at operating plants. These procedures are required d
22 h/ Appendix B to 10 CFR Part 100.
23 24 Any information collection activities mentioned in this regulatory guide are 25 contained as requirements in the proposed amendments to 10 CFR Part 50 that would provide the regulatory basis for this guide. The proposed amendments have been submitted to the Office of Management and - Budget for clearance that may be
. appropriate under the Paperwork Reduction Act. Such clearance, if obtained, t
29 would also apply to any information collection activities mentioned in this 30 guide.
31 32 33 8. DISCUSSION 34 35 Appendix 8 requires consideration of both probabilistic and deterministic 36 approaches to obtain site geologic and seismologic characteristics. The approach 37 required by Appendix A to 10 CFR Part 100 for determining the safe shutdown l 38 earthquake ground motion is deterministic and, thus, does not explicitly l 39 incorporate uncertainties about the seismic hazard into the ground motion 40 determination. Current probabilistic seismic hazard analyses rely heavily on 41 expert opinion and their results are driven by the tails of the probability 42 distributions, and, thus, need to be benchmarked by simpler deterministic 43 analysis. Therefore the role of the probabilistic analysis is to ensure that the 44 uncertainties have been included in the assessment of the seismic hazard and the 45 role of the deteministic analysis is to ensure that the resultant design 46 provides protection against a scenario based on historical seismicity and recent 47 geological history.
48' 49 Before providing specific guidance, the following synopsis of the development of 50 the Safe Shutdown Earthquake Ground Motion (SSE) is presented. The development M of the SSE follows two required, parallel paths. The first path is referred to in Figure 1 as Deterministic Analysis (DA) and the second path as Probabilistic DG-1015-1
. . - ._ _ __,__.m._.__ _ _ _ _ _ _ . _ , _ _ _ _ _ _ _
1 Analysis (PA). The' initial step in the process is to obtain the site and region 2 specific geological, seismological, and geophysical data. Branching from the l 3 first step to DA, the seismic sources around the site are identified and the 4 deterministic source earthquake (DSE) for each source is determined. Ground 5 motion is calculated using DSEs and the ground motion guidance provided in 6 Standard Review Plan (SRP) Section 2.5.2. The controlling earthquakes for this J 7 path are determined as illustrated in Figure 2. The initial step along PA is to <
8 conduct an Electric Power Research Institute (EPRI) or a Lawrence Livermore 9 National Laboratory (LLNL) seismic hazard assessment of the site (EPRI-NP-63950 lo and NUREG/CR-5250) for eastern U.S. sites. The results of this assessment are 11 compared to the collected assessments of the currently operating plants as 12 described in Appendix B of this guide. The site seismic hazard assessments are 13
- deaggregated as described in Appendix C of this guide to obtain the controlling
'14 earthquakes for PA. Ground motion based on the controlling earthquakes from PA 15 are also calculated using the guidance in SRP 2.5.2. The ground motions from the l
16 DA and PA controlling earthquakes are compared to the SSE ground motion or are l
17 used to develop the SSE.
18 19 1. Identification and Characterization of Seismic Sources 20 l 21 " Seismic source" is a general term referring to both seismogenic sources and 1 22 capable tectonic sources. A "seismogenic source" is a portion of the earth which 23 is considered to'have uniform seismicity (same DSE and frequency of recurrence).
24 A seismogenic source would not cause surface displacement. Seismogenic sources 25 cover a wide range of possibilities from a well-defined tectonic structure to 26 simply a large region of diffuse seismicity (seismotectonic province). A 27 " capable tectonic source" is a tectonic structure which can generate both j 28 earthquakes and deformation such as faulting or folding at or near the surface j 29 in the present tectonic regime. Appendix A contains definitions of these' and 30 other terms used in this regulatory guide.
31 32 Investigations of the site and region around the site are necessary to identify 33 seismic sources and determine their potential for generating earthquakes and 34 causing surface deformation. Identification and characterization of seismic 35 sources is based on regional and site geological and geophysical data, historical 36 and instrumental seismicity data, .the regional stress field, and geologic 37 evidence for prehistoric earthquakes. The bases for the identification of the
-38 seismic sources should be documented. Appendix D describes investigation 39 procedures that may be used in identifying and defining seismic sources.
40 41 The following is a general list of characteristics to be determined for a seismic 42 source:
43 )
44 'a. Source zone geometry (location and extent, both surface and subsurface).
45 46 b. Description of Quaternary (last 2 million years) displacements (sense of 47 slip on the fault, fault length and width, age of displacements, estimated i 48 displacement per event, estimated magnitude per offset, rupture length and i 49 area, and displacement history or uplift rates of seismogenic folds).
50
- 51- c. Historical and instrumental seismicity associated with each source.
52 DG-1015-2 i
V . ..
1 . _
e .
i
- d. Evidence of paleoseismicity.
, e. Relationship of the fault to other potential seismic sources in the 4 region.
5 6 f. Deterministic Source Earthquake. (Details for the determination of the l 7 DSEs are provided in section 2.) <
8 l 9 g. Recurrence model (frequency of earthquake occurrence versus magnitude). I lo 11 h. Effects of human activities such as withdrawal of fluid from or addition 12 of fluid to the subsurface, extraction of minerals, or the effects of dams
.13 '
or reservoirs.
14 15 1. Volcanism. Volcanic hazard is not addressed in this regulatory guide. It 16 will be considered on a case by case basis in regions where this hazard 17 exists. .
18 19 ,j . Other factors that can contribute to characterization of seismic sources 20 su:h as strike and dip of tectonic structures, orientations of regional 21 and tectonic stresses, fault segmentation (both along strike and down-22 dip),etc. .
?
23 24 The level of detail for investigations around the site is governed by the 25 Quaternary tectonic regime and the geological complexity of the site and region.
" Regional investigations such as geological reconnaissances and litevature reviews should be conducted within a radius of 320 km (200 miles) of the site to identify seismic sources. Geological, seismological, and geophysical investigations 29 should be carried out within a radius of 40 km;(25 miles) to identify and 30 characterize the seismic and surface deformation potential of capable tectonic 31 sources and the seismic potential of seismogenic sources, or demonstrate that ,
32 such structures are not present. Detailed geological, geotechnical, '
33 seismological, and geophysical investigations should be conducted within a radius ,
34 of 8 km (5 miles) of the site to determine the potential for tectonic deformation 35 at or near the ground surface in the site vicinity. Sites that are located such 36 that there are capable and/or seismogenic structures within a radius of 40 km (25 37 miles) will require more extensive geologic' and seismic investigations and 38 analyses (similar ,to those within a 8 km (5 mile) radius). The areas of 39 investigations may be asymmetrical and larger than specified above in areas near 40 capable tectonic sources, high seismicity, or complex geology.
41 42 For the site and the area surrounding the site, the lithologic, stratigraphic, 43 hydrologic and structural geologic conditions will need to be determined. The 44 investigations should include the determination of the static and dynamic 45 engineering properties of the materials underlying the site and an evaluation of 46 physical evidence concerning the behavior during prior earthquakes of the J 47 surficial materials and the substrata underlying the site. The properties needed I 48 to determine the behavior of the underlying material during earthquakes and the !
49 characteristics of the underlying material in transmitting earthquake ground l 50 motions to the foundations of the ' plant (such as seismic wave velocities, 51 density, water content, porosity, elastic modulii, and strength) should be determined. Geological, seismological and geophysical investigations are DG-1015-3
1 described in Appendix 0 to this guide and geotechnical investigations are 2 described in Regulatory Guide 1.132.
3 4 Where it is determined that surface deformation need not be taken into account, 5 sufficient data to clearly justify the determination should be presented.
6 Because engineering solutions cannot always be demonstrated for the effects of 7 permanent ground displacement phenomena, it is prudent to avoid a site when there 8 is potential for surface deformation.
10 ' Eastern United States 11 12 The area east of the Rocky Mountains within the North American Plate and well 13 away from the active plate margins is described as the " stable continental ;
14 region" (SCR). In the SCR characterization of seismic sources is more :
15 problematic than in the active plate margin region because there is. generally no 16 clear association between seismicity and known tectonic structures. The observed 17 geologic structures were generated in response to tectonic forces that no longer is exist and bear little correlation with current tectonic forces. Thus, a greater 19 amount of judgment must be used than for active plate margin regions, and it is 20 important to account for this uncertainty by the use of alternative models.
21 22 Based on current knowledge, seismic sources in the SCR are generally relatively 23 large areas, or seismotectonic provinces. The identification of seismic sources 24, in the SCR should consider hypotheses presently accepted for the occurrence of 25 earthquakes in the SCR (for example, the reactivation of favorably oriented zones 26 of weakness or the local amplification and release of stresses concentrated 27 around a geologic structure).
28
, 29 Western United States i
30 31 For the actit a plate margin region, where earthquakes can often be correlated 32 with tectonic structures, those structures should be assessed for their seismic 33 and surface deformation potential. In the western U.S., at least three types of 34 sources exist: (1) faults that are known at the surface, (2) buried (blind) 35 sources and, (3) subduction zone sources, such as exist in the Pacific Northwest.
36 The nature of surface faults can be determined by conventional surface and near 37 surface investigation techniques to determine strike, geometry, sense of 38 displacements, length of rupture, Quaternary history, etc.
39 40 Buried (blind) faults are often accompanied by coseismic surficial deformation 41 such as folding, uplift or subsidence. The surface expression of blind faulting 42 can be detected by the mapping of uplifted or down-dropped geomorphological 43_ features or stratigraphy, survey leveling and geodetic methods. The nature of 44 the structure at depth can often be determined by core borings and geophysical 45 techniques.
46 47 Subduction zones are seismic sources in the Pacific Northwest and Alaska. The 48 seismic sources associated with subduction zones are the interface between the 49 subducting and overriding lithospheric plates and intrasiab sources in the 50 interior of the downgoing oceanic slab. The characterization of subduction zone
- 51. seismic sources should include consideration of the following: geometry of the 52 subducting plate, rupture segmentation of subduction zones, geometry of DG-1015-4 L
l l
historical ruptures, constraints on the up-dip and down-dip extent of rupture, and comparisons with other subduction zones worldwide. l 3
4 NUREG-XXXX provides a list of references that may be useful in characterizing 5 seismic sources.
6 7 2. Deterministic Source Earthquakes (DSEs) ,
l 8 '
9 DSEs are the largest earthquakes that can reasonably be expected to occur in a 10 given seismic source in the current tectonic regime. Deterministic source 11 earthquakes are characterized by their magnitudes and, as a minimum, will be the l
12 largest historical.egr.thquake associated with each source. A larger earthquake 13
- is warranted in cases where specific geological evidence is available, e.g.,
14 paleoliquefaction evidence of larger prehistoric earthquakes or where the rate is of occurrence of earthquakes indicates the likelihood of larger than the largest 16 historical event.
17 la Eastern United States 19 20 In the SCR there is a short record of the historical seismicity and considerable 21 uncertainty about the underlying causes of earthquakes. Because of this ;
22 uncertainty, it is necessary to use considerable judgment and a variety of 23 approaches to establish the DSEs. In addition to the maximum historical 24 earthquake, the determination of the DSE earthquake for each identified 25 seismogenic source is based on the pattern and rate of seismic activity, the Quaternary (2-million years and younger) development and characteristics of the source, the current stress regime and how it aligns with the known tectonic 28 structures in the source, and paleoseismic data.
29 30 Western United States 31 33 In the Western U.S., earthquakes can often be associated with tectonic 33 structures. For faults, the magnitude of an earthquake is related to the 34 characteristics of the estimated rupture such as the length or the amount of 35 fault displacemer,t. The following empirical correlations can be used to estimate 36 DSE's from fault behavioral data and also to predict the amount of displacement 37 that might be expected for a given magnitude.
38 39 a. Surface rupture length versus magnitude (Slemmons, 1977, 1982; Bonilla and 40 others, 1984; and Wesnousky, 1988).
41 42 b. Subsurface rupture length versus magnitude (Wells and others,1989).
43 44 c. Rupture area versus magnitude (Wyss, 1979).
43 46 d. Maximum and average displacement versus magnitude (Wells and Coppersmith, 47 ir. review).
48 49 In the Pacific Northwest and Alaska, DSE's must be assessed for subduction zone 30 seismic sources. Worldwide observations indicate that the largest earthquakes are associated with the plate interface, although intraslab earthquakes (e.g.,
the 1949 Puget Sound earthquake) can also be large. OSEs for subduction zone
[N3-1015-5
l . .
j l
l 1 sources can be based on estimates of the expected dimensions of rupture or I a analogies to other subduction zones worldwide. l 3
4 NUREG-XXXX contains a list of references, some of which may be useful in l 5 developing maximum earthquakes using deterministic methodologies.
6 :
l 7 3. Probabilistic Seismic Hazard Analysis l i 8 9 A probabilistic seismic hazard analysis (PSHA) should be carried out for the 10 site. A PSHA allows the use of multi-valued models to estimate the likelihood 11 of earthquake ground motions occurring at a site. The PSHA systematically takes la into account uncertainties which exist in various parameters (such as seismic 13 sources, maximum earthquakes, and ground motion attenuation). Alternate ;
14 . hypotheses are considered in a quantitative fashion. The PSHA can be used to l 15 determine the effects of varying significant parameters, identify significant 1 16 sources in terms of magnitude and distance, and provide hazard estimates for use i 17 in seismic probabilistic risk assessments.
18 19 The results of a PSHA are specifically used to derive controlling earthquakes as ,
20 discussed in Section 4 below and Appendix C. It can also be used to estimate the '
31 probability of exceeding the SSE and demonstrate that the probability of 22 exceeding the SSE design ground motion at the site compares favorably with that
, 23 for the currently operating nuclear power plants. (The procedure for this l 24 demonstration is described in Appendix B.)
25 26 Either the Lawrence Livermore National Laboratory (LLNL) (NUREG/CR-5250) or 27 Electric Power Research Institute (EPRI) (EPRI-NP-6395-D) seismic hazard 28 analyses, including associated data bases, should be used for plant sites in the 29 SCR. However, alternative seismic hazard analyses may be used with proper 30 justification. For the PSHA, the use of the seismic sources identified in the 31 LLNL and EPRI studies are considered acceptable except in regions of the SCR with 32 high activity rates, e.g., near New Madrid and Charleston. In these cases, 33 either describe additional site specific seismic sources or show that the 34 regional seismic sources in the LLNL and EPRI probabilistic studies adequately 35 model the tectonics in the vicinity of the site.
36 37 Probabilistic methodologies similar to the LLNL and EPRI seismic hazard studies 38 have not been performed for the western U.S. For western U.S. sites, a site 39 specific PSHA must be performed and documented in such detail that a thorough 40 review can be carried out by the NRC staff (PG&E,1988; NUREG-0675; WPPSS,1988).
41 42 4. Controlling Earthquakes 43 44 Controlling earthquakes are those earthquakes that have the greatest effect on 45 the ground motion at the nuclear power plant site. There may be several 46 controlling earthquakes for a site, e.g., a moderate, nearby earthquake may 47 control the high frequency portion of the ground motion spectrum and a large, 48 distant earthquake may control the low frequency portion of the spectrum. See i 49 Figure 2.
50 51 In the Deterministic Analysis (Figure 1.), the controlling earthquakes are
! 52 determined via the following procedure.
DG-1015-6
( s ...
c .e
- a. 'For each seismic source, place the DSE at the closest approach of that source to the site. For the seismic source in which the site is located, i the DSE should be considered to occur at about 15 km from the site.
4 5~ b. Determine the DSEs that produce the largest ground motions at the site, 6 Ground motions at the site from DSEs are estimated using the procedures 7 described in. Standard Review Plan Section 2.5.2 (Vibratory Ground Motion).
8 The earthquakes producing the largest ground motions at the site are the 9 controlling earthquakes. ,
lo l 11 In the Probabilistic Analysis (Figure 1), the controlling earthquakes are I 12 determined via the following procedure.
13 14
- a. Perform a probabilistic seismic hazard analysis for the site. The is analysis will develop uniform hazard spectra at several probabilities of 16 exceedance. l 17 l
18 b. Deaggregate the probabilistic seismic hazard results to identi fy 19 controlling earthquakes; their description includes magnitude and distance 20 from the site (Appendix C). This deaggregation should be done at the 21~ probability of exceedance level discussed in Appendix B.
22 23 )
24 The controlling earthquakes thus derived from the deterministic and probabilistic 25 analyses can be compared at this stage to determine if the controlling '
x earthquakes from these two approaches are similar and also to determine if the controlling earthquake (s) which will dominate the ground motion estimates at the site is (are) easily identifiable. If the dominant controlling earthquake (s) can 29 be identified, the ground motions are determined only for this identified 30 controlling earthquake (s). If the controlling earthquakes from the two 31 approaches are dissimilar, then ground motion estimates are made for various 32 controlling earthquakes and compared to derive the final ground motion estimates 33 for use in establishing the SSE ground motion or comparing it with the SSE ground 34 motion.
35 O
l DG-1015-7
1 C. REGULATORY POSITION 2
3 1. During the site selection phase, preferred sites are those where there is 4 a minimum likelihood of surface or near surface deformation or the 5
occurrence of earthquakes on faults in the site vicinity (within a radius 6 of 8 km (5 miles)). Because of the uncertainties and difficulties in 7 mitigating the effects of permanent ground displacement phenomena such as 8 surface faulting or folding, fault creep, subsidence or collapse, the NRC 9
staff considers it prudent to select an alternate site when the potential 10 for permanent ground displacement exists at the site. '
11
- 2. Regional investigations such as geological r'econnaissances and la 13 . literature reviews should be conducted within a radius of 320 km (200 14 miles) of the site to identify seismic sources.
15 16 3. Geological, seismological, and geophysical investigation should be carried 17 out within a radius of 40 km (25 miles) to identify and characterize the 18 seismic potential of capable tectonic and seismogenic sources or 19 demonstrate that such structures are not present.
20 21 4. Detailed geological, geotechnical, seismological, and geophysical 22 investigations should be conducted within a radius of 8 km (5 miles) of 23 the site to determine the potential for tectonic deformation at or near 24 the ground surface in the site vicinity. Geological, seismological and ,
~
25 geophysical investigations are described in Appendix 0 and geotechnical 26 investigations are described in Regulatory Guide 1.!32.
27 28 5. Sites that are located such that there are capable and/or seismogenic 29 faults within a radius of 40 km (25 miles) will require more extensive 30 geologic and seismic investigations and analyses (similar to those within a 8 km (5 mile) radius). The area of investigation may be asymmetrical 31 32 and extend beyond 40 km (25 miles).
33 34 6. Seismic sources should be identified and characterized using the
! 35 information developed by the investigations. Alternative seismic sources 36 should be developed to incorporate a range of interpretations and the 37 bases for the identification of these sources should be documented.
38 Source zone geometry should be defined for each seismic source. For 39 faults, the type of slip, length of rupture, amount of displacement per 40 maximum event, and area of the rupture surface should be determined.
41 43 7. Deterministic Source Earthquakes, which are the best judgment of the 43 maximum earthquake that can reasonably be expected to occur in a given 44 seismic source should be defined for each seismic source.
45 46 8. Perform a probabilistic seismic hazard analysis (PSHA) for the site to
- 47 estimate the probability of exceeding the SSE. Either the LLNL or EPRI 48 probabilistic seismic hazard analyses with associated data bases should be 49 used for plants in the eastern United States. For western plants, a site-50 specific probabilistic seismic hazard study should be performed. Use the 51 PSHA to identify sources in terms of magnitude and distance that 52 contribute significantly to the seismic hazard at the site.
DG-1015-8 l
l
- - - - - - - . . . - . . . - .-- ~. .. . . . - - . - . . . - - .
- 9. Determine the Ces that produce the largest ground motions at the site.
Ground motions at the site from CE's are estimated using the procedures 3 described in Section 4 of this guide and Standard Review Plan Section 4 2.5.2 (Vibratory Ground Motion).
5 6
7 D. IMPLEMENTATION 8
9 The purpose of this section is_ to provide guidance to applicants and licensees 10 regarding the NRC staff's plans for using this regulatory guide.
11 12 This draft guide has been released to encourage public participation in its is development. Except in those cases in which the applicant proposes an acceptable 14 alternative method for complying with the specified portions of the Commission's 15 regulations, the method to be described in the active guide reflecting public 16 comments will be used in the evaluation of applications for a construction 17 permit, operating license, early site permit, or combined license submitted after 18 the implementation date to be specified in the active guide. This guide would 19 not be used in the evaluation of an application for an operating license 6 20 submitted after the implementation date to be specified in the active guide if 21 the construction permit was issued prior to that date.
22 ,
23, 24 REFERENCES 25 Appendix B to 10 CFR Part 100, Criteria for the Seismic and Geologic Siting of 28 Nuclear Power Plants After [ Effective Date).
29 30 Bonilla, M.G., H.A. Villabobos, and R.E. Wallace,1984, Exploratory Trench Across 31 the Pleasant Valley Fault, Nevada; Professional Paper 1274-B, USGS, p B1-814.
32 33 Cornell, A.C and E.H. Vanmarcke,1969, The Major Influence on Seismic Risk; 34 Proceedings of the Fourth World Conference on Earthquake Engineering, Santiago, 35 Chile, v. 1, p 69-83.
36 37 Electric Power Research Institute Report NP-6395-0,1989, Probabilistic Seismic 38 Hazard Evaluations at Nuclear Power Plant Sites in the Central and Eastern United 39 States: Resolution of the Charleston Earthquake Issue.
40 41 Gutenberg, B. and C.F. Richter,1954, Seismicity of the Earth and Associated 42 Phenomena; Second Edition, Princeton; Princeton University Press, 310 p.
43 44 NUREG/CR-5250, 1989 Seismic Hazard Characterization of 69 Nuclear Plant Sites 45 East of the Rocky Mountains.
46 47 NUREG-0675, Supplement No. 34, 1991, Safety Evaluation Report related to the 48 operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2.
49 50 NUREG-XXXX, Supplementary list of references for Draft Regulatory Guide OG-1015.
Pacific Gas and Electric Company,1988. Final Report of the Diablo Canyon Long ,
DG-1015-9 ,
s
1 Term Seismic Program; Diablo Canyon Power Plant, Docket Nos. 50-275 and 50-323.
2 3 Regulatory Guide 1.132, " Site Investigations for Foundations of Nuclear Power 4 Plants."
5 6 Schwartz, D.P. and K.J. Coppersmith,1984, Fault Behavior and Characteristic 7 Earthquakes: Examples from the Wasatch and San Andreas Fault Zones; Journal 8 Geophys. Res., v. 89, p. 5681-5698.
9 10 -Slemmons, D.B., 1977, Faults and Earthquake Magnitude, U.S. Army Corps of 11 Engineers, Waterways Experiment Station, Misc. Papers 5-73-1, Report 6.
12 13- Slemmons, D.B., 1982, Determination of Design Earthquake Magnitudes for 14
- Microzonation; Proc. Third International Microzonation Conference, v.1, p 119-15 130.
16 17 Wells, D.L., and K.J. Coppersmith, Updated Empirical Relationships Among 18 Magnitude, Rupture Length, Rupture Area,- and Surface Displacement; Bulletin of 19 the Seismological Society of America (in review).
20 21 Wells, D.L. , K.J. Coppersmith, X. Zhang, and D.B. Slemmons,1989, New Earthquake 22 Magnitude and Fault Rupture Parameters: Part II. Maximum and Average 23 Relationships (Abs): Seismological Research Letters, v. 60, n.l.
24 25 Wesnousky, S.G.,1988, Relationship Between Total Affect, Degree of Fault Trace 26 Complexity, and Earthquake Size on Major Strike-Slip Faults in California; (abs).
27 Seismological Research Letters, v. 59, no. 1, p. 3.
28 29 Wyss, M.,1979. Estimating Maximum Expectable Magnitude of Earthquakes from Fault 30 Dimensions; Geology, v. 7 (7), p. 336-340.
31 32 WPPSS, 1988, February 29, 1988 letter from G. Sorensen, Washington Public Power 33 Supply System to U.S.NRC.
Subject:
Nuclear Project No. 3, Resolution of Key 34 Licensing Issues, Response to Question on Seismic Hazard.
O I
DG-1015-10 l
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- 0:0:0'0008800% %000%%% '0%0MM%G:0%G 000%..< ..o i
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F t PROBABILISTIC' SITE DETERMINISTIC p ANALYSIS (PA)
= _ ANALYSIS- (DA)
- g. Geological, Selsmological 3 and GeophysicalInvestigations
,, i e
' i *
- a. Conduct an EPRI or LLNL
- Identify Seismic Seismic Hazard Assessment Sources ;
S 7. Compare to Operating ~
Y $ Determine Deterministic Source e Plants to Set Probability Earthquakes for Each Source E of Exceedance Level
'?
vs p Determine Controlling Earthquakes (CEs) Ms & Ds Determine Controlling Earthquakes (CEs) Ms & Ds
, i i E ,
i a
Compara CEs Derived From PA and DA Z
g Develop SSE Ground Motion c _( GM) or Compare with CE GMs
I 1
2 3 u,.
4 s aa 5
6 I l S
9 10 k . .<
11 ss 12-13 0 " "
14
- 15 -
16 g ,,
17 8=== 1 18 19 ,
20 l
21 22 unsu< l 23 24 25 26 27 1 28 29 ' I 3 30 %"T2.
31 32 -
- l 33 .
34 g
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l 35 / .
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40 l '
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l 44 c.oi o,i 3 io e.a.ei..ca 47 48 49 50 Figure 2. Schematic representation of the detemination of the controlling 51 earthquakes for the deterministic analysis path.
4 DG-1015-12 l
.- . o l 5
t Anoendix A to Reaulatory Guide DG-1015 3
4 Definitions 4 5
6 7
8 Seismic Source ,
9 10 A " seismic source" is a general term referring to both seismogenic sources and.
11- capable tectonic sources.
j 12-13 . Seismonenic Source 14 15' A "seismogenic source" is a portion of the earth that has uniform earthquake
, 16 potential (same expected maximum earthquake and frequency of recurrence) distinct ,
l 17 from the surrounding area. A seismogenic source will not cause surface '
f 18 displacement. Seismogenic sources cover a wide range of possibilities from a 19 wet defined' tectonic structure to simply a large region of diffuse seismicity 20 (s01smotectonic province) thought to be characterized by the same earthquake
- 21. ret erreace model. A seismogenic source is also characterized by its involvement 22 in the current tectonic regime as reflected in the Quaternary (approximately the 23 .last ? L.' lion years).
l 24 25 Canable' Tectonic Source I A " capable tectonic source" is a tectonic structure which can generate both
! .2. earthquakes and tectonic surface deformation such as faulting or folding at or 1
~29 near the surface in the present seismotectonic regime. It is characterized by l 30 at.least one of the following characteristics:
31 32 a. Presence of surface or near surface deformation of landforms or geologic 33 deposits of a recurring nature within the last approximately 500,000 years 34 or at least once in the last approximately 50,000 years.
! 35 36 b. A reasonable association with one or more large earthquakes or sustained 37 earthquake activity which are usually accompanied by significant surface l -38 deformation.
39 40 c. A structural association with a capable tectonic source having I
41' characteristics'(a) of this paragraph such that movement on one could be 42 reasonably expected to be accompanied by movement on the other.
43 44 In some cases, .the geologic evidence of past activity at or near the ground 45 surface along a particular capable tectonic source may be obscured at a 46 particular site. This - might occur, .for example, at a site having a deep 47 overburden. For these cases, evidence may exist elsewhere along the structure 48 from which an evaluation of its characteristics in the vicinity of the site can 49; be reasonably based. Such evidence shall be used in determining whether the 50 structure is a capable tectonic source within this definition.
=$
l Notwithstanding the foregoing paragraphs, structural association of a structure t
l DG-1015-13
. . 1 1 with geologic structural features which are geologically old (at least pre-2 Quaternary) such as many of those found in the eastern region of the lkiited 3 States shall, in the absence of conflicting evidence, demonstrate that the 4 structure is not a capable tectonic source within this definition.
5 6
7 Deterministic Source Earthouake (DSE)'
8 9 A DSE is the largest earthquake that can reasonably be expected to occur in a 10 given seismic source in the current tectonic regime, and is used in a 11 deterministic analysis. It is generally based on the maximum historical 12 earthquake associated with that seismic source, unless recent geological evidence 13 warrants a larger earthquake, or where the rate of occurrence of earthquakes 14 indicates the likelihood of larger than the largest historical event.
15 16 17 Controllina Earthauakes (CE) 18 2; Controlling Earthquakes are the earthquakes which produce the largest ground 24 motions estimated at the site. There may be several Ces for a site.
21 22 Stable Continental Recion 23 24 A " stable continental region" (SCR) is comprised of continental crust, including 25 continental shelves, slopes and attenuated continental crust and excludes active 26 plate boundaries and zones of currently active tectonics directly influenced by 27 plate margin processes. It exhibits no significant deformation associated with 28 the major Mesozoic-to-Cenozoic (last 240 million years) orogenic belts. It 29 excludes major zones of Neogene (last 25 million years) rifting, volcanism or 30 suturing.
31 32 Safe Shutdown Earthauake 33 34 The Safe Shutdown Earthquake Ground Motion is the vibratory ground motion for
! 35 which certain structures, systems, and components shall be designed to remain 36 functional.
37 38 Intensity 39 40 The intensity of an earthquake is a measure of its effects on humans, human-built 41 structures, and on the earth's surface at a particular location. Intensity is 42 described by a numerical value on the Modified Mercalli scale.
43 44 Tectonic Structure 45 46 A tectonic structure is a large-scale dislocation or distortion usually within 47 the earth's crust. Its extent is on the order of miles.
48 49 Maonitude i 50 51 An earthquake magnitude is a measure of the strength of an earthquake as 52 determined by seismographic observations.
DG-1015-14 l
l
e .
Nontectonic Deformation 3
4 Nontectonic deformation is distortion of surface or near surface soils or ro.;ks 5 that is not directly attributable to tectonic activity. Such deformation 6 includes features associated with subsidence, karst terrane, glaciation or 7 deglaciation, and growth faulting.
8 e
f 1
DG-1015-15
. _ .m _ _ . - _ _.. .- _.._ _ _ _._ __ _ _ _____.___ ______
1
-2 ADDendix B to Reculatory Guide DG - 1015 -
3 4 Probabilistic Comoarison of Safe Shutdown Earthouake 5 to Ooeratino Plants 6
7 8 B.1 Introduction i 9
- 10 This appendix outlines a procedure to calculate the probability of exceeding the 11 Safe Shutdown Earthquake Ground Motion (SSE). This procedure can be used (1) to 12 compare the calculated probability of exceeding the SSE to those for the 13 currently operating plants as required by Appendix B to 10 CFR Part 100; and (2) 14 - to establish controlling earthquakes in the probabilistic hazard analysis as 15 discussed in Appendix C to this regulatory guide. Uniform hazard spectra 16 (spectra that have a uniform probability of exceedance over the frequency range 17 of interest) should be calculated to estimate the probability of exceeding the 18 SSE design response spectrum.
19 20 B.2 Procedure
- 21 22 The following procedure is one acceptable approach to assure that the probability 23 of exceeding the SSE compares favorably with that for the currently operating 24 nuclear power plants as of [date).
25 LL1 Eastern U.S. Sites.
26 27
- 28 There are two state-of-the-art approaches (EPRI NP-6395-D,1989 and NUREG/CR-29 5250, 1989) currently available to calculate the probabilistic seismic hazard for 30 sites east of the Rocky mountains (Eastern U.S.). These approaches, however, 31 produce different hazard estimates fu a given site. Therefore, the staff is 32 recommending the following interim procedure until the differences between the 33 two hazard methods are resolved. This procedure relies on relative measures to i 34 assure that the annual probability cf exceeding the SSE is comparable to that of !
' ' 35 operating plants. The procedure is based on. studies conducted for the Eastern 36 Seismicity Issue and the IPEEE program (NUREG-1407,-1990). Either the LLNL or 37 EPRI methodology can be used to carry out the following calculations, with the 38 appropriate. set of limits associated with each method. If any analysis other ,
than the LLNL or EPRI methods is used in the eastern U.S., probabilities of 39 l 40 exceeding the SSE would need to be developed for all operatir.; plant sites in l 41 addition to the site under consideration in order to make the appropriate l 42 comparison.
43 44 Step 1. Calculate Uniform Hazard Response Spectra (UHRS) with various return 45 periods. Figure B.1 shows a sample set of median UHRS for various 46 return periods. The UHRS should be developed at the same location as the location of the SSE (i'.e. either at the free ground surface 47 or at a hypothetical rock outcrop).
49-50 Step 2. Calculate composite annual probabilities of exceeding the SSE and 51 compare those probabilities with operating plants using median 52 hazard estimates. (Although the median estimates are used for the DG-1015-16
e e l
l purpose of the carrying out the procedure outlined in this appendix, l the hazard analysis should be performed with consideration of 3 uncertainties to develop complete insights.) The procedure is 4 illustrated in Figure B.2.
5 6 (a) Estimate the annual probabilities of exceeding the SSE l 7 spectrum at two discrete frequencies (5 and 10 Hz) using the 8 UHRS.
9 10 (b) Calculate the composite annual probability using the following 11 formula:
, 12 13 . Composite Probability = 1/2(al) + 1/2(a2) l 14 15 where al and a2 represent annual probabilities of exceeding 16 SSE spectral ordinates at 5 and 10 Hz, respectively.
17 18 Examole: From Figure B.2, for a median UHRS derived using the 19 LLNL methodology, at points al and a2 corresponding to 5 and i 20 10 Hz.
21 l 22 Composite Probability = 1/2(4E-5) + 1/2(8E-5) 23 - 6E-5. -
24 I 25 (c) Figure B.3 shows the distribution of median probabilities of exceeding SSEs for operating Eastern U.S. plants using LLNL hazard estimates. This figure also indicates a limit;
.o approximately 50% of the currently operating plants have a 29 probability of exceeding the SSE ground motion below this 30 limit. (Limits for both the current EPRI and LLNL seismic 31 hazard studies are listed in Table B.l.) The SSE is adequate 32 when the probability of exceeding the SSE compares favorably l 33 to the limits shown in these figures.
34 l 35 Table 8.1 !
36 1 37 38 Method Probability of Exceedance Limits for Median Hazard Estimates 39 LLNL lE-4 40 EPRI 3E-5 41 42 43 For the hypothetical example the calculated probability of
, 44 exceedance of 6E-5 is less than the limit of IE-4 and thus the 45 probability of exceeding the SSE compares favorably with that 46 of operating plants.
47 EH3-1015-17
._ _._ __ .m_ ~ . _ _ _ _ _ _ _ _ _ . _ _ _ _ ~ . _ _ _
1 Figures B.4 presents the same information resulting from the 2 use of the EPRI UHRS estimates. This limit should be used 3 when the EPRI method is used to calculate the probability of 4 exceeding the SSE.
5 6
7 L L 2 Western U.S. Sites 8
9 For the Western U.S. (WUS) sites, a probabilistic data base, such as that 10 compiled in the LLNL and EPRI studies, is not available. To date no procedure 11 exists, similar to that described above, to compare the probability of exceeding 12 the SSE to other sites in the WUS. In addition, the probaLilistic hazard at a 13 site in the WUS may be governed by clearly identifiable seism c sources, such as 14 . faults (or folds) observed at the surface, which have better defined seismicity is characteristics. Therefore, for WUS sites, a site-specific analysis should be 16 developed using suitable methodologies to estimate the probability of exceeding 17 the SSE and to identify significant contributors to the hazard (e.g., NUREG-0675, 18 1991).
19 20 21 REFERENCES 22 23 Electric Power Research Institute Report NP-6395-D, "Probabilistic Seismic Hazard 24 Evaluations at Nuclear Power Plant Sites in the Central and Eastern United 25 States: Resciution of the Charleston Earthquake Issue," 1989.
26 27 NUREG/CR-5250, " Seismic Hazard Characterization of 69 Nuclear Plant Sites East 28 of the Rocky Mountains," 1989.
29-30 NUREG-1407, " Procedural and Submittal Guidance for the Individual Pl ant 31 Examination of External Events (IPEEE) for Severe Accident Vulnerabilities,"
32 1990.
33 34 NUREG-0675, Supplement No. 34, " Safety Evaluation Report related to the operation 35 of Diablo Canyon Nuclear Power Plant, Units 1 and 2," 1991. -
36 37 DG-1015-18
. . l l
ed 4
5 6
7 8
9 10 El L2
'13 '
3 - 5 14 go ,
15 16 17 18 19 .
20 88 2
~
31 32 23 I
24 25 '
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to' -
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29 t
y j
30 g "
t 31 l 32 * .
e 33 -
to stivant PCatoos :
34 cunvt 5 = soooo. YCAsts 35 cumvC 4 Moo, vt#ts 36 W 3* 2000
- 1 37 cuevC 2 = som. W5 l
38 CURvC 1 = S00. WMts n ~. ~* =
39 ., '
n ~~=a=
40 to , , . . _ -
- To To f0tt00 (SCC) 42 43 44 Fig. B.1 Median Uniform Hazard Response Spectra 4 .
47 DG-1015-19 l
l ..
1 2 3 to 3
4 5
6 7
8 , .
9 to 10 11 12 w w nsc w s 13 o A 1 E-5 14 q .5 - A. 2E-5
--- 3E-5 15 16 6
g to'
[
/ - - 1 E 1 j 2E 4 19 y 20 21 e 22 80 23 24 25.
26 27 go' a a - " ~ " " * * ~
28 a - = = =
29 7e Io PCRico (SCC) O *
~~
30 -
31 32 33 Figure B.2 Procedure to Compute Probability of Exceeding 34 ' Design Basis 35 36 37 Comp. Prob. - 1/2(al) + 1/2(a2) 38 39 DG-1015-20 mi . . _ _ _ . _
O' * * -
- l 1
4 5
6 +
7' '
8 9
10 11 12 1;c - - - - i - - - - i
- - - - y.
. 33 .
1 ordered using med. log. haz, :
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18
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20 m : "
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41 -
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- e e 43 "
e 44 45 46 1 47 Probability of Exceeding SSE 48 4',9 Figure 8.3 Probability of Exceeding SSE Using Median LLNL Hazard 50 Estimates 51 DG-1015-21
9..
1 2
3 4
5 6
7 8
9 10 11 1. 0 _
12 - -
13 : ordered using med. log. haz :
14 - .9 .- --
15 -
C 16 o .- .
17 ~- .8 - ~
18 I 19 o : :
2 0 'C 7 .- ""
21 m - * .
22 23 5 0 .~ ~
24 e- -
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25 * -
26 .5 7 3E-5 i 27 C - -
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30 : :
31 - -
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33 :
34 - -
35 * ~ - T 36 : :
37 *g 38 -
39 :
40 0*
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41 e- = * * " " *
- e e a e 42 w w w w 43 44 Probability of Exceeding SSE 45 46 Figure B.4 Probability of Exceeding SSE Using Nedian 47 EPRI Hazard Estimates DG-1015-22
e e Accendix C to Reculatory Guide OG-1015 3
4 5 Determination of Controllina Earthauakes from the Probabilistic Analysis 6
7 8
9 10 C.1 Introduction 11.
-12 This appendix outlines a procedure to determine controlling earthquake (s) from 13 - the probabilistic hazards analysis for a site. The ground motions from these 14 controlling earthquakes should be determined following the procedures outlined Controlling earthquakes should 15 in Section 2.5.2 of the Standard Review Plan.
16 be determined for the median seismic hazard limit used to satisfy the requirement 17 discussed in Section C.2 below and Appendix B of this Regulatory Guide to demon-18 strate that the probability of exceeding the saf r shutdown earthquake ground 19 motion (SSE) compares favorably with that of the currently operating nuclear 20 power plants.
21 22 C.2 Procedure 23 controlling 24 , The following procedure is one acceptable approach to determ ine 25 earthquakes from an probabilistic hazards analysis.
C.2.1 Eastern U. S. Sites 28-29 As discussed in Appendix B of this Regulatory Guide there' are two approaches 30 (NUREG/CR-5250,1989 and EPRI NP-6395-D,1989) currently available to calculate 31 probabilistic seismic hazards 'for sites east of the Rocky mountains (Eastern 32 U.S.). Either of these methods can be used to carry out the following 33 calculations, with the approprirte set of limits associated with each method.
34 35 Step 1. Perform the site-specific hazard analysis using the LLNL or EPRI 36 method and associated data. From this analysis, compute median 37 hazard curves for the average of the 5 and 10 Hz spectral 38 velocities , S,..i.. That is a curve showing probability of exceeding 39 various levels of the average of the 5 and 10 Hz spectral velocity.
40 41 Step 2. Using the appropriate probability of exceedance level, Pc, (e.g.,
42 for the median 5,,,,, hazard curve derived from the LLNL method, P c is 43 1E-4 according to Figure B.3(c) and Table B.1 of Appendix B), enter 44 the hazard curve of step 1 at P, to determine the corresponding 45 spectral velocity.
46 47 Step 3. Deaggregate the median of the average of the 5 and 10 Hz hazard 48 curves as a function of magnitude and distance by calculating the 49 contribution to the hazard for all of the earthquakes in a selected 50 set of magnitude and distance bins, to determine the relative
"' contribution to the hazard, K., for each bin centered at Magnitude e and Distance d. H., is the probability of exceeding S,(P,)
D3-1015-23
. -- - .- . . , ~ _ - . . . - - -.-.-. ... - - . . . . - - . . . . - -
o . v J
i i computed for a bin at, magnitude m and distance d.
2 3 ' Step 4. Compute the magnitude of the controlling earthquake for the median 4 . estimste using the contributions H computed in Step 3.
i 5 i 6^
k' 7 l 1
8-
'd8 R = Z Z mH / Z Z H.
11 md md
.j. 13 .
i 13 '
14 , -The distance of the controlling earthquake from..the site- is l
! 15 determined from i
! 16 i j' 17 L M D - md I I dH./ad II H.
f 20 1 21 1 32 ~ Step 5. Using the same P, and steps 1 through 4 as above, also determine ,
! 23 controlling earthquakes for median spectral response for the average ;
4
-24 of the I and 2.5 Hz spectral responses, and for the median estimates 25 of the peak ground acceleration. ]
'26 '
37 28 Step ti. The ground motion corresponding to the controlling earthquake is i 39 determined as outlined in Section 2.5.2 of the Standard Review Plan.
30 31 l 32 C.2.2 Western U. S. Sites 33 34 For the Western U. S. Sites, a probabilistic data base, such as compiled in the 35 LLNL or EPRI studies, is not available. In a region of active tectonics there 36 is less uncertainty about the significant contributors to the seismic hazard and 37 the controlling earthquakes can generally be defined deterministically. For 38 regions of lower, less ar.tive tectonics, an analysis similar to the one outlined 39 above in Steps 1-4 can be performed. Step I would be omitted and the 5, level 40 used would'corrcspond to the value selected for the SSE.
41 42 43 C.3 Examole for Eastern U. S. Site 44
,45 To illustrate the application of the above procedure, calculations are performed 46 for an eastern U. S. site using the LLNL methodology given in NUREG/CR-5250, 47 48 Step 2 49 90 Table C.1 gives the probability of exceeding various levels of the average of the 51 5 and 10 Hz spectral velocity hazard curves from the LLNL study.
92 93 DG-1015-24
Table C,1 a Average of 5 and 10 Hz S, Curves for the Site 4
Spectral Probability of Exceedance 5
6 Velocity (Median) 7 ( S,-cm/ s) 8 2 .
2.6E-3 9 5 3.7E-4 10 5.8E-5 10 ,
11 12 Entering Table C.1 with the probabirity of exceedance (P,) values given in Table 13 8.1, and by interpolating, the corresponding value for S,(P,) is as given in 14 Table C.2.
15 16 17 18 Table C.2 i 19 20- Median 2' S,(P,)-cm/s 8 24 25 Step 3 26 27 For this example, to deaggregate the hazard and determine the H, , it is first 28 necessary to compute the contribution to the average hazard for the 5 and 10 Hz 29 spectral velocities for the matrix of magnitudes and distance bins such as given 30 in Table C.3.
31
[N3-1015-25
- . - . = - - - . . - . . ..- _- -
Q 9 1
2 3 Table C.3 4 ,
9 Magnitudes and Distance Bins Used in Example 6 )
i l
7 Distance Magnitude Range of Bin 8 Range of i 9 Bin (km) 5 - 5.5 5.5 - 6 6 - 6.5 6.5 - 7 7 - 7.5 >7.5 10 0-25 l
11 25-50 i
12 50-100 i 1
13 100-150 14 150-200 15 >200 16 17 For each bin a complete hazard analysis is performed to give the contribution to 18 the hazard from all earthquakes within the bin, e.g., all earthquakes with 19 magnitudes 6 to 6.5 and distance 25 to 50 km from the site. The results for this 20 bin are given in Table C.4. ,
21 22 23 Table C.4 24 25 Contribution to the Hazard from All Earthquakes in the Range of 26 6 s M s 6.5 and distances 25 s d s 50 to the average of the 5 27 and 10 Hz spectral velocity 28 29 Spectral Median 30 Velocity, S, Probability of Exceedance 31 5 1.4E-5 32 10 3.lE-6 33 12.5 1.lE-6 34 35 The value of H. (Probability of exceeding S,(P,)) for this bin is obtained by 36 entering Table C.4 with the S,(P,) values given in Table C.2 and computing H. by 37 interpolation. The values for H for this bin are given in Table C.5.
38 i
l I
! DG-1015-26 l
l.
l J Table C.5 ,
4 l 5 Value for H for the bin 6 s m s 6.5 and !
6 25 s d s 50 for the Example Site j 7
8 Median 9 H. 5.0E-6 10 11 12 -
Table C.6 gives the complete matrix of the H. values for the example site.
13 14 15 16 Table C.6 17 l 18 H,,, Values for All Bins Based on the Median Hazard 19 (Note: If H.-s 1.E-10, it is listed as 0) 20 l
l- 21 Distance Magnitude Range of Bin 22 Range Bin 5 - 5.5 5.5 - 6 6 - 6.5 6.5 - 7 7 -7.5 >7.5 ,
! 0-25 2.0E-5 1.lE-5 2.4E-6 0 0 0 l 24 25-50 6.2E-6 8.9E-6 5.0E-6 6.5E-9 0 0 i
- 25 50-100 6.0E-7 2.3E-6 6.8E-6 8.4E-7 0 0 26 100-150 1.6E-9 1.6E-7 1.5E-6 2.8E-6 0 0 27 '150-200 0 1.lE-9 2.lE-8 4.6E-7 0 0 28 >200 0 0 0 6.0E-9 0 0 l 29 30 l 31 Step 4 33 To compute R, D for the example site, the values of H. given in Table C.6 are 34 used with m and d values corresponding to the midpoint of the magnitude of the 35 bin (5.25, 5.75, 6.25, 6.75, 7.25, 7.75) and centroid of the ring area (16.7, 36 38.9, 77.8, 126, 176 and somewhat arbitrarily 300km).
37 30 Thus for the example site, the controlling earthquakes, in if, 6 values are given 40 in Table C.7.
41 DG-1015-27
e e
.k I 2
Table C.7 l
) 3 :
4 Magnitude and Distance of Controlling Earthquake from the t 5
LLNL Probabilistic Analysis i
6 7
Based on Median
. Hazard Estimates 5.8 .
J g g
- 7.0 g 32 41 j .g 13 .
14 t
' 15
- 16 C.4 Examoles for Western U. S. Sites I
17 18 Sir.ce a general approach for the western U.S. sites is not available, two 19 specific cases illustrating determination of controlling earthquakes are .
20 discus:cd below.
21 22 i 23 C.4.1 - Diablo Canyon l
- 24 !
25 The Diablo Canyon site is located on the California coast. A logic-tree approach 26 has been used to assign weights to variables associated with faults near the site 27 and determine maximum magnitude distributions (NUREG-0675, Supplement 34). The 28 logic tree app ich was also part of the probabilistic seismic hazard analysis. 1' 29 The result was that-the Hosgri fault zone was the most significant source. The 30 controlling earthquake for the Diablo Canyon site is a magnitude 7.2 event on the
- 31. Hosgri fault zone at the closest distance of this fault zone to the site (4.5 l '
32 km). The controlling earthquake magnitude is larger than the maximum historical 33 . earthquake (the 1927 magnitude 7.0 Lompoc earthquake) which may have occurred on 34 a structure related to the Hosgri.
35 36 C.4.2 - WNP-3 ,
37 i 38 The WNP-3 site is located in western Washington and lies above the Cascadia 39 subduction zone. The staff considered four controlling earthquakes for the site 40 (January 4, 1991 letter from Mendonca to Mazur).
41 42 a. The applicant proposed that a maximum random earthquake in the crust near 43 the site is magnitude 5-1/2 to 6. . This earthquake is based on the largest :
44 historical ecrthquakes in the Coastal Plain seismotectonic province (about ,
45 magnitude 5) and the resolution of geological studies in the site region.
- 46 47 b. The maximum earthquake associated with the Olympia Lineament 35 km 48 northeast of the site is a magnitude 7.5 based on estimated maximum ,
'49 rupture length.
50 51 c. The maximum magnitude earthquake for the intraslab subduction zor.e source DG-1015-28
' ~
i I
- o. . ,
-f F is about magnitude 7-1/2 based on the maximum historical event associated i
- with the Cascadia subduction zone intraslab source (the 1949 magnitude 7.1 3 Puget Sound earthquake) and comparisons with intraslab sources in other -
, 4 subduction zones worldwide.
5 -
6 d. The interface subduction zone source is capable of great (larger than 7' magnitude 8) earthquakes. This maximum magnitude is still under review in a light of ongoing geological studies. At this time the staff considers the 9 maximum magnitude to be 8-1/4 based on arguments about the likely lo dimensions of rupture and comparisons with other subduction zones with 11 slow convergence rates.
12 13 . REFERENCES ,
14 >
15 Electric Power Research Institute Report NP-6395-D, "Probabilistic Seismic Hazard 16 Evaluations at Nuclear Power Plant Sites in the Central and Eastern United 17 States: Resolution of the Charleston Earthquake Issue," 1989.
18 19 NUREG/CR-5250, " Seismic Hazard Characterization of 69 Nuclear Plant Sites East 20 of the Rocky Mountains," 1989.
'21 22 Letter from Marvin Mendonca, NRC to D.W. Mazur, Washington Public Power Supply 23 System, "NRC Review of Seismic Report for WNP-3," January 4,1991.
24 ,
25 NUREG-0675, Supplement No. 34, " Safety Evaluation Report related to the operation of Diablo Canyon Nuclear Pwer Plant, Units 1 and 2," 1991.
e DG-1015-29
e a
] ,
-l
! 1
! 1 2
ADDendix 0 to Reaulatory Guide DG-1015 l i 3
. Geoloaical. Seismoloaical and Geophysical Investiaations to :
, 4 Characterize Seismic Sources l 5
- 6 0.1 Introduction
! 7
- 8 9 Seismic sources define areas where future earthquakes are likely to occur.
- 10 Geological and seismological investigations provide the information needed to j 11 characterize source parameters, including the size and geometry of the seismic
[ 12 sources, earthquake recurrence models, and deterministic source earthquakes 13 (DSE). The amount of data available about earthquakes and their causative t 14 sources varies substantially between the western U.S. and the stable continental :
15 region (SCR) and also from region to region within these broad areas. In active 4
16 tectonic regions the focus will be on the identification of both capable tectonic sources and seismogenic sources and the methods described in section 02 can be
- 17
. 18 applied. In the SCR east of the Rocky Mountains, seismogenic sources play a
! 19 significant role because of the difficulty in unequivocally correlating
. 20 earthquake activity with known tectonic structures. ;
21 1 22 In the SCR a number of significant tectonic structures exist which have been 23 suggested as potential seismogenic sources (i.g. New Madrid fault zone, Nemaha i 24 Ridge, Meers fault, Ramapo fault zone, Clarendon-Linden fault). There is no -
! 25 clear procedure to follow to characterize the DSE magnitude associated with such i
26 possible seismogenic sources; therefore, it is most likely that the determination
!' 27 of the seismogenic nature of the source will be inferred rather than demonstrated 28 by strong correlations with seismicity and/or geologic data. Furthermore, it is 29 not.known what relations exist between observed tectonic structures in a given 30 seismogenic source and the current earthquake activity loosely correlated with 31 that source. Generally, the observed tectonic structure resulted from ancient !
32 tectonic forces that are no longer present, and thus the structural extent may 33 not .be a very meaningful indicator of the size of future earthquakes in the 34 source. Careful analysis of the historical record and the results of regional '
35 and site studies and judgment play key roles. If, on the other hand, such strong 36 correlations and/or data exist between seismicity and seismic sources, then 37 approaches used for active tectonic regions can be applied.
38 39 The following is a general list of characteristics to be determined for a seismic :
-40 source:
41 42 a. Source zone geometry (location and extent, both surface and subsurface). l 43 '
44 b. Description of Quaternary (last 2 million years) displacements (sense of 45 slip on the fault, fault length and width,, age of displacements, estimated ,
46 displacements per event, estimated magnitudes per offset, rupture length l 47 and area, and displacement history or uplift rates of seismogenic folds). l 48 49 c. Historical and instrumental seismicity associated with each source. r 50 51 d. Paleaseismicity.
52- ,
DG-1015-30
. . . _ . *Y _ _ . _ , ._ _ . . _ . - _. .,
s- e _
- e. Relationship of the fault to other potential seismic sources in the
. region.
3 4 f. Deterministic Source Earthquake.
5 6 g. Recurrence model (frequency of earthquake occurrence versus magnitude).
7 8 h. Effects of human activities such as withdrawal of fluid from or addition 9 of fluid to the subsurface,. extraction of minerals, or the effects of dams 10 or reservoirs.
11 12 1. Volcanism. Volcanic hazard is not addressed in this regulatory guide. It 13 -
will be considered on a case by case basis in regions where this hazard 14 exists.
15 16- J. Other factors that can contribute to characterization of seismic sources
. 17 such as strike and dip of tectonic structures, orientations of regional
. 18 and tectonic stresses, fault segmentation (both along strike and downdip),
l 19. etc.
i 20 l 21 D.2. Investications to Characterize Seismic Sources 22 j 23 a. General
! 24 i 'S Investigations of the site and region around the site are necessary to identify '
- both seismogenic sources and capable tectonic sources and determine their j potential for generating earthquakes and for causing surface deformation. Where i 28 it is determined that surface deformation need not be taken into account, 29 sufficient data to clearly justify the determination should be presented in the l 30 license application or early site review.
- 31 32 In the siting of nuclear power plants, engineering solutions are generally i 33 available to mitigate the potential vibratory effect of earthquakes through 34 design. However, such solutions cannot always be demonstrated as being adequate j i 35 for mitigation of the effects of permanent ground displacement phenomena such as :
) 36 surface faulting or folding,' subsidence, ground collapse or fault creep. For !
l 37 this reason, it is prudent to select an alternative site Ghen the potential for j
! 38 permanent ground displacement exists at the site (IAEA,.1991). In most of the i 39 eastern U.S. tectonic structures at seismogenic depths, as determined from 40 earthquake hypocenters, apparently bear no relationship to geologic structures i
.41 exposed at the ground surface. Young faults either do not extend to the ground L 42 surface or there is insufficient geologic material of the appropriate age
- 43 available to date the faults. Seismogenic faults are not always exposed at ground
- 44 surface in the western U.S. as demonstrated by the buried (blind) reverse sources 4
45 of the 1983 Coalinga,1988 Whittier Narrows and 19891.onra Prieta earthquakes.
{ 46 These factors emphasize the need to not only conduct thorough investigations at
- 47 the ground surface but also to identify structures at seismogenic depths.
i 48
! 49 The level of detail for investigations should be governed by the current and late 4 50 Quaternary tectonic regime and the geological complexity of the site and region. l j Whenever faults or other structures are encountered at a site (including in the
{
SCR) either in outcrop or excavations, it is necessary to perform many of the j DG-1015-31 4
d
~
1 1 investigations described below to demonstrate whether or not they are capable 2 tectonic sources. ,
3 4 Regional investigations should extend to a distance of 320 km (200 miles) from 5 the site and data presented at a scale of 1:500,000 or smaller. Investigations :
6 of greater detail should be conducted to a distance of 40 km (25 miles) from the Detailed 7 site and the data presented at a scale of 1:50,000 or smaller.
8 investigations should be carried out within a radius of 8 km (5 miles) from the site and data presented at a scale of 1:5000 or smaller. Data from 9
10 investigations within the site area (approximately I km2) should be presented 11 at a scale of 1:500 or smaller. The areas of investigations may be asymmetrical l 12 and larger than those described above in regions of late Quaternary activity or 13 historical seismic activity (felt or. instrumentally recorded data) or where a +
14 site is located near a capable tectonic source such as a fault zone.
15 .
16 Regional and site information needed to assess the integrity of the site with 17 respect to potential ground motions and surface deformation caused by capable 18 tectonic. sources include determination of: (1) the lithologic, stratigraphic, l r 19 geomorphic, hydrologic, geotechnical and structural geologic characteristics of 20 the site and the area surrounding the site, including its geologic history; (2) 21 geologic evidence of fault offset or other distortion such as folding at or near J 22 ground surface at or near the site; and (3) determination of whether or not any 23 faults or other tectonic structures any part of which are within a radius of 8 i 24 km (5 miles) are capable tectonic sources. This information will be used to i 25 evaluate tectonic structures underlying the site, whether buried or expressed at '
26 the surface, with regard to their potential for generating earthquakes and for 27 causing surface deformation at or near the site. The evaluation should consider i 28 the possible effects caused by human activities such as withdrawal of fluid from l 29 or addition of fluid to the subsurface, extraction of minerals, or the loading 30 effects of dams or reservoirs. !
31 32 b. Reconnaissance investiaations.1.iterature Review and Other Sources of i 33 Preliminary Information l 34 35 Site and regional investigations can be planned based on field reconnaissances 36 data from previous investigations and reviews of available documents. Possible 37 sources of information may include universities, consulting firms and government 38 agencies. A detailed list of possible. sources of information is given in 39 Regulatory Guide 1.132.
40 41 c. Detailed Investications to Characterize Seismic Sources 42 43 The following methods . are . suggested but they are not all-inclusive and 44 investigations should not be limited to them. Some procedures- will not be 45 applicable to every site and situations will occur requiring investigations which 46 are not included in the following discussion. -It is anticipated that new 47- technologies will be available in the future that will be applicable to these I 48 investigations. )
49 50 Surface exploration needed to assess neotectonic conditions of the geology of the 51 area around the site is dependent on the site location and may be carried out DG-1015-32
~
- ~
e e with the use of any appropriate combination of geological, geophysical, seismological and geotechnical engineering techniques.
3 4 (1) . Geological interpretations of aerial photographs and other remote-sensing 5 imagery, as appropriate for the particular site conditions, to assist in 6 identifying- rock outcrops, faults and other tectonic features, fracture 7 traces, geologic contacts, lineaments, soil conditions, and evidence of 8 landslides or soil liquefaction.
9 ..
10 (2) Mapping of topographic, geologic, geomorphic and hydrologic features at 11 scales and contour intervals suitable for analysis, stratigraphy 12 (particularly - Quaternary), surface tectonic structures such as fault
. zones, and Quaternary geomorphic features. For offshore sites, coastal J 13 -
14 sites, or sites located near lakes or rivers this includes topography, 15 geomorphology (particularly mapping marine and fluvial terraces),
16 bathymetry, geophysics (such as seismic reflection), and hydrographic 17 surveys to the extent needed for evaluation.
18 19 (3) Identification and evaluation of vertical crustal movements by:
20 (a) geodetic land surveying to identify and measure short term crustal ;
21 movements (Reilinger and others, 1984; Mark and others, 1981) and l 22 (b) geological analyses such as analysis of regional dissection and 23 degradation patterns, marine and lacustrine terraces and shorelines, 24 fluvial adjustments such as changes in stream longitudinal' profiles or 25 terraces and other long term changes such as elevation changes across lava :
flows, etc. (Rockwell and others, 1984) 28 (4) Analysis of offset, displaced or anomalous landforms such as displaced i 29 stream channels or changes in stream profiles or the upstream migration of
, 30 knickpoints (Sieh,1984; Sieh and Jahns,1984; Sieh and others,1989; 31 Weldon and Sieh, 1985; Swan and others, 1980; PG&E, 1988), abrupt changes !
32 in fluvial deposits or terraces, changes in paleochannels across a fault 33 (Swan and others, 1980), or uplifted, downdropped or laterally displaced 34 marine terraces (PG&E, 1988).
35 36 (5) Analysis of Quaternary sedimentary deposits within or near tectonic zones
}7 such as fault Zones and including: (a) fault related or fault controlled f 38 deposits including sag ponds, graben fill deposits, and colluvial wedges 39 formed by the erosion of a fault paleoscarp, and (b) non-fault related, !
40 but offset deposits including alluvial' fans, debris cones, fluvial terrace 41 and lake shoreline deposits.
42 43 (6) Identification and analysis of deformation features caused by vibratory 44 ground motions including seismically induced liquefaction features (sand 45 boils, explosion craters, lateral spreads, settlement, soil flows), mud 46 volcanoes, landslides, rockfalls, deformed lake deposits or soil horizons, 47 shear zones, cracks or fissures (Obemeier and others,1985; Amick and 43 others, 1990).
49 50 (7) Estimation of the ages of fault displacements by analysis of the
~
morphology of topographic fault scarps associated with or produced by surface rupture. Fault scarp morphology is useful in estimating age of DG-1015-33 f
1 last displacement, approximate size of the earthquake, recurrence 2 intervals, slip rate and the nature of the causative fault at depth 3 (Wallace, 1977, 1980, 1981; Crone and Harding, 1984).
4 5 (8) Listing of all historically reported earthquakes which can reasonably be 6 associated with seismic sources any part of which is within a radius of 7 320 km (200 miles) of the site, including date of occurrence and the 8 following measured or estimated data: highest intensity, magnitude, 9 epicenter, depth, focal mechanism, stress drop, etc. Historical 10 seismicity includes both historically reported and instrumentally recorded 11 data. For pre-instrumentally recorded data, intensity should be converted 12 to magnitude, the procedure used to convert it to magnitude should be 13 clearly documented, and epicenters should be determined based on intensity 14 . contours. Methods to convert intensity values to magnitudes in the is central and eastern U.S. are described in Nuttli (1979), Street and i 16 Turcotte (1975), and Street and Lacroix (1979).
17 18 (9) Seismic monitoring in the site area should be established as soon as
-19 possible after site selection.
20 21 Subsurface investigations that should be accomplished in the site area or within 22 the region to identify and define seismogenic sources and capable tectonic 23 sources may include:
24 ,
25 (1) Geophysical investigations such as air or ground magnetic and gravity 1 26 surveys, seismic reflection and seismic refraction surveys, borehole 27 geophysics, and ground penetrating radar.
28 29 (2) Core borings to map subsurface geology and obtain samples for testing 30 such as age dating.
31 32 (3) Excavating and logging trenches across geological features as part of the 33 _ neotectonic investigation and to obtain samples for age dating those 34 features.
35 36 At some sites, deep soil, bodies of water, or other material may obscure geologic 37 evidance of past activity along a tectonic structure. In such cases the analysis 38 of evidence elsewhere along the structure can be used to evaluate its 39 characteristics in the vicinity of the site (PG&E, 1988; NUREG-0675, 1991).
40 41 An important part of the geologic investigations to identify and define potential 42 seismic sources - is the age-dating of geologic materials. The following 43 techniques are useful in dating Quaternary deposits:
44 45 (1) Radiometric Dating Methods 46 ,
47 (a) Carbon 14 for dating organic materials (upper limit ranges from 48 30,000 up to 100,000 years) (Callender,1989).
4 9 ._ (b) Potassium argon for dating volcanic rocks ranging in age from about 50 50,000 to 10 million years (Callender, 1989).
51 (c) Uranium series uses the relative properties of various decay 52 products o f '"U o r '"U . Ages range from 10,000 to 350,000 DG-1015-34
- l. .
(Callender, 1989). '"U/'"U can yield between 40,000 and 1,000,000 years (Muhs and Szabo, 1982)
(d) Fission track uses minerals such as zircon and apatite, with 4 fissionable uranium in volcanic rocks. Although some interpretation !
5 is required in counting tracks, the technique has no inherent age 6 range limitations if suitable materials are available (Callender, 1989). i 8 (e) Thermoluminescence (TZ) is best used for stratigraphic correlation 9 and determining relative ages rather than absolute ages. The 10 maximum age is 10 million years (Callender,1989).
11 (f) Electron spin resonance (ESR) is used to date quartz that formed in 12 fault gouge during the fault event (Ikeya and others, 1982).
13
- l 14 (2) Other Quantitative Numerical Methods l 15 16 (a) Paleomagnetic dating requires material containing magnetic-17 susceptible minerals with sufficient stratigraphic and time ranges j 18 to provide several reversals. An independent time datum for i 19 correlation with the polarity time scale is required (Callender, l 20 1989).
l 21 (b) Thicknesses of weathering rind development on the margins of clasts, 22 such as caused by obsidian hydration, can be used to estimate the 4
- 23 age of deposits (Coleman and Pierce, 1981). !
24- (c) Cation-ratio dating of desert varnish on rock surfaces by chemical 25 analysis (Dorn, 1983).
l (d) Tephrochronology, which is the identification and correlation of l undated and dated volcanic ashes by geochemical and petrographic l .. analyses (Sheets and Grayson, 1979; Self and Sparks, 1981).
29 (e) Amino-acid racemization uses organic material and is based on time-30 dependent- diagenetic conversion of one form of amino-acid polymer 31 structure to another (Bada and Helfman, 1975; Bada and Protsch, 32 1973). -
33 (f) Lichenometry is used to estimate ages from sizes of lichens growing 34 on gravel or boulders (such as glacial deposits) (l.ocke and others, 35 1979). .
36 (g) Soil profile development is used to determine age based on measured 37 amounts of accumulated pedogenic materials (Machette, 1978).
38 (h) Dendrochronology is used to determine the ages of trees that were 39 affected by a tectonic event or other phenomena such as landsliding 40 or flooding (Page,1970; Sieh,1978; Atwater and Yamaguchi,1991).
41 i 42 (3) Relative Age Dating Methods 43 44- (a) Relative degree of soil profile development of B and C horizons can 45 provide at least an order of magnitude estimate of the ages of l 46 buried soils or relict surface soils on surficial . deposits 47 (Callender, 1989; Machette, 1982). For B horizons the diagnostic
! 48 characteristics include: thickness, depth, amount, texture, type of l 49 clay, soil structure and color, and amount of Fe oxides or Fe-Al-50- organic accumulation (Callender, 1989). For C horizons the l 51 important diagnostic characteristics are thickness, depth, stage of
- development and amount of pedogenic carbonate and other soluble I
DG-1015-35
1- salts (Macfadden and Tinsley,1982; Hardin,1982). Other references 2 for this subject include Matti and others, 1982; Pearthree and i 3 Calvo, 1982; Pearthree and others, 1983; Keller and others, 1984, 4 and Chadwick and others, 1984.
5 (b) Relative degree'of weathering of surface and subsurface clasts in 6 sedimentary deposits such as glacial moraines is useful but requires 7 independent
^
means of age calibration (Callender,1989).
8 9 In the SCR it may not be possible 'to demonstrate, in an absolute manner, the age 10 of last activity of a tectonic structure. In such cases the NRC staff will 11 accept association of such structures with geologic structural features or 12 tectonic processes which are geologically old (at least pre-Quaternary) as an age 13 indicator in the absence of conflicting evidence.
14
'15 These investigative procedures should also be applied, where possible, to 16 characterize offshore structures (faults or fault zones, and also folds, uplift 17 or subsidence related to faulting at depth) for coastal sites or those sites is located adjacent to landlocked bodies of water. Investigations of offshore 19 structures will rely heavily on seismicity, geophysics and bathymetry rather than 20 conventional geologic mapping methods which can be used effectively onshore.
21 However, it is often useful to investigate similar features onshore to learn more 22 about the significant offshore features.
23 24 25 d. Distinction Between Tectonic and Nontectonic Deformation 26 27 Nontectonic defermation like tectonic deformation can pose a substantial hazard 28 to nuclear power plants but there are likely.to be differences in the approaches 29 used to resolve the issues raised by the two types of phenomena. Therefore, non-30 tectonic deformation should be distinguished from tectonic deformation at a site.
31 In past nuclear power plant licensing activities, surface displacements caused -
32 by phenomena other than tectonic phenomena have been confused with tectonically-33 induced faulting. Such features include faults on which the last displacement was 34 induced by glaciation or deglaciation, collapse structures, such as found in .
35 karst terrain, and growth faulting, such as occurs in the Gulf Coastal Plain or 36 in other deep soil regions subject to extensive subsurface fluid withdrawal.
37 38 Glacially induced faults generally do not represent a deep seated seismic or i 39 fault displacement hazard because the conditions that created them are no longer However, residual stresses from Pleistocene glaciation may still be
,40 present.
41 present in glaciated regions although they are of less concern than .ctive 42 tectonically induced stresses. These features should be investigated with respect .
43 to their relationship to current in-situ stresses.
44 ,
45- The nature of faults related to collapse features can usually be defined through 3 46- geotechnical investigations and can either be avoided, or if feasible, adequate 47 engineering fixes can be provided.
48 49 Large, naturally occurring growth faults as found in the coastal plain of Texas 50 and Louisiana can pose a surface displacement hazard even though offset most 51 likely occurs at a much'less rapid rate than that of tectonic faults. They are 52 not regarded as having the capacity to generate damaging earthquakes, can often DG-1015-36
3 i
be identified and avoided in siting, and their displacements can be monitored.
3 Some greyth faults and antithetic faults related to growth faults are not easily 4
identified; therefore, investigations described above with respect to capable tectonic fault; and fault zones should be applied in regions where growth faults 5 are known to be present. Local human-induced growth faults can be monitored and 6 controlled or avoided.
7 s ,.
9 If questionable features cannot be demcnstrated to be of non-tectonic origin they should be treated as tectonic deformation.
10 11 REFERENCES 12' :
13 . Amick, D., R. Gelinas, G. Maurath, D. Moore, F. Billington, and H. Kemppinen, 14 1990, Paleoliquefaction Features Along the Atlantic Seaboard; U.S. Nuclear 15 Regulatory Commission NUREG/CR-5613, 146p.
16 17' Atwater, 8. F., and D. K. Yamaguchi, 1991, Sudden, Probably . Coseismic 18 Submergences of Holocene Trees and Grass in Coastal Washington State; Geology, 19 V. 19, p. 706-709.
20 21 Bada, J. L., and P. M. Helfman,1975, Amino Acid Racemization Dating of Fossil 22 -Bones; World Archeology.
23 i '24 Bada, J.' L., and R. Protsch,1973, Racemization Reaction of Aspartic Acid and its l 25 Use in' Dating Fossil Bones; Proc, Nat. Acad. Sci. USA, vol. 70, p. 1331-1334.
Callender, J. F.,1989, Tectonics and Seismicity; Chapter 4 in Techniques for >
ed Determining Probabilities of Events and Processes Affecting the Performance of 29 Geologic Repositories, NUREG/CR-3964 SAND 86-0196, Vol.1, Edited by R. L. Hunter 30 and C. J. Mann, p.89-125.
31 32 Chadwick, 0. A., S. Hecker, and J. Fonseca,1984, A Soils Chronosequence at 33 Terrace Creek: Studies of Late Quaternary Tectonism in Dixie Valley, Nevada; 34 Open-file Report 84-0090, U.S. Geological Survey, 32 pp.
35 36 Colman, S. M., and K. L. Pierce,1981, Weathering Rinds on Andesitic and Basaltic 37 Stones as a Quaternary Age Indicator, Western United States; Prof. Paper 1210, as U.S. Geological Survey, 56pp. l 39 Crone, A. J., and S. T. Harding,1984, Relationship of Late Quaternary Fault 40 41 Scarps to Subjacent Faults, Eastern Great Basin, Utah; Geology, vol.12, p. 292-42 295.
43 44 Dorn, R. I., 1983, Cation-Ratio Dating: A New Rock Varnish Age-Determination 45- Technique; Quaternary Research, vol. 20, p. 49-73.
46 47- Harden, J. W., 1982, A Quantitative Index of Soil Development from Field 48 Descriptions: Examples from a Chronosequence in Central California; Geoderma, vol. 28, p. 2-18. ;
49 50
- 1 Ikeya, M., T. Miki, _ and K. Tanaka,1982, Dating of a Fault by Electron Spin Resonance on Intrafault Materials; Science, vol. 215, p.1392-1393.
CH3-1015-37
. - . -J
1 1
International Atomic Energy Agency, 1991, Earthquakes and Associated Topics in ;
2 Relation to Nuclear Power Plant Siting; Safety Series No. 50-SG-SI (Rev.1).
3 4 Keller, E. A., M. S. Bonkowski, R. J. Korsch, and R. J. Shlemen,1982, Tectonic 5 Geomorphology of the San Andreas Fault Zone in the Southern Indio Hills, 6
Coachella Valley, California; Geol. Soc. Amer. Bull ., vol. 93, p. 45-56.
7 s Locke, W. W., J. T. Andrews, and P. J. Webber,1979, A Manual for Lichenometry; 9
Technical Bull. 26, British Geomorphological Research Group, Norwich, Univ. of 10 East Anglia.
11 12 Machette, M. N.,1978, Dating Quaternary Faults in the Southwestern United States 13 by Using Buried Calcic Paleosols; U.5/ Geological Survey Jour. Research, vol. 6, 14 - p. 369-381. ,
15 16 Machette, M. N., 1982, Soil Dating Techniques, Western Region (United States);
17 Open-file Report 0FR-82-840, U.S. Geological Survey, p. 137-140.
18 19 Mark, R. K., J. C. Tinsley, E. 8. Newman, T. D. Gilmore, and R. O. Castle,1981, 20 An Assessment of the Accuracy of the Geodetic Measurements that Led to the 21 Recognition of the Southern California Uplift; Jour. Geophys. Research, vol. 86, 22 p. 2783-2808.
23 24 Matti, J. C., J. C. Tinsley, D. M. Morton, and L. D. McFadden, 1982, Holocene 25 Faulting History as Recorded by Alluvial Stratigraphy Within the Cucamonga Fault 26 Zone; A Preliminary View; in J. C. Tinsley, J. C. Matti, and L. D. McFadden, 27 eds., Guidebook, Field Trip No.12, Geol. Soc. Amer., Cordillera Section, p. 29-28 44. +
29 30 McFadden, L. D., and J. C. Tinsley, 1982, Soil Profile Development in Xeric 31 Climates: A Summary; in J. C. Tinsley, J. C. Matti, and L. D. McFadden, eds.,
32 Guidebook, Field Trip No.12, Geol. Soc. Amer., Cordillera Section, p.15-19.
33 34 Muhs, D. R., and B. J. Szabo,1982, Uranium-Series Age of the Eel Point Terrace, 35 San Clemente Island, California; Geology, vol.10, p. 23-26.
36 37 NUREG-0675, Supplement No. 34, 1991, Safety Evaluation Report Related to the 38 Operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2.
39 40 Nuttli, O. W.,1979, The Relation of Sustained Maximum Ground Acceleration and 41 Velocity to Earthquake Intensity and Magnitude, State-of-the-art for Assessing 42 Earthquake Hazards in the Eastern United States; U.S. Army Corps of Engineers 43 Misc. Paper 5-73-1, Report 16.
44 45 Obermeier, S. F., G. S. Gohn, R. E. Weems, R. L. Gelinas, and M. Rubin, 1985, 46 Geologic Evidence for Recurrent Moderate to Large Earthquakes Near Charleston, 47 South Carolina; Science, vol. 227, p. 408-411.
48 49 Pacific Gas and Electric Company,1988, Final Report of the Diablo Canyon Long 50 Term Seismic Program; Diablo Canyon Power Plant, Docket Nos. 50-275 and 50-323.
51 57 D3-1015-38
i
)
~ l l Page R.,1970, Dating Episodes of Faulting From Tree Rings: Effects of the 1958 l Rupture of the Fairweather Fault on the Tree Grcwth; Geol. Soc. Amer. Bull., vol. i 3 81, p. 3085-3094. !
! 4 l 5 Pearthree, P. A., and S. S. Calvo,1982, Late Quaternary Faulting West of the 6 Santa Rita Mountains South of Tucson, Arizona; M. S. Thesis, Univ. of Arizona, 7 Tucson, AZ, 49 pp.
8 9 Pearthree, P. A., C. M. Menges, and L. Mayer,1983, Distribution, Recurrence, and 10 Possible Tectonic Implications of Late Quaternary Faulting in Arizona; Open-file 11 Report 83-20, Arizona Bureau of Geology and Mineral Technology, 51 pp.
12 13 . Reilinger, R., M. Bevis, and G. Jurkowski,1984, Tilt from Releveling: An 14 Overview of the U.S. Data Base; Tectonophysics, vol.107, p. 315-330.
15 i 16 Rockwell, T. K., E. A. Keller, M. N. Clark, and D. L. Johnson,1984, Chronology 1 17- and Rates of Faulting of Ventura River Terraces, California; Geol. Soc. Amer.
18 Bull., vol 95, p. 1466-1474.
19 20 Sheets, P. D., and D. K. Grayson, eds.,1979, Volcanic Activity and Human 21 Ecology; Academic Press, New York.
22 23 Sieh, K. E., 1978, Prehistoric Earthquakes Produced by Slip on the San Andreas 24 Fault at Pallett Creek, California; Journal Geophys. Research, vol. 83, p. 3907-25 3939.
Sieh, K. E., 1984, Lateral Offsets and Revised Dates of Prehistoric Earthquakes 28 at Pallett Creet, Southern California; Jour. Geophys. Research, vol. 89, no. 89, .
29 p. 7641-7670. )
30 31 Sieh, K. E. and R. H. Jahns,1984, Holocene Activity of the San Andreas Fault at 32 Wallace Creek, California; Geol. Soc. Amer. Bull., vol. 95, p. 883-896.
33 34 Sieh, K., M. Stuiver, and- D. Brillinger,1989, A More Precise Chronology of 35 Earthquakes Produced by the San Andreas Fault in Southern California; Journal of 36 Geophysical Research, vol. 94, p. 603-623.-
37 38 Self, S., and R. J. S. Sparks, eds.,1981, Tephra Studies; Proc. NATO Advanced 39 Studies Institute, Tephra Studies as a Tool in Quaternary Research, D. Reidel 40 Publ . Co. , Dordrecht, Holland.
41 42 Street, R. L., and A. Lacroix, 1979, An Empirical Study of New England 43 Seismicity; Bulletin of the Seismological Society of America, vol. 69, p.159-44 176.
45 46 Street, R. L., and F. T. Turcotte, 1977, A Study of Northeastern North America 47 Spectral Moments, Magnitudes and Intensities; Bulletin of the Seismological Society of America, vol. 67, p. 599-614.
l 48
!= 49 l 50 Swan, F. H., III, D. P. Schwartz, and L. S. Cluff,1980, Recurrence of Moderate to Large Magnitude Earthquakes Produced by Surface Faulting on the Wasatch Fault l
Zone; Bull . Seismol . Soc. Amer., vol. 70, p.1431-1462.
DG-1015-39 l . . . .
- ~ . _ _ _ _ _ _ _ _ _ . , , _ _ , ,,
?
1 U.S. NRC,1979, Site Investigations for Foundations of Nuclear Power Plants; 2 Regulatory Guide 1.132, 25 pp.
3 4 Wallace, R. E.,1977 Profiles and Ages of Young Fault Scarps, North-Central 5 Nevada; Geol. Soc. Amer. Bull., vol. 88, p. 1267-1281.
6 7 Wallace, R. E.,1980, Discussion--Nomographs for Estimating Components of Fault 8 Displacement from Measured Height of Fault Scarp; Bull. Assoc. Engineering 9 Geologists, vol. 17, p. 39-45.
10' 11 Wallace, R. E.,1981, Active Faults, Paleoseismology, and Earthquake Hazards:
12 Earthquake Prediction--An International Review; Maurice Ewing Series 4, Amer.
13 Geophys. Union, p. 209-216. <
14 .
15 Weldon, R. J.,111, and K. E. Sieh,1985, Holocene Rate of Slip and Tentative-Recurrance Interval for Large Earthquakes on the San Andreas Fault, Cajon Pass, 16 17 Southern California; Geol _. Soc. Amer. Bull . , vol . 96, p. 793-812.
18 1
DG-1015-40
e .. .
')
l REGULATORY ANALYSIS The A separate regulatory analysis was not prepared for this regulatory guide.
5 draft regulatory analysis " Proposed Revision of 10 CFR Part 100 and 10 CFR Part 6 50," provides the regulatory basis for this guide and examines the costs and !
7 benefits of the rule as implemented by the guide. A copy of the draft regulatory 8 analysis is available for inspection and copying for a fee at tho NRC Public 9 Document Room, 2120 L Street NW. (Lower Level), Washington, DC, as Enclosure 2 10 to Secy 92-???. Single copies of the draft regulatory analysis are available 11 from Mr. Leonard Soffer, Office of Nuclear Regulatory Research, Mail Stop NL/5-12 324, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 13 492-3916 or Dr. Andrew J. Murphy, Office of Nuclear Regulatory Research, Mail 14 Stop NL/S-217A, U.S. Nuclear Regulatory Commission, Washington, DC 20555, 15 telephone (301) 492-3860.
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i l 1 DRAFT REGULATORY GUIDE DG-1016 1 2 SECOND PROPOSED REVISION 2 TO REGULATORY GUIDE 1.12 i 3 NUCLEAR POWER PLANT INSTRUMENTATION FOR EARTH 0VAKES j 4
5 6
7 A. INTRODUCTION 8
9 In 10 CFR Part 20, " Standards for Protection Against Radiation," licensees are 10 required to make every reasonable effort to maintain radiation exposures as 11 low as is reasonably achievable. Paragraph (c) of 150.36, " Technical 12 Specifications," to 10 CFR Part 50, " Domestic Licensing of Production and ;
13 Utilization Facilities," requires the technical specifications of a facility 14 to include surveillance requirements to ensure that the necessary quality of 15 systems and components is maintained, that facility operation will be within 16 safety limits, and that the limiting conditions of operation will be met.
17 Paragraph IV(a)(4) of Proposed Appendix S, " Earthquake Engineering Criteria 18 for Nuclear Power Plants," to 10 CFR Part 50 would require that suitable 19 instrumentation be provided so that the seismic response of nuclear power
'20 plant features important to safety can be evaluated promptly. Paragraph 21 IV(a)(3) of Proposed Appendix S to 10 CFR Part 50 would require shutdown of 22 the nuclear power plant if vibratory ground motion exceeding that of the )
23 Operating Basis Earthquake (OBE) ground motion occurs.' l 24 25 This guide is being developed to describe seismic instrumentation acceptacle 26 to the NRC staff for satisfying the requirements of Parts 20 and 50 and the 27 Proposed Appendix 5 to Part 50.
28 29 Any information collection activities mentioned in this draft regulatory guice 30 are contained as requirements in the proposed amendments to 10 CFR Part 50 31 that would provide the regulatory basis for this guide. -The proposed 32 amendments have been submitted to the Office of Management and Budget for 33 clearance that may be appropriate under the Paperwork Reduction Act. Such
,. 34
' Guidance is being developed in Draft Regulatory Guide DG-1017, " Pre-l 35 Earthquake Planning and Immediate Nuclear Power Plant Operator Post-
! 36 Earthquake Actions," to provide plant shutdown criteria.
DG-1016 - 1 APr 7,1992
4 1:
1-l 1 clearance, if obtained, would also apply to any information collection 2 activities' mentioned in this guide.
S 3 l
4
- 5 i 6 B. DISCUSSION 7
l' l 8 Wnen as earthquake occurs, it is important to assess immediately the effects
! 9 of the earthquake at the nuclear power plant. State-of-the-art solid-state l 10 - digital time-history accelerographs installed at appropriate locations will 11 provide time-history data on the. seismic response of the free-field, 12 containment structure, and other Category I structures. The instrumentation 13 should be located so that a comparison and evaluation of such response may be 14 made with the design basis and so that occupational radiation exposures are 15 maintained as low as reasonably achievable (ALARA).
16 17 Free-field instrumentation data would be used to determine if the OBE ground i 18 motion has been exceeded (see Draft Regulatory Guide DG-1017, " Pre-Earthquake 19 Planning and Immediate Nuclear Power Plant Operator Post-Earthquake Actions").
20 Foundation-level instrumentation would provide data on the actual seismic 21 input to the containment and other buildings and would quantify differences 22 between the vibratory ground motion at the free-field and foundation-level.
23 Instrumentation is not located on equipment, piping, or supports since 24 experience has shown that data obtained at these locations are obscured by 25 vibratory motion associated with normal plant operation.
26 27 The guidance being developed in Draft Regulatory Guide DG-1017 is based on the 28 assumption that the nuclear power plant has operable seismic instrumentation, 29 including the equipment and software required to process the data within four 30 hours after an earthquake. This is necessary because the decision to shut 31 down the plant will be made in part, by comparing the recorded data against 32 OBE exceedance criteria. The decision to shut down the plant is also based on 33 the results of the operator walkdown inspections which take place within eight 34 hours of the event.
35 36 It may not be necessary that identical nuclear power units on a given site 37 each be provided with seismic instrumentation if essentially the same seismic DG-1016 - 1 Apr 7,1992
s O O 4
1 response at each of the units is expected from a given earthquake.
2 3 An evaluation of seismic instrumentation operational experience noted that 4 ~ instruments have been out of service during plant shutdown and sometimes
! 5 during plant operation. The instrumentation system should be operable at all 6 times. If .the seismic instrumentation is inoperable, the guidelines being
. 7 developed in Appendix B to Draft Regulatory Guide DG-1017 should be used to l 8 determine if the Operating Basis Earthquake ground motion has been exceeded.
9 l
l 10 - Information pertaining to instrumentation characteristics, installation, 11 activation, remote indication, and maintenance is provided in this guide to
! 12 ensure (1) that the data provided are comparable with the data used in the 13 design of the nuclear power plant, (2) that exceedance of the Operating Basis 14 Earthquake can be determined, and (3) that the equipment will perform as 15 required.
16 17 Appendix A to this guide provides definitions to be used with this guidance.
18 19 20 21 C. REGULATORY POSITION 22 23 The type, locations, operability, characteristics, installation, actuation, 24 remote indication, and maintenance of seismic instrumentation described below 25' are acceptable to the NRC staff for satisfying the requirements in 10 CFR 26 20.1(c), 10 CFR 50.36(c), and Paragraph IV(a)(4) of Proposed Appendix S to 10 27 CFR 50'for ensuring the safety of nuclear power plants.
28 29 1. Seismic Instrumentation Tvoe and Location i 30 !
31 1.1 State-of-the-art solid-state digital instrumentation that will 32 enable the quick processing of data at the plant site should be 33 used.
i 34 35 1.2 A triaxial time-history accelerograph should be provided at each 36 of the following locations:
37 DG-1016 - 3 Apr 7,1992 o ;
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)
1 1. Free-field, j l 2 !
3 2. Containment foundation. ]
4 '. l
! 5 3. Two elevations (excluding the foundation) on a structure ;
j 6 internal to the containment.
7 i
8 4. Two. independent Category I structure foundations (for
)
9 instance, the diesel generator building and the auxiliary i 10 - building) where the response is different from that of the
(
11 containment structure.
12 j 13 5. An elevation (excluding the foundation) on each of the j 14 independent Category I structures selected in 4 above. I 15 16 6. If seismic isolators are used, instrumentation should be 17 placed on both the rigid and isolated portions of the 18 structures at approximately the same elevations.
19 20 1.3 The specific locations for instrumentation should be determined by 21 the nuclear plant designer to obtain the most pertinent 22 information consistent with maintaining occupational radiation l
23 exposures ALARA for the location, installation, and maintenance of l l 24 seismic instrumentation. In general:
25 26 1. A design review of location, installation, and maintenance l 27- of proposed instrumentation for maintaining exposures ALARA 28 should be performed by the facility in the planning stage in 29 accordance with Regulatory Guide 8.8, "Information Relevant 30 to Ensuring that Occupational Radiation Exposures at Nuclear 31 Power Stations Will Be As Low As Is Reasonably Achievable "
'32 33 2. Instrumentation should be placed in a location with as low a ;
34 dose rate as is practical, consistent with other 35 requirements.
36 37 3. Instruments should be selected to require minimal DG-1016 - 4 Apr 7,1992 l-
e .
_1 maintenance and in-service inspection, and minimal time and 2 numbers of personnel to conduct installation and 3 maintenance.
4 5 2. Instrumentation at Multi-Unit Sites 6
-7 Instrumentation in .ddition to that installed for a single unit will not 8
be required if essentially the same seismic response is expected at the 9
other units based on the seismic analysis used in the seismic design of 10 the plant. However, if there are separate control rooms, annunciation 11 should be provided to both tontrol rooms as specified in Regulatory 12 Position 7.
13 14 3. Seismic Instrumentation Ooerability 15
- 16 The seismic instrumentation should operate during all modes of plant
! ,17 operation, including periods of plant shutdown. The maintenance and 18 repair procedures should provide for keeping the maximum number of
- 19 instruments in service during plant operation and shutdown.
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- 21 4. Instr e tation Characteristics 22 23 4.1 Me design should include provisions for in-service testing. The 24 instruments should be capable of periodic channel checks during !
25 normal plant operation.
26 '
27 4.2 The instruments should have the capability for in-place functional 28 testing.
29 30 4.3 The instrumentation on the foundation and at elevations within the 31 same building or structure should be interconnected for common 32 starting and common timing, and the instrumentation should contain 33 provisions for an external remote alarm to indicate actuation.
34 35 4.4 The pre-event memory of the instrumentation should be sufficient 36 to record the onset of the earthquake; for example, it should have 37 the ability to record the 3 seconds prior to seismic trigger DG-1016 - 5 Apr 7,1992 s ~ -
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actuation. It should operate continuously during the period in 2
which the earthquake (xceeds the seismic trigger threshold and for 3
a minimum of 5 seconds beyond the last seismic trigger signal.
.4 The instrumentation should be capable of a minimum of 25 minutes 5
of continuous recording.
6 Acceleration Sensor (s).
7 4.5 8
9 1. The dynamic range should be 1000:1 zero to peak, for 10 .- example, 0.00lg to 1.09 11 12 2. The frequency range should be 0.20 Hz to 50 Hz, or an 13 equivalent demonstrated to be adequate by computational 14 techniques applied to the resultant accelerogram.
15 16 4.6- Recorder.
17 18 1. The sample rate should be at least 200 samples per second.
19 20 2. The bandwidth should be at least from 0.20 Hz to 50 Hz.
21 22 3. The dynamic range should be 1000:1.
23 24 4.7 . Seismic Trigger.
25 26 The actuating level should be adjustable for a minimum of 0.005g 27 to 0.02g.
28 29 5. Instrumentation Installation 30 31 5.1 The instrumentation should be designed and installed so that the 32 vibratory transmissibility over the amplified region of the design j
33 spectral frequency range is essentially unity, tbst is, the 34 mounting is rigid.
l 35 36 5.2 The instrumentation should be oriented so that the horizontal axes 37 are parallel to the orthogonal horizontal axes assumed in the
- DG-1016 - 6 Apr 7,1992
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1 seismic analysis.
2 j 3 5.3 Protection agdnst accidental impacts should be provided.
4 5 6. Instrumentatian A3 tuation 6
7 6.1 Both vertical and horizontal input vibratory ground motion should 8 actuate the same time-history accelerograph. One or more seismic 9 triggers may be used to accomplish this.
10 -
11 6.2 Spurious triggering should be avoided.
12 13 6.3 The seismic trigger mechanisms of the time-history accelerograph !
14 should be set for a threshold ground acceleration of not more than 15 0.02g.
16 17 7. Remote Indication 18 19 Activation of the free-field or any foundation-level time-history 20 accelerograph should be annunciated in the control room. If there are 21 two or more control rooms at the site, annunciation should be provided 22 to each control room.
23 1 24 8. Maintenance 25-26 8.1 The purpose of the maintenance program is to ensure that the 27 equipment will perform as required. As stated in Regulatory 28 Position 3, the maintenance and repair procedures should provide 29 for keeping the maximum number of instruments in service during 30 plant operation and shutdown.
31 32 8.2 Systems are to be given channel checks every two weeks for the 3 first three months of service after startup. Failures of devices 34 normally occur during initial operation. After the initial three-
, 35 month period and three consecutive successful checks, monthly 36 channel check are sufficient. The monthly channel check is to l
37 include checking the batteries. The channel functional test DG-1016 - 7 Apr 7,1992
l should be performed every 6 months. Channel calibration should be 1
2 performed during refueling.
3 D. IMPLEMENTATION 4
5 6 The purpose of this section is to provide guidance to applicants and licensees 7 regarding the NRC staff's plans for using this regulatory guide.
8 9 This proposed revision has been released to encourage public participation in 10 its development. Except in those cases in which the applicant proposes an 11 acceptable alter.;ative method for complying with the specified portions of the 12 Commission's regulations, the method to be described in the active guide 13 reflecting public comments will be used in the evaluation of applications for 14 a construction permit, operating license, con.bined license, or design 15 certification submitted after the implementation date to be specified in the 16 active guide. This guide would not be used in the evaluation of an 17 application for an operating license submitted after the implementation date 18 to be specified in the active guide if the construction permit was issued 19 prior to that date.
20 DG-1016 - 8 Apr 7,1992
e a APPENDIX A 1
DEFINITIONS 2
3 4
Acceleration Sensor. An instrument capable of sensing absolute acceleration 5 and transmitting the data to a recorder.
6 7
Channel Calibration (Primary Calibration). The determination and adjustment, 8 if required, of an instrument, sensor, or system such that it responds within 9 a specific range and accuracy to an acceleration, velocity, or displacement 10 - input, as applicable,' traceable to the National Institute of Standards.and 11 Technology (NIST), or an acceptable physical constant.
12 13 Channel Check. The qualitative verification of the functional status of the l
I 14 instrumen? sensor. This check is an "in-situ" test and may be the same as a j
15 channel functional test.
I 16 17 Channel Functional Test (Secondary Calibration). The determination without 18 adjustment that an instrument, sensor, or system responds to'a known input, 19 not necessarily traced to the National. Institute of Standards and Technology 20 (NIST), of such character that it will verify the instrument, sensor, or 21 system is functioning in a manner that can be calibrated.
22 23 Containment - See Primary Containment and Secondary Containment.
24 25 Operatina Basis Earthauake Ground Motion (OBE). The vibratory ground motion 26 for which those features of the nuclear power plant necessary for continued 27 operation without undue risk to the health and safety of the public will 28 remain functional. The value of the Operating Basis Earthquake Ground Motion 29 is set by the applicant.
30 31 Primary Containment. The principal structure of a unit that acts as the 32 ' barrier, after the fuel cladding and reactor pressure boundary, to control the l 33 release of radioactive material. It includes (1) the containment structure 34 and its access openings, penetrations, and appurtenances, (2) the valves, 35 pipes, closed systems, and other components used to isolate of the containment l 36- atmosphere.from the environment, and (D those systems or portions of systems 37 that, by their system functions, extene the containment structure boundary DG-1016 - 9 Apr 7,1992 l
1 (e.g., the connecting steam and feedwater piping) and provide effective 2 isolation.
3 4 Recorder. An instrument capable of simultaneously recording the data versus-5 time from acceleration sensor (s).
6 7 Secondary Containment. The structure surrounding the primary containment that 8 acts as a further barrier to control the release of radioactive material.
9 10 . Seismic Isolator. A device (for instance, laminated elastomer and steel) i 11 installed between the structure and its foundation to reduce the acceleration 12 of the isolated structure, and the attached equipment and components.
- 13 2
14 Seismic Triaaer. A device that starts the time-history accelerograph.
15 16 Time-History Acceleroaraoh. An instrument capable of measuring and l ,17 permanently recording the absolute acceleration versus time. The components 18 of the time-history accelerograph (acceleration sensor, recorder, seismic i 19 trigger) may be assembled in a self-contained unit or be separately located.
20 i.
] 21 Triaxial. Describes the function of an instrument or group of instruments in 22 three mutually orthogonal directions, one of which is vertical.
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1 REGULATORY ANALYSIS 2
3 A separate regulatory analysis was not prepared for this regulatory guide.
4 The draft regulatory analysis, " Proposed Revision of 10 CFR Part 100 and 10 5 CFR Part 50," provides the regulatory basis for this guide and examines the 1 6 costs and benefits of the rule as implemented by the guide. A copy of the 7 draft regulatory analysis is available for inspection and copying for a fee at 8 the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC, 9 as Enclosure 2 to Secy 92-???. Single copies of the draft regulatory analysis 10 are available from Mr. Leonard Soffer, Office of Nuclear Regulatory Research, 11 Mail Stop NL/S-324, U.S. Nuclear Regulatory Commission, Washington, DC 20555, 12 telephone (301) 492-3916 or Dr. Andrew J. Murphy, Office of Nuclear Regulatory 13 Research, Hail Stop NL/S-217A, U.S. Nuclear Regulatory Commission, Washington, 14 DC 20555, telephone (301) 492-3860.
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I ENCLOSURE 9 l
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4 DRAFT REGULATORY GUIDE DG-1017 l
- a PLANT SHUTDOWN l 1
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1 DRAFT REGULATORY GUIDE DG-1017 2 PRE-EARTHQUAKE PLANNING AND IMMEDIATE NUCLEAR POWER 3 PLANT OPERATOR POST-EARTHQUAKE ACTIONS 4 !
5 .
6 7 A. INTRODUCTION 8
9 Paragraph IV(a)(4) of Proposed Appendix S, " Earthquake Engineering Criteria 10 - for Nuclear Power Plants," to 10 CFR Part 50, " Domestic Licensing of ;
11 Production and Utilization Facilities," would require that suitable l 12 instrumentation' be provided so that the seismic response of nuclear power 13 plant ' features important to safety can be evaluated promptly. Paragraph l 14 IV(a)(3) of Proposed Appendix S to 10 CFR Part 50 would require shutdown of 15 the nuclear power plant if vibratory ground motion exceeding that of the 16 Operating Basis Earthquake Ground Motion or significant plant damage occurs.
17 Proposed Paragraph 50.54(ee) to 10 CFR 50 would require licensees of nuclear 18 power plants that have adopted the earthquake engineering criteria in Proposed 19 Appendix S to'10 CFR 50 to shut down the plant if the criteria in Paragraph 20 IV(a)(3) of Proposed Appendix S are exceeded.
21 22 This guide is being developed to provide guidance acceptable to the NRC staff 23 for a timely evaluation after an earthquake of the recorded instrumentation 24 data and for determining whether plant shutdown would be required by the 25 proposed amendments to 10 CFR Part 50.
26 27 Any information collection activities mentioned in this draft regulatory guide 28 are contained as requirements in the proposed amendments to 10 CFR Part 50 29 that would provide the regulatory basis for this guide. The proposed 30 amendments have been submitted to the Office of Management and Budget for 31 clearance that may be appropriate under the Paperwork Reduction Act. Such 32 clearance, if obtained, would also apply to any information collection 33 ' Guidance is being developed in Draft Regulatory Guide DG-1016, Second 34 Proposed Revision 2 to Regulatory Guide 1.12 " Nuclear Power Plant 35 Instrumentation for Earthquakes," to describe seismic instrumentation 36 acceptable to the NRC staff.
DG-1017 - 1 Apr 7, 1992
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1 activities mentioned in this guide.
2 i 3 !
l 4 i l 5 B. DISCUSSION t
6 7 When an earthquake occurs, ground motion data are recorded by the seismic 8 instrumentation.8 These data are used to make an early determination of the 9 degree of severity of the seismic event. The data from the seismic 10 - instrumentation, coupled with information obtained from a plant walkdown, are 11 used to make the initial determination of whether the plant should be shut 12 down, if it has not already been shut down by operational perturbations 13 resulting from the seismic event. If on the basis of these initial 14 evaluations (instrumentation data and walkdown) it is concluded that the plant 15 shutdown criteria have not been exceeded, it is presumed that the plant will
! 16 not be shut down. Guidance is being developed on post shutdown inspections l 17 and plant restart; see Draft Regulatory Guide DG-1018, " Restart of a Nuclear .
18 Power Plant Shut Down by a Seismic Event."
19 ,
20 21 The Electric Power Research Institute has developed guidelines that will !
22 enable licensees to quickly identify and assess earthquake effects on nuclear i 23 power plants. These guidelines are in EPRI NP-5930, "A Criterion for 24 Determining Exceedance of the Operating Basis Earthquake," July 1988, EPRI NP-25 6695, " Guidelines for Nuclear Plant Response to an Earthquake," December 1989, ;
26 and EPRI TR-100082, " Standardization of Cumulative Absolute Velocity,"
27 December 1991.*
28 29 This guide is based on the assumption that the nuclear power plant has 30 operable seismic instrumentation. If the seismic instrumentation is 31 inoperable, the guidelines being developed in Appendix A to this guide would
! 32 be used to determine whether the Operating Basis Earthquake Ground Motion has I 33 been exceeded.
i 34 h 35- 8 Copies may be obtained from the Research Reports Center (RRC), Box 50490, '
36 Palo Alto, California 94303.
DG-1017 - 2 Apr 7, 1992
n a 1 Shutdown of the nuclear power plant would be required if the vibratory ground 2 motion experienced exceeds that of the Operating Basis Earthquake (0BE) ground 3 motion. Two criteria for determining exceedance of the OBE are provided in 4 EPRI NP-5930: a threshold response spectrum ordinate criterion and a 5 cumulative absolute velocity criterion (CAV). A procedure to standardize the 6 calculation of the CAV is provided in EPRI TR-100082. In addition, a spectral 7 velocity threshold has also been recommended by EPRI since some structures 8 have fundamental frequencies below the range specified in EPRI NP-5930. The 9 staff now recommends 1.0 to 2.0 Hz for the range of the spectral velocity 10 - limit since some structures have fundamental frequencies below 1.5 Hz. The 11 former range was 1.5 to 2.0 Hz.
12 13 Decisions on continued operation will be made by the staff in conjunction with 14 the licensee on a case-by-case basis consistent with applicable regulations.
15 Therefore, the staff does not endorse the philosophy discussed in EPRI NP-16 6695, Section f..J./ (first paragraph, last sentence), pertaining to plant 17 shutdown considerations following an earthquake based on the need for 18 continued power generation in the region.
19 20 Appendix B to this guide provides definitions to be used with this guidance.
21 22 23 24 C. REGULATORY POSITION 25 26 1. Base-line Data 27 28 1.1 Information Related to Seismic Instrumentation 29 30 A file containing information on all the seismic instrumentation 31 should be kept at the plant. The file should include:
32 33 1. Information on each instrument type such as make, model, and 34 serial number; manufacturers' data sheet; list of special 35 features or options; performance characteristics; examples 36 of typical instrumentation readings and interpretations; 37 operations and maintenance manuals; repair procedures DG-1017 - 3 Apr 7, 1992
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I 1 (manufacturers' recommendations for repairing common
- l 2 problems); and a list of any special requirements, e.g.,
3 maintenance, operational, installation.
.4 l 5- 2.- Plan views and vertical sections showing the locations of 6 each seismic instrument and the orientation of the instru-7 ment axis with respect to a plant reference axis.
8 ,
9 3. A complete service history of each seismic instrument. The 10 - service history should include information such as dates of 11 servicing, description of completed work, and calibration 12 records and data (where applicable).
13 14 4. The response spectrum and cumulative absolute velocity (see 15 Regulatory Position 4). These data should be obtained af ter 16 the initial installation and each servicing of the free-17 field instrumentation using a suitable earthquake time 18 history (e.g., the October 1987 Whittier, California 19 earthquake) or manufacture's calibration standard.
20 21 1.2 Planning for Post Earthquake Inspections 22 23 The selection of equipment and structures for inspections and the 24 content of the base line inspections as described in Sections 25 S.3.1 and 5.3.2.1 of EPRI NP-6695, " Guidelines for Nuclear' Plant 26 Response to an Earthquake," are acceptable to the NRC staff for 27 satisfying the requirements in Paragraph IV(a)(3) of Proposed 28 Appendix S to 10 CFR Part 50 for ensuring the safety of nuclear 29 power plants.
30 31 2. Immediate Postearthauake Actions 32 33 The guidelines for immediate postearthquake actions specified in :
34 Sections 4.3.1 and 4.3.2 (including Section 5.3.2.1 and items 7 and 8 of 35 Table 5-1) of EPRI NP-6695 are acceptable to the NRC staff for 36 -satisfying the requirements indicated in Paragraph IV(a)(3) of Proposed 37~ Appendix S to 10 CFR Part 50.
DG-1017 - 4 Apr 7, 1992
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1 1 3. Evaluation of Ground Motion Records I 2
3 21 Data Identification 4
5 A record collection log should be maintained at the plant, and all 6 data should be identifiable and traceable with respect to:
7 8 1. The date and time of collection, 9
10 2. The make, model, serial number, location, and orientation of 11 the instrument (sensor) from which the record was collected.
12 3.2 Data Collection 13 14 1. Only personnel trained in the operation of the instrument 15 should collect the data.
16 17 2. Procedures for removing and storing records from each 18 seismic instrument should be preplanned and perforned in 19 accordance with established procedures.
20 21 3. Extreme caution should be exercised to prevent accidental 22 damage to the recording media and instruments during data 23 collection and subsequent handling.
24 25 4. As data are collected and the instrumentation is inspected, 26 notes should be made regarding the condition of the 27 instrument and its installation, for example, instrument 28 flooded, mounting surface tilted, whether fallen objects 29 might have struck the instrument or the instrument mounting i 30 surface.
31 32 5. For validation of the collected data, a reference signal 33 (see Regulatory Position 1.1(4)) should be added to the 34 record without affecting the previously recorded data.
35 36 6. If the instrument operation appears to have been normal, the 37 instrument should remain in service without readjustment or DG-1017 - 5 Apr 7, 1992
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- 1. change that would defeat attempts to obtain postevent-2 calibration.
- 3 4 3.3 Record Evaluation i 5 ;
6 Records should be analyzed according to the panufacturer's ;
7- specifications and the results of the analysis should be-8 ., evaluated. Any record anomalies, invalid data, and nonpertinent q 9 signals should be noted, with any known causes.
-10 ,
11 4. Determinina OBE Exceedance 12 ; l 13 The evaluation to determine if the OBE was exceeded should be performed :
14 using data obtained from the three components of the free-field ground f 15 motion (i.e., two horizontal and one vertical). The evaluation may be 16 performed on uncorrected earthquake records. It was found in a study of 17 uncorrected versus corrected earthquake records (EPRI NP-5930) that the j 18 use of uncorrected records is conservative. The evaluation should3 19 consist of a check of the response spectrum, cumulative absolute i velocity limit, and the operability of the instrumentation. I 20 21 22 4.1 Response Spectrum Check
't3 24 The OBE response spectrum is exceeded if any one of the three 25 components (two horizontal and one vertical) of the 5 percent 26 damped free-field ground motion response spectra is larger than:
27 28 1. The corresponding design response spectral acceleration (OBE 29 spectrum if used, otherwise 1/3 of the Safe Shutdown !
30 Earthquake (SSE) spectrue) or 0.2g, whichever is greater, 31 for frequencies between 2 to 10 Hz, or l 32 )
33 , 2. The corresponding design response spectral velocity (0BE l 34 spectrum if used, otherwise 1/3 of the SSE spectrum) or a 35 spectral velocity of 6 inches per second, whichever is 36 greater, for frequencies between 1 to 2 Hz. !
37 DG-1017 - 6 Apr T,1992 ,
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~1 i- 1 4.2 Cumulative Absolute Velocity (CAV) Limit 2
3 The CAV should be calculated as follows: For each component of 4
the free-field. ground motion, (1) the absolute acceleration (g 5
units) time-history is segmented into 1-second intervals, (2) 6 7,
each 1-second interval that has at least 1 exceedance of 0.0259 is integrated over time, (3) all the integrated values are summed 8- together to arrive at the CAV. Additional guidance on.how to 9 determine the CAV is provided in EPRI TR-100082.
10 11 Tiie CAV Limit is exceeded if any CAV calculation is greater than 12 0.16 g-second.
13 14 4.3 Instrument Operability Check 15 16 After an earthquake at the plant site, the response spectrum and 17 CAV should be obtained using the calibration standard (see 18 Regulatory Position 1.l(4)) to demonstrate that the system was 19 functioning properly, s 20 21 5. Criteria for Plant Shutdown ,
22 23 If the OBE vibratory ground motion is exceeded or significant plant-24 damage occurs, the plant must be shut down.
25 26 5.1 OBE Exceedance. If the response spectrum check and the CAV limit, 27 performed in accordance with Regulatory Position 4.1 and 4.2, were 28- exceeded, the OBE was exceeded and plant shutdown is required. If 29 either limit does not exceed the criterion, the earthquake motion 30 did not exceed the OBE. The determination of whether or not the 31 OBE has been exceeded should be performed even if the plant 32 automatically trips off-line as a result of the earthquake, or l 33 l 34 5.2 Damace. The plant should shutdown if the walkdown inspections, 35 performed in accordance with Regulatory Position 2 (Section 4.3.2 j 36 of EPRI NP-6695), discover damage.
! 37 DG-1017 - 7
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, a 1 6. Pre-Shutdown Insoections 2
3 The pre-shutdown inspections described in Section 4.3.4 of EPRI NP-6695,
" Guidelines for Nuclear Plant Response to an Earthquake," are acceptable 4
5 to the NRC staff for satisfying the requirements indicated in Paragraph
/ 6 IV(a)(3) of Proposed Appendix S to 10 CFR 50 for ensuring the safety of 7
nuclear power plants subject to the following:
8 9 6.1 Delete the last sentence in the first paragraph of Section 4.3.4.
10 6.2 The following paragraph in Section 4.3.4 is repeated to emphasize 11 that the plant should shut down in an orderly manner.
12 13
" Prior to initiating plant shutdown following an earthquake, 14 visual inspections and control board checks of safe shutdown 15 systems should be performed by plant operations personnel, 16 and the availability of off-site and emergency power sources 17 should be determined. The purpose of these inspections is 18 to determine the effect of the earthquake on essential safe 19 shutdown equipment which is not normally in use during power 20 21 operation so that any resets or repairs required as a result of the earthquake can be performed, or alternate equipment 22 In can be readied, prior to initiating shutdown activities.
23 order to ascertain possible fuel and reactor internal 24 damage, the following checks should be made, if possible, 25 26 before plant shutdown is initiated .... "
27 28 If the OBE was not exceeded and the walkdown inspection indicates no 29 damage to the nuclear power plant, shutdown of the plant is not 30 required. The plant may continue to operate (or restart following a 31 post-trip review, if it tripped off-line due to the earthquake).
32 33 34 D. IMPLEMENTATION 35 l
36 The purpose of this section is to provide guidance to applicants and licensees
)
1 37 Apr 7, 1992 DG-1017 - 8
I regarding the NRC staff's plans for using this regulatory guide. l 2
3 This draft guide has been released to encourage public participation in its 4 development. Except in those cases in which the applicant proposes an 5
acceptable alternative method fcr complying with the specified portions of the 6
Commission's regulations, the method to be described in the active guide 7
reflecting public comments will be used in the evaluation of applications for 8
a construction permit, operating license, combined license, or design 9
certification submitted after the implementation date to be specified in the 10 . active guide. This guide would not be used in the evaluation of an 11 application for an operating license submitted after the implementation date 12 to be specified in the active guide if the construction permit was issued 13 prior to that date.
14 DG-1017 - 9 Apr 7, 1992 0 ..
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APPENDIX A 2
3 INTERIM OPERATING BASIS EARTHQUAKE EXCEEDANCE GUIDELINES 4
5 6 Draft Regulatory Guide DG-1017 is based on the assumption that the nuclear 7 power plant has operable seismic instrumentation. If the seismic.instrumenta-8 tion is inoperable, the following should be used to determine whether the 9 ' Operating Basis Earthquake Ground Motion (OBE) has been exceeded:
10 11 1. For plants at which instrumentally determined data are available only at 12 the foundation level, the Cumulative Absolute Velocity (CAV) Limit (see 13 Regulatory Position 4.2 of.this guide) is not applicable, and a 14 determination of OBE exceedance is based on the response spectrum check A
15 described in Regulatory Position 4.1 of this regulatory guide.
16 comparison is made between the foundation level design response spectra If the 17 and data obtained from the foundation level instruments.
18 response spectrum check at any foundation is exceeded, the ORE is
-19 exceeded and shutdown is warranted.
20 21 2. For plants at which no instrumental data are available, the OBE will be 22 considered to have been exceeded and shutdown to be warranted if one of 23 the following applies:
24 25 1. The earthquake resulted in Modified Mercalli Intensity (MMI) VI or 26 greater within 5 km of the plant, 27 28 2. The earthquake was felt within the plant and was of magnitude 6.0 29 or greater, or 30 31 3. The earthquake was of magnitude 5.0 or greater, and occurred 32 within 200 km of the plant.
33 34 3. A postearthquake plant walkdown should be conducted (see Regulatory 35 Position 2 of this guide).
36 37 4. If plant shutdown is warranted under the above guidelines, the plant DG-1017 - 10 Apr 7, 1992 l
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I should be shut down in an orderly manner (see Regulatory Position 6 of 2 this guide).
3 4 8212:
5 The U.S. Geological Survey, National Earthquake Information Center, 6 determinations of epicentral location, magnitude, and intensity will 7 usually take precedence over other estimates; however, regional and 8 local determinations will be used if they are considered to be more 9 accurate. Also, higher quality damage reports or a lack of damage 10 - reports from the nuclear power plant site or its immediate vicinity will 11 take precedence over more distant reports.
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APPENDIX B 2
3 DEFINITIONS 4
5 Desian Response Soectra. Response spectra used to design Seismic Category 1 6 structures, systems, and components.
7 8 Ooeratina Basis Earthauake Ground Motion (OBE). The vibratory ground motion I
9 for which those features of the nuclear power plant necessary for continued 10 . operation without undue risk to the health and safety of the public will 11 remain functional. The val'ue of the Operating Basis Earthquake Ground Motion 12 is set by the applicant.
13 14 Soectral Acceleration. The acceleration response of a linear oscillator with 15 prescribed frequency and damping.
16 17 Soectral Velocity. The velocity response of a linear oscillator with pre-18 scribed frequency and damping.
19 20 21 DG-1017 - 12 Apr 7, 1992
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1 REGULATORY ANALYSIS 2
3 A separate regulatory analysis was not prepared for this regulatory guide.
4 The draft regulatory analysis, " Proposed Revision of 10 CFR Part 100 and 10 5 CFR Part 50," provides the regulatory basis for this guide and examines the 6 costs and benefits of the rule as implemented by the guide. A copy of the 7 draft regulatory analysis is available for inspection and copying for a fee at 8 the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC, l 9 as Enclosure 2 to Secy 92-???. Single copies of the draft regulatory analysis l 10 are available from Mr. Leonard Soffer, Office of Nuclear Regulatory Research, l 11 Mail Stop NL/S-324, U.S. Nuclear Regulatory Commission, Washington, DC 20555, 12 telephone (301) 492-3916 or Dr. Andrew J. Murphy, Office of Nuclear Regulatory 13 Research, Mail Stop NL/S-217A, U.S. Nuclear Regulatory Commission, Washington, 14 DC 20555, telephone (301) 492-3860.
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DG-1017 - 13 Apr 7, 1992
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ENCLOSURE 10 4 I
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DRAFT REGULATORY GUIDE DG-1018
! PLANT RESTART l
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s' O 1 DRAFT REGULATORY GUIDE DG-1018 2 RESTART OF A NUCLEAR POWER PLANT SHUT DOWN 3 BY A SEISMIC EVENT i
4-5 6
7 A. INTRODUCTION 8
l l 9 Paragraph IV(a)(3) of Proposed Appendix S, " Earthquake Engineering Criteria
- 10 - for Nuclear Power Plants," to 10 CFR Part 50, " Domestic Licensing of 11 Production and Utilization Facilities," would require shutdown of the nuclear 12 power plant if vibratory ground motion exceeding that of the Operating Basis l 13 Earthquake Ground Motion occurs.' Prior to resuming operations, the licensee 14 must demonstrate to the Commission that no functional damage has occurred to 15~ those features necessary for continued operation without undue risk to the 16 health and safety of the public.
17
! ~18 This guide is being developed to provide guidelines that are acceptable to the
! 19 NRC staff for performing inspections and tests of nuclear power plant l
20 equipment and structuras prior to restart of a plant that has been shut down l 21 by a seismic event.
22 23 Any information collection activities mentioned in this draft regulatory guide l 24 are contained as requirements in the proposed amendments to 10 CFR Part 50 25 that would provide the regulatory basis for this guide. The proposed
! 26 amendments have been submitted to the Office of Management and Budget for 27 clearance that may be appropriate under the Paperwork Reduction Act. Such l 28 clearance,,if obtained, would also apply to any information collection 29 activities mentioned in this guide.
30 31 32 33 j 34 8 Guidance is being developed in Draft Regulatory Guide DG-1017, " Pre-
! 35 Earthquake Planning- and Immediate Nuclear Power Plant Operator Post-36 Earthquake Actions," to provide plant shutdown criteria.
DG-1018 - 1 Feb 7, 1992
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- 8. DISCUSSION 2
3 Data from seismic instrumentation' and a walkdown of the nuclear power plant l
4 are used to make the initial determination of whether the plant should be shut l
5 down after an earthquake, if the plant has not already shut down from 6 operational perturbations resulting from the seismic event.'
I 8 The Electric Power Research Institute has developed guidelines that will 9 enable licensees to quickly identify and assess earthquake effects on nuclear 10 - power plants, EPRI NP-6695, " Guidelines for Nuclear Plant Response to an 11 Earthquake,"' December 1989. This regulatory guide addresses sections of 12 EPRI NP-6695 that relate to post-shutdown inspection and tests, inspection 13 criteria, inspection personnel, documentation, and long-term evaluations.
14 15 16 17 C. REGULATORY POSITION 18 19 After a plant has been shut down by an earthquake, the guidelines for 20 inspections and tests of nuclear power plant equipment and structures that are 21 specified in Sections 5.3.2 (including Tables 2-1, 2-2, and 5-1), 5.3.3 22 (includes Table 5-1),5.3.4,5.3.5, and the long-term evaluations that are 23 specified in Section 6.3 (all sections and subsections) of EPRI NP-6695 would 24 be acceptable to the NRC staff for satisfying the requirements in Paragraph 25 IV(a)(3) of Proposed Appendix 5 to 10 CFR 50.
26 27 Coincident with the long-term evaluations, the plant should be restored to its 28 current licensing basis. Exceptions to this must be approved by the Director, 29 Office of Nuclear Reactor Regulation.
30 31 2 Guidance is being developed in Draft Regulatory Guide DG-1016, Second 32 Proposed - Revision 2 to Regulatory Guide 1.12, " Nuclear Power Plant 33 Instrumentation for Earthquakes," that will describe seismic l
34 instrumentation acceptable to the NRC staff.
l l 35 8 Copies may be obtained from the Research Reports Center (RRC), Box 50490, 36 Palo Alto, California 94303.
DG-1018 - 2 Feb 7, 1992
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D. IMPLEMENTATION 2
3 The purpose of this section is to provide guidance to applicants and licensees 4 regarding the NRC staff's plans for using this regulatory guide.
5 6 This draft guide has been released to encourage public participation in its ;
7 development. Except in those cases in which the applicant proposes an '
8 acceptable alternative method for complying with the specified portions of the 9 Commission's regulations, the method to be described in the active guide 10 reflecting public comments will be used in the evaluation of applications for 11 a construction permit, operating license, combined license, or design 12 certification submitted after the implementation date to be specified in the 13 active guide. This guide would not be used in the evaluation of an 14 application for an operating license submitted after the implementation date 15 to be specified in the active guide if the construction permit was issued 16 prior to that date.
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o e 1 REGULATORY ANALYSIS 2
3 A separate regulatory analysis was not prepared for this regulatory guide.
4 The draft regulatory analysis, " Proposed Revision of 10 CFR Part 100 and 10 5 CFR Part 50," provides the regulatory basis for this guide and examines the 6 costs and benefits of the rule as implemented by the guide. A copy of the 7 draft regulatory analysis is available for inspection and copying for a fee at 8 the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC, 9 as Enclosure 2 to Secy 92-???. Single copies of the draft regulatory analysis 10 . are available from Mr. Leonard Soffer, Office of Nuclear Regulatory Research, 11 Mail Stop NL/S-324, U.S. Nuclear Regulatory Commission, Washington, DC 20555, 12 telephone (301) 492-3916 or Dr. Andrew J. Murphy, Office of Nuclear Regulatory 13 Research, Mail Stop NL/S-217A, U.S. Nuclear Regulatory Commission, Washington, 14 DC 20555, telephone (301) 492-3860.
15 i
DG-1018 - 4 Feb 7, 1992
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DRAFT STANDARD REVIEW PLAN SECTION 2.5.2 l PROPOSED REVISION 3 i
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' STANDARD REVIEW PLAN 2.5.2 PROPOSED REVISION 3 4
4 3 2.5.2 VIBRATORY GROUND MOTION 4 REVIEW RESPONSIBILITIES 5 Primary - Structural and Geosciences Branch (ESGB) l l 6 Secondary - None .
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< 7 , AREAS OF REVIEW
- 8 The Structural and Geosciences Branch review covers the 9 seismological and geological. investigations carried out to 10
- :tsblich evaluate th; ::::lersti:n for the safe sh'ttdown 11 earthquake (SSE) and th Oper: ting 50:10 carthquche (OSE) for the 3
12- site. The ::f: :hutd:en :::thqu k: i: th t :::thqu k: th:t 4s
) 13 %: d upon :n Ov:lu;ti:n ;f th :: 10 = : rthquch; p;tenti:1
- gi:n:1 :nd 1;;;l ge:1;gy :nd ::i ::1;;y :nd 14 eeS!d: ring th 15 epm %+ic-chernet:rictic: Of 10:01 cuh urf::: ::teri:1. It is th:t 16 eeeMNpH;k th:t pr:du;;; th: :: i;; vibr:t:ry gr;;nd ti:n for 1 17 chich ::fcty rel t:d :tructur;;, y:tr , nd ;;;g:nent: ::: l 18 d::igned t: ::::in functi:ncl. The :p :: ting 50:10 certhqu:he i l 30 that ::rthquch; th:t, ::ncid: ring th: ::;i:nci :nd-12::1 g;;1;;y,
- i = 1;;;, ni :pecifi h=r :teri ti== cf l== 1 zub:;rf:::
- tcri:1, ::ald :::::::bly b; : p :t:d t: Offect the plant cit:
22 during th: Oper: ting lif: ef the pl nt; it i: th:t :: thqu k: that 23 pr: duce: the vibretcry ground ::tien for which th::: fe ture: Of 24 the nue10 r p :: pl:nt nece;; ry -f;r : ntinu d perati:n without 25 undu: rich to the h::lth :nd ::f y of the publi; ar d: igned to 26 remain function:1. The SSE repreJents the potential for earthquake 27 ground motion at the site and is the vibratory ground motion for 28 which all safety related structures, systems and components are 29 designed to ensure public safety. The SSE is based upon a detailed 30 evaluation of the earthquake potential, taking into account 31 regional and local geology, seismicity, and specific 32 characteristics of local subsurface matcrial. It is defined as the 33 free-field ground response spectra at the plant site and is 34 described by horizontal and vertical response spectra corresponding 35 to the expected ground motion at the free-field ground surface or 36 a hypothetical rock outcrop.
37 Seismological and geological investigations are described in 38 Regulatory Guide DG1015, Identification and Characterization of 39 Seismic Sources. These investigations describe the seismicity of 40 the site region and correlation of earthquake activitycharacterized, with seismic 41 sources. Seismic sources are identified and 42 including the Deterministic Source Earthquake (DSE) associated with 43 each seismic source. All seismic sources, any part of which is February 10, 1992 2.5.2-1
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1- within 320 km (200 miles) of the site, must be identified. Sources 2 at larger. distances which are capable of earthquakes large enough i to af fect the site must also be identified. Seismic sources can be 3
4 4 capable tectonic sources or seismogenic sources; a seismotectonic 5 province is a type of seismogenic source.
6- The principal regulation used by the staff in determining the scope 7 and adequacy of the submitted seismologic and geologic information 8 and attendant procedures and analyses is App ndix A," "Ceisci: and 9 C;;1 gic Citing Criteri for Nucle:r P ver Plant: Appendix B, 4
10 " Criteria for the seismic' and Geologic Siting of Nuclear Power 11 Plants after (effective date]" to 10 CFR Part '100 (Ref. 1).
12 , Additional guidance (regulations, regulatory guides, and reports)
- 13 is provided to the staff through References 2 through 8.
- 14 . Specific areas of review include seismicity (Subsection 2.5.2.1),
15 geologic and tectoniccorrelation characteristics of the site and region of earthquake activity with
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16 (Subsection 2.5.2.2),
17 ' geologic structure or tectonic provinces (Subsection 2.5.2.3)',
18 maximum earthquake potential (Subsection 2.5.2.4) , seismic wave j 19 transmission characteristics of the site (Subsection 2.5.2.5), and -
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20 safe shutdown earthquake (subsection 2.5.2.6) r-end Oper ting 50:10
- rthq h ' Cub:::tien 2.5.2.?). Both deterministic and 21 22 probabilistic evaluations are used to assess the SSE.
1 23 The geotechnical engineering aspects of the site and the models and 24 methods employed in the analysis of soil and foundation response to 25 the ground motion environment are reviewed under SRP Section 2.5.4. i 26 The results of the geosciences review are used in SRP Sections 27 3.7.1 and 3.7.2.
28 II. ACCEPTANCE CRITERIA 29 The applicable regulations (Refs. 1, 2, and 3) and regulatory '
30 guides (Refs. 4, 5, and 6) and basic acceptance criteria pertinent
- 31. to the areas of this section of the Standard Review Plan are:
10 CFR Part 100, App ndix A, "C:10:10 :nd 0:01;gi Citing 32 1.
33 Crit ria fer M;;1 :: P;;;r 01:nte." Appendix B, " criteria for 34 the seismic and Geologic Siting of Nuclear Power Plants af ter 35 (effective date)." These criteria describe the kinds of geologic and seismic information needed to determine site 36 37 suitability and identify geologic and seismic factors required 38 to be taken into account in the siting and design of nuclear 39 power plants (Ref. 1).
-10 CFR - Part 50, Appendix A, " General Design Criteria for 40 2.
" Design 41 Nuclear Power Plants"; General Design Criterion 2, Bases- for Protection Against Natural Phenomena."
This 42 of the 43 criterion requires that safety-related portions 44 structures, systems, and components important to safety shall February 10, 1992 2.5.2-2 .
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' be designed to. withstand the effects of earthquakes, tsunami, and seiche without loss of capability to perform their safety
. functions'(Ref. 2).
4 3. 10 CFR Part 100, " Reactor Site Criteria. " This part describes 5 criteria that guide the evaluation of the suitability of 6 proposed sites for nuclear power and testing reactors (Ref.
7 3).
8 4.- Regulatory Guide.l.132, " Site Investigations for Foundations 9 of Nuclear-Power Plants." This guide describes programs of 10- site investigations related to geotechnical aspects that would La normally meet the needs for evaluating the safety of the site L2 from the standpoint of the performance of foundations under It 13 anticipated loading conditions including earthquake.
14 provides general guidance and recommendations for developing 15 site-specific investigation programs as well as specific 16 guidance for conducting subsurface investigations, including 17 the spacing and depth of borings as well as sampling intervals 18 (Ref. 4).
19 5. Regulatory Guide 4.7, " General Site Suitability Criteria for 20 Nuclear Power Stations." This guide discusses the major site 21 characteristics related to public health and safety which the 22 NRC. staff considers in determining the suitability of sites 2' for nuclear power stations (Ref. 5).
A 6. Regulatory Guide 1.60, " Design Respons? Spectra for Seismic 25 Design of Nuclear Power Plants." Thi: guid: gi/:: :n: :-e-t-hed 36 ::: pt:bi: t; th: M2C :::ii f : d:iining thegr rc p;nz: sp :tra 27 ;;rr;;p:nding00:to210:
th: For cxp;;ted :::i=u und :::cler: tion design purposes smoothed response ';
28 fRef. f). !
29 spectra are generally used - for example, a standard spectral 30 shape which has been used in the past is Regulatory Guide 1.60 31 (Ref. 6) . These smoothed spectra are still acceptable when an 32 appropriate peak acceleration is used as the high frequency 33, asymptote and the smoothed spectra compare favorable with site 34 specific response spectra derived from the deterministic and 35 probabilistic procedures discussed in Subsection 2.5.2.6.
36 The primary required investigations are described in 10 CFR Part 37 100,Section IV(a) of Appendix A B(Ref. 1) and regulatory guide 38 DG1015. The acceptable procedures for determining assessing the 39 seismic design bases are given in Section V(a), (b), and (c).-and The ' seismic design bases are 40 certi n "I(:) cf th; ppendix.
41 conservative determination of the SSE predicated and th: 025.
on a reasonable,As defined in Sections 4M IV and V of 10 CFR Part 100, Appendix A B(Ref. 1), the SSE nd 005 cre is based on 43 43 44 consideration of the regional and local geology and seismology and 45 on the characteristics of the subsurface materials at the site and eee is described'in terms of the vibratory ground motion that they 4F February 10, 1992 2.5.2-3
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- f. A could produce at the site. No comprehensive definitive rules can 2 be promulgated regarding the investigations needed to establish the
- 3- seismic design bases; the requirements vary from site to site.
Seismicity. In meeting the requirement of Reference l 4 2.5.2.1
! 5 1, this subsection is accepted when the complete historical record 6 of earthquakes in the region is listed and when all available ,
y 7 parameters are given for each earthquake in the historical record.
8 The listing should include all earthquakes having Modified Mercalli
- 9- Intensity (MMI) greater than or equal to IV or magnitude greater 10 than or equal to 3.0 that have been reported in all t tenic 11 pr; vin
- : for all seismic sources, any parts of which are within 12 , 320 km (200 miles) of the site. A regional-scale map should be i 13 presented showing all listed earthquake epicenters and should be
- 14 supplemented by a larger-scale map showing earthquake epicenters of 15 all known events within 80 km (50 miles) of
- the site. The 16 following information concerning each earthquake is required 17 whenever it is available: epicenter coordinates, depth of focus, 18 origin time, highest intensity, magnitude, seismic moment, source i 19 mechanism, source dimensions, distance from the site, and any 20 ' strong-motion recordings (references from which the information was 21 obtained should be identified) . All magnitude designations such as i' 22 %, M, M, M,, , etc., should be identified. In addition, any t
23 reported earthquake-induced geologic failure, such as liquefaction, l 24 landsliding, landspreading, and lurching should be described '
25 completely, including the level of strong motion that induced failure and the physical properties of the materials. The 26 27 completeness of the earthquake history of the region is determined 28 by comparison to published sources of information (e.g., Refs. 9 1 When conflicting descriptions of individual 29 through 13). ,
30 earthquakes are found in the published references, the staf f should l 31 determine which is appropriate for licensing decisions. >
- 32. 2.5.2.2 Geoloaic and Tectonic Characteristics of 2,Site and 3, and 33 Reaion. In meeting the requirements of References 1, 34 this subsection is accepted when all g::legie ntructure: uithin the 35 regi n and tectoni ::tivity seismic sources that are significant 36 in determining the earthquake potential of the region are 37 identified, or when an adequate investigation has been carried out 38 to provide reasonable assurance that all significant tectonic 39 structures seismic sources have been identified. Information 40 presented in Section 2.5.1 of the applicant's safety analysis 9 and 41 report (SAR) and information .from other sources (e.g. , Refs
.42 14 through 18) dealing with the current tectonic regime should be 43 developed into a coherent, well-documented discussion to be used as 44 the basis characterizing the earthquake-generating potential of 45 seismogenic sources and capable tectonic sources the identified 46 ge:1 gic etructures. Specifically, each t :tenic provine: seismic 47 source, any part of which is within 320 km (200 miles) of the site, 48 must be identified. The staff interprets seismotectonic provinces 49 to be regions of uniform : rthquche p t nti 1 (0i stcetni February 10, 1992
. 2.5.2-4
4 p.
I provinces) seismicity (same DSE and frequency of recurrence) l distinct from the seismicity of the surrounding area. The proposed seismotectonic provinces may be based on seismicity studies, 4 differences in geologic history, differences in the current 5 tectonic regime, etc. The staff considers that the most important 6 factors for the determination of seismotectonic provinces include 7 both (1) development and characteristics of the current tectonic in th:
8 regime of the region that is most likely reflected about 5 in the Quaternary 9 n:: tectonic: (Pcot-Mice:n: Or 10 (approximately the last. 2 million years and younger geologic 11 history) and (2) the pattern and level of historical seismicity.
12 Those characteristics of geologic structure, tectonic history, i 13 'present and past stress regimes, and seismicity that distinguish i 14 the various seismotectonic provinces and the particular areas 15 within those provinces where historical earthquakes have occurred 16 should be described. Alternative regional tectonic models derived 17 from available literature sources, including previous SARs and NRC 18 staff Safety Evaluation Reports (SERs), should be discussed. The 19 model that best conforms to the observed data is accepted. In 20 addition, in those areas where there are capable f: ult: tectonic sources, the results of the additional investigative requirements 21 22 described in 10 CTO P rt 100, App ndi% A, 00 ti:n IV(0) (3) (Ref.
23 43 r SRP 8ection 2.5.1 must be presented. The discussion should be 24 augmented by a regional-scale map showing the tectonic provin ::
locations of geologic 25 seismic ser.rces, earthquake epicenters, structures and other features that characterize the seismotectonic locations of any capable fcult tectonic provinces, and the ;
.. sources.
29 2.5.2.3 Correlation of Earthouake Activity with C0:10cic Otructur i 30 Seismocenic sources, Canable Tectonic Sources or !
31 SeismoTectonic Provinces. In meeting the requirements of Reference 32 1, acceptance of this subsection is based on the development of the 33 relationship between the history of earthquake activity and the 34 geologic :tructure: :: ::i:::tectenic provin;;: seismic sources of 35 a region. The applicant's presentation is accepted when the 36 earthquakes discussed in Subsection 2.5.2.1 of the SAR are shown to Or tectonic provine:
i 37 be associated with either g;;logi: tructur 38 capable ~ tectonic sources or seismogenic sources. Whenever an 39 earthquake hypocenter or concentration of earthquake hypocenters structures, the 40 can be reasonably correlated with geologic 41 rationale for the association should be developed considering the 2 characteristics of the geologic structure (including geologic and 43 geophysical data, seismicity, and the tectonic history) and the 44 regional tectonic model. The discussion should include the earthquake 45 identification of the methods used to locate 46 hypocenters, an estimate of their accuracy, and a detailed account 47 tho' compares And contrasts the geologic structure involved in the l 48 earthquake Occivity with other areas within the seismotectonic i
49 province. Particular attention should be given to determining the Fa capability of faults with which instrumentally located earthquake I
rebruary 10, 1992 i 2.5.2-5
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1 hypocenters are associated. l The presentation should be augmented by regional maps, all of the 2
3 same scale, showing the tectonic provine:: seismic sources, the 4 earthquake epicenters, and the locations of geologic structures and '
5 measurements used to define provinces. Acceptance of'the proposed 6 t;;tenic provin;:: seismic sources is based on the staff's 7 independent review of the geologic and seismic information.
8 2.5.2.4 Maximum Earthouake Potential and Controlline 9 'Earthauake (CE). In meeting the requirements of Reference 1, this 10 subsection is accepted when the vibratory ground motion due to the 11 i : i;; cr:dible : rth;;;he DSE ~ associated with each 5;:1;;ic 12 Otructurc ;r th: :: i;;; hi:teri; : rthq :h: :::: icted with :::h 13 '
be:teni; provimee seismic source has been assessed and when the 14 earthquake (s) that would produce the mewieum most severe The vibratory me*ieum 15 ground motion at the site has beenlargest determined.
earthquake that can 16 cr:dib1: : rthq :he DSE is the reasonably be expected to occur on a 50:1 gi tructer: given 17 Considerable 18 seismic source in the current tectonic regime. l' 19 judgement is involved in estimating the magnitude of the DSE.
20 Suggested procedures for estimating the DSE are given in Regulatory 21 Guide DG1015. C:1;;i; ::th:n : i:2010;ical cvid:n : ::y th: ::xi;;; hi teri: ::rthq rrent :
- he.
22 : xi=== ::rthq :h; 1:rg:
. Earthquakes associated with each g::1 gi Otructur: :: t :t nic 23 24 previn : seismic source must be identified. Where an earthquake is 25 associated with geologic structure, the :: inur credible : rthq :he 26 DSE that could occur on that structure should be evaluated, taking 27 into account significant factors, for example, the type of the 28 faulting, fault length, fault slip rate, rupture length, rupture 29 area, moment, and earthquake history (e.g., Refs. 19 through 22).
In order to determine the :: i;; cr:dibi : rthq h: DSE that 30
.31 could occur on those faults that are shown or assumed to be capable 32 tectonic sources, the staff accepts conservative values based on l.
33 historic experience in the region and specific considerations of ,
the earthquake history and geologic history of movement on the l
34 35 faults. Where the earthquakes are associated with a seismotectonic l'
.36 province,' the largest historic earthquake within the province
-should be identified. Isoseismal maps should also be presented for ;
37 The ground motion at the site 38 the most significant earthquakes. seismic energy 39 should be evaluated assuming appropriate l l
40 transmission effects and assuming that the :: 1 :: ::rthq :he DSE 41 associated with each ;;;1;;i tructur: cr with ::ch tc teni j l
provine: seismic source occurs at the point of closest approach of 42-43 the structure or province to the site. (Further description is 44 provided in Subsection 2.5.2.6.)
f f
l 45 The earthquake (s) that would produce the most severe vibratory If different j 46 ground motion at the site should be defined. i
! 47 potential earthquakes would produce the most severe ground motion
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February 10, 1992 2.5.2-6 i
4 a in different frequency bands, these earthquakes should be specified. The description of the potential earthquake (s) is to include the maximum intensity or magnitude and the distance from 4 the assumed location of the potential earthquake (s) to the site.
5 ~For the seismotectonic province surrounding the site, the DSE is 6 assumed to occur within 25 km of the site. The staff independently 7 evaluates the site ground motion produced by the 4 rgest rthquak l 8 DSE associated with each g:01 gi tructure er tect:.i prefi,cc l 9 seismic source. Controlling earthquakes (CE) are those earthquakes l 10 that have the greatest effect on the ground motion at the nuclear 11 power plant site. Acceptance of the description of the pet-en&ie4-12 controlling earthquake (s) that would produce the largest ground t
13 motion at the site is based on the staff's independent analysis.
l 14 2,5.2.5 Seismic Wave Transmission Characteristics of the Site.
l 15 In meeting the requirements of Reference 1, this subsection is
! 16 accepted when the seismic wave transmission characteristics
- 17 (amplification or deamplification) of the materials overlying l 18 bedrock at the site are described as a function of the significant 1 19 frequencies. The following material properties should be l 20 determined for each stratum under the site
- seismic compressional 21 and shear wave velocities, bulk densities, soil index properties 22 and classification, shear modulus and damping variations with 23 strain level, and water table elevation and its variation. In each 24 case, methods used to determine the properties should be described
?" in subsection 2.5.4 of the SAR and cross-referenced in this subsection. For the =:xi=u: ::rthquche controlling earthquake, j . determined in Subsection 2.5.2.4, the free-field ground motion i 28 (including - significant frequencies) must be determined, and an 29 analysis should be performed to determine the site effects on 30 different seismic wave types in the significant frequency bands.
31 If appropriate, the analysis should consider the effects of site 32 conditions and material property, variations upon wave propagation 33 and frequency content.
34 The free-field ground motion (also referred to as control motion) 15 should be defined to be on a ground surface and should be based on 36 data obtained in the free field. Two cases are identified l 37 depending on the soil characteristics at the site and subject to 38 availability of appropriate recorded ground-motion data. When data 39 are available, for example, for relatively uniform sites of soil or 40 rock with smooth variation of properties with depth, the control 41 point (location at which the control motion is applied) should be 42 specified on the soil surf ace at the top of the finished grade.
43 The free-field ground motion or control motion should be consistent 44 with the properties of the soil profile. For sites composed of one 45 or more thin soil layers overlying a competent material, or in case 46 of insufficient recorded ground-motion data, the control point is l 47 specified on an outcrop or a hypothetical outcrop at a location on
! 48 the top of the competent material. The control motion specified 40 should be consistent with the properties of the competent material.
I February 10, 1992 l
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Where vertically propagating shear waves may produce the maximum '
2 ground motion, a one-dimensional equivalent-linear analysis (e.g. ,
Refs. 25, 26, and 27) 3 Ref. 23 or 24) or nonlinear analysis (e.g.,
4 may be appropriate and is reviewed in conjunction with geotechnical 5 and structural engineering. Where horizontally propagating shear 6 waves, compressional waves, or surface waves may produce the 7 maximum ground motion, other methods of analysis (e.g., Refs. 28 8 and 29) may be more appropriate. However, since some of the 9 . variables are not well defined and the techniques are still in the 10 developmental stage, no generally agreed-upon procedures can be 11 promulgated at this time. Hence, To the staf f must use discretion in insure appropriateness, site 12 reviewing any method of analysis.
13 .. response characteristics determined from analytical procedures 14 should be compared with historical and instrumental earthquake 15 data, when available.
2.5.2.6 Safe Shutdown Earthauake. In meeting the 16 17 requirements of Reference 1, this subsection is accepted when the is vibratory ground motion specified for the SSE is described in terms 19 of the free-field response spectrum and is at least as conservative 20 as that which would result at the site from the acximum carthquche 21 CEs (determined in Subsection 2.5.2.4) considering the site If 22 transmission effects (determined in Subsection 2.5.2.5).
23 several different acximur potentici carthquckcc CEs produce the 24 largest ground motions in dif f.sent frequency bands (as noted in 25 Subsection 2.5.2.4), the vibratory ground motion specified for the 26 SSE must be as conservative in each frequency band as that for each 27 carthquake.
28 The staff reviews the free-field response spectra of engineering 29 significance (at appropriate damping values). Ground motion When may the 30 vary for different foundation conditions at the site.
31 site effects are significant, this review is made in conjunction 32 with the review of the design response spectra ir hection 3.7.1 to ensure consistency with the free-field motion ne staff normally 33 34 evaluates response spectra on a case-by-c7su basis. The staff 35 considers compliance with the following conditions acceptable in 36 the evaluation of the SSE. In all these procedures, the proposed 37 f ree-field response spectra shall be considered acceptable if they 38 equal or exceed the estimated 84th percentile ground-motion spectra described in 39 from the maximum cr controlling Ocethquckc CE 40 Subsection 2.5.2.4.
41 The following steps summarize the staff review of the SSE.
42 1. Both horizontal and vertical component site-specific response should be developed statistically from response 43 spectra 44 spectra of recorded strong motion records that are selected to 45 have similar source, propagation path, and recording site 46 properties as the controlling earthquake (s). It must be 47 ensured that the recorded motions represent free-field February 10, 1992 2.5.2-8
4 ,
1 conditions and are free of or corrected for any soil-structure interaction effects that may be present because of locations 4
and/or housing of recording instruments. Important source properties include magnitude and, if possible, fault type, and 5 tectonic environment. Propagation path properties include G distance, depth, and attenuation. Relevant site properties 7
8 include shear the amplitude velocity profile and other factors that affect of waves at different frequencies.
10 9
sufficiently large number of site-specific time historiesA 11 and/or response spectra should be used to obtain an adequately broadband 12 parameters.
spectrum to encompass the uncertainties in these 13 An 84th percentile response spectrum for the 14
- records should be presented for each damping value of interest 15 and compared to the SSE free-field and design response spectrum (e.g., Refs. 30, 31, 32, 16 and 33). The staff 17 considers direct estimates of spectral ordinates preferable to scaling of spectra to peak accelerations.
18 19 United States, relatively little information isInavailable the Eastern on 20 magnitudes for the larger historic earthquakes; hence, it may 21 be appropriate to rely on intensity observations (descriptions of earthquake effects) to estimate magnitudes of historic 22 events (e.g. , Refs. 34 and 35) . If the data for site-specific 23 24 response spectra were not obtained under geologic conditions 25 similar to those at the site, corrections for site effects 26 should spectra.
be included in the development of the site-specific 28 2.
Where a large enough ensemble of strong-motion records is not 29 available, response spectra may be approximated by scaling 30 that ensemble of strong-motion data that represent the best 31 estimate of source, propagation path, and site properties 32 (e.g., Ref. 36). Sensitivity studies should show the effects of scaling.
33 3.
34 If strong-motion records are not available, site-specific peak 35 ground acceleration, velocity, and displacement (if necessary) 36 should be determined for appropriate magnitude, distance, and foundation conditions. Then . response spectra may be 37 determined by scaling the acceleration, velocity, and 38 displacement values by appropriate amplification factors 39 (e.g., Ref. 37). Where only estimates of peak ground 40 41 acceleration are available, it is acceptable to select a peak 42 acceleration and use this peak acceleration as the high 43 frequency asymptote to standardized response spectra such as 44 described in Regulatory Guide 1.60 (Ref. 6) for both the horizontal and vertical components of motion with the 45 appropriate amplification factors.
46 For each controlling
- 47 earthquake, the peak ground motions should be determined using 48 current relations between acceleration, velocity, necessary, and, if displacement, earthquake size (magnitude or 49 intensity) , and source distance. Peak ground motion should be l
l February 10, 1992 2.5.2-9
e t Relationships 1
determined f rom state-of-the-art relationships. for example, in )
a between magnitudo 'and ground motion are found, 3 References 38, 39, 40, and 41 and relationships between ground motion and intensity are found, for example, in References 41, 4
5 42, and . 4 3. : Due to the limited' data for high intensities VIII, the greater than - Modified. Mercalli Intensity (MMI) i 6 i 7
available empirical ~ relationships between intensity and peak for determining the l 8 ground motion may not be . suitable 9 appropriate reference acceleration for seismic design.
10 4. Response spectra developed by theoretical-empirical modeling 11-of ground motion may . be used to supplement' site-specific spectra if the input parameters and the appropriateness of the 12 13
- model are thoroughly documented (e.g., Refs. 19, 44,- 45-and 46, and 53). Modeling is particularly useful for sites near capable f ult: tectonic sources or for deeper structures that 14 may experience ground motion that is different in-terms of 15 16 ;
17 frequency content and wave type from ground motion caused by 18 more distant. earthquakes..
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19 5. Probabilistic estimates of seismic hazard should be calculated
- 20. (e.g., Refs. 41 and 47) and the underlying assumptions and 21 associated uncertainties should be documented to assist in the 22 staff's overall deterministic approach. The probabilistic ,
23 studies should highlight which seismic sources are significant to the site. "nifer; h:: rd :p :tr: ' pectra th:t h:ve : ,
24 l 25 unifer: pr:bability Of n ::d:n : ;ver th: frequency range ef 26 intercet) chsuing unc;rt:inty ch:21d bc : lcul ted for 0.01, 27 0.001, nd 0.0001 n70:1 pr b bilitic: ;f ex:: d ne: Ot th: ,
28 cit;. The probability of exceeding the SSE response spectra !
29 should also be estimated and comparison of results made with are
~
30 other probabilistic studies. Suggested procedures 31 contained in DG1015.
I 32 The time duration and number of cycles of strong ground motion is 33 required for analysis of site foundation liquefaction potential and 34 for design of many plant components. The adequacy of the time
- 35 history for structural analysis is reviewed under SRP Section 3 . 7 .' 1. The time history is reviewed in this SRP section to confirm l
36 j- 37 that it is compatible with the seismological and geological !
l 38 conditions in the site vicinity and with the accepted SSE model.
39- At present, models for deterministically computing the time history l
j 40 of strong ground motion from a given source-site configuration may
-41 be limited. It is therefore acceptable to use an ensemble of 42 ground-motion time histories from earthquakes with similar size, or 43 site-source characteristics, and spectral characteristics ;
44 results of a statistical analysis of such an - ensemble. Total 45 duration of the motion is acceptable when it is as conservative as 46 values determined using current studies such as References 48, 49, ;
47 50, and 51.
February 10, 3992' 2.5.2-10
,- r. _ -- - - - - - - _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _______ _
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2.5.2."
requirc ent: Of 2 ference 1, 02 rstin: 0 :i: Earthruck:. In ::: ting the thi: sub:::tien 1: ::::pt:bic when the vibratory gr und :: tion for th: OSS i; d:: ribed and the resp;ns:
4 sp :tru: (t pproprict: d:: ping v 1ues) at th: cit: specified.
5 Probability icul: tion: (e.g., Refs. 41, 4", and 52) ch;uld be i 6 u::d t estinct: the prebebility of ex;;; ding the OSE during th:
7 Operating life of-the pl:nt. The ::ximus vibratory ground r.ction 8
of the OSS chculd b: :t les:t en: h:1f th: ::ximu vibratory ground 9 =ction of the CCE unle:: lever 000 :n be justified On the b :i:
10 of probability calculation:. It h:: b n Ot:ff pr etic: t :: pt 11 th: OSE if the return pcried i: On th: Ord: Of hundr d: Of 7:Or:
12 (0 5., 2 f. 31).
l 13 -
III. REVIEW PROCEDURES 14 Upon receiving the applicant's SAR, an acceptance review is 15 conducted to determine compliance with the investigative 16 requirements - of 10 CFR Part 100, Appendix A B (Ref. 1). The 17 reviewer also identifies any site-specific problems, the resolution 18 of which could result in extended delays in completing the review.
19 After SAR acceptance and docketing, those areas are identified 20 where additional information is required to . determine the
.21 - earthquake hazard. These are transmitted to the applicant as draf t 22' requests for additional information.
A site visit may be conducted during which the reviewer inspects the geologic conditions at the site and region around the site as 25 shown in outcrops, borings, geophysical data, trenches, and those 26 geologic conditions exposed during construction if the review is 27 for an operating license. The reviewer also discusses the 28 questions with the applicant and his consultants so that it is 29 clearly understood what additional information is required by the 30 staff to continue the review. Following the site visit, a revised 31 set of requests for additional information, including any i 32 additional questions that may have been developed during the site 33 visit, is formally transmitted to the applicant.
34 The reviewer evaluates the applicant's response to the questions, l
35 prepares requests for additional clarifying information, and 36 formulates positions that may agree or disagree with those of the 37- applicant. These are formally transsitted to the applicant.
38 The safety analysis report and amendments . responding to the i 39 requests for additional information are reviewe'd-to determine that 40 the information presented by the applicant is acceptable according 41 to the criteria described in Section II (Acceptance Criteria) 42 above. Based on information supplied by the applicant, obtained 43 from site visits or.from staff consultants or literature sources, 44 the reviewer independently identifies and evaluates the relevant i
l 45 ::i:::t;;tenic pr; vin;;; seismogenic sources and capable tectonic I
February 10, 1992 f 2.5.2-11
'9 RO
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sources, evaluates the capability of faults in the region, and i 2
I determines the earthquake potential for each province and :::h 3 ::p:ble f;;1t er tecteni: :tructure seismogenic source or capable tectonic source using procedures noted in Section II (Acceptance 4
5 Criteria) above. .The reviewer evaluates the vibratory ground 6
motion that the pct:nti:1 ::rthquake: controlling earthquakes could f 7 produce at the site and define: compares that ground motion to the 8
safe shutdown earthquake- nd Operating besi: : rthquake.
9 IV. EVALUATION FINDINGS 10 If the evaluation by the staf f, on completion of the review confirms that of the i 11, .' geologic and seismologic aspects of the plant site, -
the applicant has met the requirements or guidance of applicable portions of References 1 through 6, the conclusion in the SER 12 13 14 states that the information provided and investigations performed 15 support the applicant's conclusions regarding the seismic integrity .
16 of the subject nuclear power ~ plant site. In addition to the l' 17- conclusion, this section of the SER includes (1) d;finition: an evaluation of tcetenic pr rir.cc: seismogenic sources and capable tectonic sources; (2) evaluations of the capability of geologic 18 19 20 structures in the region; (3) deter in: tion: evaluation of the'GGE spectra . based -on 21 : rthq Ch:( t DSEs and free-field response 22 evaluation of the p;t nti:1 controlling earthquakes; and (4) time ,
23 history of strong ground motion, nd (5) detcr;incti n Of th: 005 about any fre -field rc:penz: p::tr . Staff reservations ;
24 :
25 significant deficiency presented in the applicant's SAR are stated 26 in sufficient detail to make clear the precise nature of the 27 concern. The above evaluation determinations or redeterminations
[
28 are made by the staf f during both the construction permit (CP) and operating license (OL) phases of review.
29 30 OL applications are reviewed for any new information developed 31 subsequent to the CP safety evaluation report (SER). The review '
' 32 will' also determine whether the CP recommendations have been 33 implemented. '
A typical OL-stage summary finding for this section of the SER l
l 34 ;
35 follows 36 In our review of the seismologic aspects of the plant site we j 37 have conside' red pertinent information gathered since our 38- initial seismologic review which was made in conjunction with y 39 the issuance of the Construction Permit.site This new and information near-site 40
~
includes . data gained from both '
l l 41 investigations as well as from a review of recently published 42 literature.
of our recent review of the seismologic 43 As a result 44 information, we have determined that our earlier conclusion 45 regarding the safety of the plant from a seismological February 10, 1992 2.5.2-12 i
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l standpoint remains valid. These conclusions can be summarized as follows:
3 1. Seismologic information provided by the applicant and 4 required by Appendix A B to 10 CFR Part 100 provides an 5 adequate basis to establish that no capabic fault:
6 seismic sources exist in the plant site area which would 7 cause earthquakes to be centered there.
8 2. The response spectrum proposed for the safe shutdown 9 earthquake is the appropriate free-field response 10- spectrum in conformance with Appendix A B to 10 CFR Part li .
100.
^12 The new information reviewed for the proposed nuclear power 13 plant is discussed in Safety Evaluation Report Section 2.5.2.
14 The staff concludes that the site is acceptable from a l 15 seismologic standpoint and meets the requirements of (1) 10 l 16 CFR Part 50, Appendix A (General Design Criterion 2), (2) 10 CFR Part 100, and (3) 10 CFR Part 100, Appendix A B. This 17 i la conclusion is based on the following:
l l 19 1. The ppplicant has met the requirements of:
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- a. 10 CFR Part 50, Appendix A (General Design i Criterion 2) with respect to protection" against l .. natural phenomena such as faulting.
l
, 23 b. 10 CFR Part 100 (Reactor Site Criteria) with l 24 respect to the identification of geologic and 25 seismic information used in determining the i 26 suitability of the site.
l 27 c. 10 CFR Part 100, Appendix A 'Scic=ic cnd Cecicgic l l 28 Citing Critcric fcr Muclear rcuer Plantc) Appendix
- 29 B (Criteria for the seismic and Geologic siting of 30 Nuclear Power Plants after (effective Date]) with seismic 31 respect to obtaining the geologic and 32 information necessary to determine (1) site l
l 33 suitability and (2) the appropriate design of the ,
34 plant. Guidance for complying with this regulation is contained in Regulatory Guide 1.132, " Site 35 36 Investigations for Foundations of Nuclear Power 37 Plants," Regulatory Guide 4.7, " General Site Suitability for Nuclear Power Stations," and 38 39 Regulatory Guide 1.60, " Design Response Spectra for r 40 Seismic Design of Nuclear Power Plants."
41 V. IMPLEMENTATION i
Februaqr 10, 1992 2.5.2-13
'^
e o 1 The following is intended to provide guidance to applicants and 2 licensees regarding the NRC staff's plans for using this SRP i 3 section.
4 Except in those cases in which the applicant / licensee proposes an 5 acceptable alternative. method for complying with specific portions 6 of the Commission's regulations, the methods described herein will 7 be used by the staff in its evaluation of conformance with 8 Commission. regulations.
9 Implementation schedules for conformance to parts of the method 10 discussed herein are contained in the referenced regulatory guides 11 . and NUREGs (Refs. 4 through 8).
12 The provisions of this SRP section apply to reviews of construction 13 permit (CP), operating license (OL), preliminary design approval 14 (PDA), final design approval (FDA), and combined license (CP/OL) 15 applications docketed after the date of issuance of .this SRP 16 section.
17 VI. REFERENCES 10 CFR Part 100, Appendix *
" Sci:=ic :nd C01cgic Siting 18 1.
19 Criteri: fcr Nucle:r Peter P1:nts." Appendix B, " Criteria for 20 the seismic and Geologic Siting of Nuclear Power Plants After 21 [ Effective Date)."
10 CFR Part 50, Appendix A, General Design Criterion 2, 22 2.
23 " Design Bases for Protection Against Natural Phenomena."
24 3. 10 CFR Part 100, " Reactor Site Criteria."
25 4. Regulatory Guide 1.132, " Site Investigations for Foundations 26 of Nuclear Power Plants."
27 5. Regulatory Guide 4.7, " General Site Suitability Criteria for 28 Nuclear Power Stations."
29 6. Regulatory Guide 1.60, " Design Response Spectra for Seismic 30 Design of Nuclear Power Plants."
31 7. Regulatory Guide 1.70, " Standard Format and Content of Safety 32 Analysis Reports for Nuclear Power Plants." 1 33 8. NUREG-0625, " Report of Siting Policy Task Force" (1979).
34 9. NUREG/CR-1577, "An Approach to Seismic Zonation for Siting
- 35 Nuclear Electric Power Generating Facilities in the Eastern 36 United States," prepared by Rondout Associates, Inc. Barstow, , for the U.S. Nuclear Regulatory Commission. Authored by N.
37 38 K. Brill, O. Nuttli, and P. Pomeroy (1981).
February 10, 1992 2.5.2-14
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l l
- 10. C. W. Stover et al., 1979-1981, Seismicity Maps of the States I of the U.S., Geological Survey Miscellaneous Field Studies
. Maps. '
4 5
- 11. " Earthquake History of'the United States," Publication 41-1, National Oceanic and Atmospheric Administration, U.S.
6 Department of Commerce (1982).
7 12. T. R. Toppozada, C. R. Real, S. P. Bezore, and D. L. Parke, 8 " Compilation of Pre-1900 California Earthquake History, Annual 9- Technical Report-Fiscal Year 1978-79, Open File Report 79-6
(. .10 SAC (Abridged Version)," California Division of Mines and j 11 ,
Geology (1979).
12 13. P. W. Basham, D. H. Weichert,- and M. J. Berry, " Regional 13 Assessment of Seismic Risk in Eastern Canada," Bulletin 14 Seismological Society of America, Vol. 65,. pp. 1567-1602 15 (1979).
16 14. P. B. King, "The Tectonics of North America - A Discussion to 17 Accompany the Tectonic Map of North America, Scele 18 1:5,000,000," Professional Paper 628, U.S. Geological Survey 19 (1969).
l 20 15. A. J. Eardley, ' Tectonic Divisions of Ndrth America," Bulletin
~~
American Association of Petroleum Geologists, Vol. 35 (1951).
.. 16. J. B. Hadley and J. F. Devine, "Seismotectonic Map of the ,
l 23 Eastern United States," Publication MF-620, U.S. Geological '
24 Survey (1974).
i 25 17. M. L. Sbar and L. R. Sykes, " Contemporary Compressive Stress 26 and Seismicity in Eastern North -America: An Example of Intra-27 Plate Tectonics," Bulletin Geological Society of America, Vol.
28 84 (1973).
L 29 18. R. B. Smith and M. L. Sbar, " Contemporary Tectonics and 30 Seismicity of,the Western United States with Emphasis on the 31 .Intermountain Seismic Belt," Bulletin Geological Society of 32 America, Vol. 85 (1974).
33 19. NUREG-0712, " Safety Evaluation Report (Geology and Seismology) l 34' Related to the Operation of San Onofre Nuclear Generating 35 Station, Units 2 and 3" (1980).
36 20. D. B. Slemmons, " Determination of Design Earthquake Magnitudes <
" 37 for - Microzonation,". Proceedings of the Third International
-38 Earthquake Microzonation Conference (1982).
l 39 21. M. G. Bonilla, R. K. Mark, and J. J. Lienkaemper, " Statistical da Relations Among Earthquake Magnitude, Surface Rupture, Length February 10, 1992 2.5.2-15
o o and Surface Fault Displacement," Bulletin of the Seismological 1
2 Society of America, Vol. 74, pp. 2379-2411 (1984).
T. C. Hanks and H. Kanamori, "A Moment Magnitude Scale,"
2348-2350 3 22. Vol. 84, pp.
4 Journal -of Geophysical Research, 5 (1979).
B. Seed, " SHAKE-A Computer 6 235 P. B. Schnabel, J. Lysmer, and H.
7 Program for Earthquake Response Analysis of Horizontally Earthquake Engineering 8 Layered Sites," Report No. EERC 72-12, Berkeley (1972).
9 Research Center, University of California, 10 24.
E. Faccioli and J. Ramirez, " Earthquake Response of Nonlinear 11
- Hysteretic Soil Systems," International Journal4,of pp.Earthquake 261-276 12 Engineering and Structural Dynamics, Vol.
13 (1976).
14 25. I'.
V. Constantopoulos, " Amplification Studies for a Nonlinear 15 Hysterstic soil Model," Report No. R73-46, Department of Civil
-16 Engineering,. Massachusetts Institute of Technology (1973) .
- 26. V. L. Streeter, E. B. Wylie, and F. E. Richart, " SoilAmerican Motion 17 Methods," Proc.
Computation by Characteristics 18 19 . Society of Civil Engineers, Journal of the Geotechnical 20 Engineering Division, Vol. 100, pp. 247-263 (1974).
B. Joyner and A. T. F.
Chen, " Calculations of Nonlinear 21 27. W.
Earthquakes," Bulletin Seismological Ground Response in 22 23 Society of America, Vol. 65, pp. 1315-1336 (1975).
Seed, _ " Dynamic Response of f
24 28. T. Udaka, J. Lysmer, and H. B.
25 Horizontally Layered Systems Subjected to Traveling Seismic on Earthquake Waves," Proc. 2nd U.S. National Conf.
26 27 Engineering.(1979).
28 29. L. A'.' Drake, " Love and Raleigh Waves in an Irregular Vol. Soil 70, 29 Layer," Bulletin Seismological Society of America, 30 pp. 571-582 (1980).
31 30. NUREG/CR-4861, " Development of Site-Specific Response Spectra" 32 -(1987).
33 31. NUREG-0011, " Safety Evaluation Report Related to operation of 34 Sequoyah Nuclear Plant, Units 1 and 2" (1979).
35 32. NUREG-0793, " Safety Evaluation Report Related to the Operation 36 of Midland Plant, Units 1 and 2" (1982).
+
37 33. NUREG-0847, " Safety Evaluation Report Related to the Operation 38 of Enrico Fermi Atomic Power Plant, Unit No. 2" (1981).
February 10, 1992 2.5.2-16
- 34. R. L. Street and F. T. Turcotte, "A Study of Northe:: stern North American Spectral Moments, Magnitudes, and Intensities,"
4 Bulletin 614 Seismological Society of America, Vol. 67, pp. 599-(1977).
5 35. O. W. Nuttli, G. A. Bollinger, and D.
6 W. Griffiths, "On'the 7 Relation Between Modified Mercalli Intensity and Body-Wave Magnitude," Bulletin Seismological Society of America, Vol.
8 69, pp. 893-909 (1979).
9 36. T. H. Heaton, F. Tajima, and A. W. Mori, " Estimating Ground 10 11 Motions Vol. 8, Using Recorded Accelerograms" Surveys in Geophysics, pp. 25-83 (1986).
12 37.
13 NUREG/CR-0098, " Development of Criteria for Seismic Review of Selected Nuclear Power Plants" (1978).
14 38. W. B. Joyner and O. M.
15 Boore, " Peak Horizontal Acceleration 16 and the Velocity from Strong Motion Records Including Records from 17 1979 Imperial Valley, California Earthquake," Bulletin Seismological Society of America, Vol. 71, 2011-2038 (1981).
18 39. K. W. Campbell, "Near-Source Attenuation of Peak Horizontal 19 20 Acceleration," Bulletin Seismological Society of America, Vol.
71, pp. 2039-2070 (1981).
- 40. O. W. Nuttli and R. B. Herrmann, " Consequences of Earthquakes e
23 in the Mississippi Valley," Preprint 81-519, American Society of Civil Engineers Meeting, 14 pp. (1981).
24 41.
25 NUREG/CR-5250, " Seismic Hazard Characterization of 69 Nuclear Plant Sites East of the Rocky Mountains" (1989).
26 42.
27 M. D. Trifunac and A. G. Brady, "On the Correlation of Seismic 18 Intensity Scales with Peaks of Recorded Strong Ground Motion,"
Bulletin Seismological Society of America, Vol. 65 (1975).
19 43.
lo NUREG-0402, " Analysis of a Worldwide Strong Motion Data Sample 31 to Develop an Improved Correlation Between Peak Acceleration, 32 Seismic Intensity and Other Physical Parameters," prepared by Computer Sciences Corporation for the U.S. Nuclear Regulatory Commission. Authored by J. R. Murphy and L. J. O'Brien as (1978).
35 44.
36 NUREG-0717, " Safety Evaluation Report Related to the Operation of Virgil C. Summer Nuclear Station, Unit No. 1" (1981).
37 45.
38 NUREG/CR-1340, " State-of-the-Art Study concerning Near-Field Earthquake Ground Motion" (1980).
39 46.
NUREG/CR-1978, " State-of-the-Art Study Concerning Near-Field February 10, 1992 2.5.2-17
.* n k 1 Earthquake Ground Motion" (1981).
2 47. " Seismic Hazard Methodology for the Central and Eastern United 3
States," Electric Power Research Institute, Report NP-4726 4 (1986).
5 48. R. Dobry, I. M. Idriss, and E. Ng, " Duration Characteristics 6
of Horizontal Components of Strong-Motion Earthquake Records,"
7.
Bulletin Seismological Society America, Vol. 68, pp.1487-1520 8 (1978). t
'9 49. B. A. Bolt, " Duration of Strong Ground Motion," Proceedings of 10 the Fifth World Conference on Earthquake Engineering (1973).
11 50. W. W. Hays,
" Procedures for Estimating Earthquake Ground
'12 . Motions," Professional Paper 1114, U.S. Geological Survey 7 13 (1980).
H. Bolton Seed, I. M. Idriss, F. Makdisi, and N. Banerjee, Irregular Stress Time Histories by 14 51.
15 " Representation of 16 Equivalent' Uniform Stress Series in Liquefaction Analysis,"
17 National Science Foundation, Report EERC 75-29, October 1975.
Perkins, P. C. Thenhaus, S. L.
18 52. S. T. Algermissen, D. M.
19 Hanson, and B.'L. Bender, "Probabilistic Estimate of Maximum 20 Acceleration and Velocity in Rock in the Contiguous United 21 States," U. S. Geological Survey Open-File' Report 82-1033 22 (1982).
NUREG-0675, Supplement No. 34, " Safety Evaluation Report 23 53.
24 Related to the Operation of Diablo Canyon Nuclear Power Plant, 25 Units 1 and 2", (1991).
26 y
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! February 10, 1992 l 2.5.2-18
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ENCLOSURE 12 M
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i DRAFT PUBLIC ANN 0UNCEMENT l The Nuclear Regulatory Commission-(NRC) announced that it is issuing proposed regulations ~to amend and to update the criteria _used in decisions re.garding l l power reactor siting, including geologic, seismic, and earthquake engineering I
considerations for future nuclear power plants. Existing reactor licensees .
L 'would be unaffected b)Fthese proposed changes. The proposed revisions would l l allow the NRC to benefit from experience gained in the application of the I procedures and methods used in the current regulation and to incorporate ]
advancements in.the earth sciences and earthquake engineering since the !
regulation was issued in 1973. '
L The proposed regulation primarily consists of two separate changes, namely, L the source term and dose considerations, and the seismic and earthquake i engineering considerations of reactor siting. The source-term and dose: :
revisions would eliminate the use'of a source term and dose calculations to I fix the size cf the exclusion area. lnstead, a minimum exclusion area i distance from the reactor to the exclusion area boundary of 0.4 miles would be l required. Population density criteria around nuclear power reactor sites i would also be incorporated into the regulation. '
In_ the seismic area, both probabilistic and deterministic evaluations would be employed. The Safe Shutdown Earthquake (SSE) would be employed in plant design, whereas the Operating Basis Earthquake (OBE) would require a plant ;
!. shutdown and inspection, were it to occur. l l
The Commission is issuing the proposed revisions in the Federal Reaister for a '
ninety-day public comment period.
l, l
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- 4 m- A--A- nA-- Dwm .+ - n - ~-s. -a sa~ ~ asw.,-- 2 ENCLOSURE 13 l
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~g UNITED STATES l ['
y g
E NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 -
~s.,**.../ i i
The Honorable Peter H. Kostmayer, Chairman 1 Subcommittee on Energy and the Environment Committee on Interior and Insular Affairs United States House of Representatives l Washington, DC 20515 l Dear Mr. Chairman- l t
j Enclosed for the information of the Subcommittee are copies of a public announcement and a proposed revision to Title-10 of the Code of Federal L Regulations which is to be published in the Federal Reaister.
l The Nuclear Regulatory Commission is proposing to amend its regulations to l update the criteria used in decisions regarding power reactor siting, i
including geologic, seismic, and earthquake engineering considerations for l future nuclear power plants. The proposed revisions would allow the NRC to i benefit from experience gained in the application of the procedures and '
i methods set forth in the current regulation and to incorporate the rapid l advancements in the earth sciences and earthquake engineering.
l The proposed regulation pri$arily consists of two separate changes, namely, the source term and dose considerations, and the seismic and earthquake engineering considerations of reactor siting. The source term and dose revisions would eliminate the use of a source term and dose calculationt, to fix the size of the exclusion area. Instead, a minimum exclusion area distance from the reactor to the exclusion area boundary of 0.4 miles would be required. Population density values around nuclear power reactor sites would also be incorporated into the regulation.
In the seismic area, both probabilistic and deterministic evaluations would be l
employed. The Safe Shutdown Earthquake (SSE) would be employed in plant
! design, whereas the Operating Basis Earthauake (0BE) would require a plant l shutdown and inspection, were it to occur.
The Commission is issuing the proposed revisions for a ninety-day public l- comment period. i Sincerely, Dennis K. Rathbun, Director Office of Congressional Affairs
Enclosures:
- 1. Public Announcement
- 2. Federal Register Notice cc: Representative John J. Rhodes
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l ENCLOSURE 14 1
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, , o UNITED STATES '
! o NUCLEAR REGULATORY COMMISSION 3 E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o,, wAssiNcTON, D, C. 20555
%, f,8 January 15, 1992 >
I The Honorable Ivan Selin
. Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555 f-
Dear Chairman Selin:
SUBJECT:
PROPOSED 10 CFR'PART 50 AND PART 100 (NONSEISMIC) RULE
- l CHANGES AND PROPOSED UPDATE OF SOURCE TERM l During the 381st meeting of the Advisory Committee on Reactor Safeguards, January 9-11, 1992, we discussed the NRC staff proposal to decouple nonseismic siting requirements from plant requirements through revisions to 10'CFR Part 50 and Part 100 and the proposal I
~
to update the fission product source term used for siting analyses.
Our Subcommittees on Safety Philosophy, Technology and criteria; l
Severe Accidents; and Regulatory Policies and Practices discussed j these matters during a joint meeting on January 7-8, 1992. During l these meetings, we had the benefit of discussions with representa-l tives of the NRC staff and industry and the documents referenced.
i The staff has proposed separation of Part 100, " Reactor Site Criteria," from those requirements for - plant design which more
, properly belong in Part 50. A two-stage program to accomplish this l has been described. In Stage 1, radiological dose criteria and
! reference to fission product release quantities (the " source term")
from TID-14844 will be moved to 10 CFF. 50.34 without.other change.
Also, Part 100 will be augmented by adding to it certain quantita-tive criteria, without change, now specified in Regulatory-duide 4.7, " General Site Suitability Criteria for Nuclear Power Sta-tions."
In Stage 2, further changes will be made in Part 50 to update source term requirements by incorporating technical information about severe accidents developed since Part 100 was promulgated in 1962. Details of all changes to Part 50 have not been developed, but a preliminary description of the source term itself, and its derivation, was provided. This new source term will ultimately be described in a technical document which will be referenced in the
! regulations instead of TID-14844. The staff plans to issue this j technical document in advance of further work on the Part 50 change j so that it can be reviewed by industry and the reactor safety
! community. The present plan is to issue it for public comment in April 1992.
l
4 S '
The Honorable Ivan Selin 2 January 15, 1992 6
We were told that Stage 1 work is progressing on a. schedule which f' should make the revised rules available prior to completion of the review-of evolutionary plant applications and that-schedules for Stage 2 work are compatible with,those for passive plant reviews.
We believe the first stage proposal is reasonable and should proceed as the staff indicated. For Stage 2, development of a new i surrogate source _ term 'is also proceeding along the right line.
However, a major part of the source term proposal, a description of mechanisms for depletion of the source term within containment by '
engineered safety features and natural processes, remains to be developed. Also, confirmation of details of the proposal through the public comment period is needed. Beyond that, we have major concerns about the Stage 2 program, not so much with what is being done as with what is not. '
e There is no plan for a Stage 2 upgrade of Part 100. Stage 1 merely provides a more logical arrangement and more completely codifies technical information on siting which was developed 30 years ago. The basis for key requirements such a the 0.4-mile radius for an exclusion zone, a 10-mile radius for an emergency planning zone, and a maximum population density for the low population zone has not been reexamined or justified with up-to-date information. We question the appropriateness of codifying Regulatory Guide 4.7 guidance.
While Stage 1 is acceptable in the interlm, further work i should be carried out to consider what other changes may be appropriate in light of the large amount of experience and information that has been accumulated since 1962. In particu-lar, the relation of these requirements to the commission's Safety Goal Policy should be established.
e There is no plan to incorporate meteorological requirements in Part 100. Enough is now known about meteorology to define, in advance, .that certain sites - would be unacceptable ~ for a nuclear power plant.
e While the source term is an important part of Part 50 require-ments for containment performance, it is not the most impor-tant part by any means. Far more significant to risk are characteristics which will govern whether a containment will continue to perform its function or fail during a severe accident and whether mitirjation s;' stems will operate effec-tively or not. The staff has done a good job.in its prelimi-nary development of the fission product source term discussed above. This source term is, in essence, a surrogate for the spectrum of possible fission product releases defined in a way so that it can be of practical use in containment design.
This surrogate was developed using engineering judgment and the information about severe accidents and fission product 1 .
I
- 6. ' ev .
The Honorable Ivan Selin 3 January 15, 1992 releases developed over the last decade or.so. The ACRS recommended in a May 17, 1991 report that criteria to accommo-date severe accident phenomena in containment design. be developed. This report recommended use of an " energy source term" which would be a surrogate:for'the spectrum of severe accident challenges to. containment. integrity.
- Another critical issue for designers, in an updated source term, is the-timing of fission product releases and how that timing relates to the requirements for closure times of large containment isolation valves. In existing plants, such valves are required to close.in about 5 seconds. Preliminary work described by the staff shows that closure times of 10-30 seconds could easily be justified. Further work may indicate minimum closure times could be increased to 1 minute or more.
If required closure times can be-increased enough, it may be possible to justify a class of valves for use in containment isolation which will be more generally reliable. This work i should be pursued with that object in mind.
l The staff should be complimented on the work it has begun to decouple siting and plant design criteria and to update source term requirements. However, the additional areas we have described should also be covered as the program progresses. The Committee' will be interested'in following this work as it develops.
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Sincerely,
. K David A. Ward Chairman Referenc 4:
- 1. Mensrandum dated October 11, 1991 from Themis P. Speis,
,, Nuclear Regulatory Commission, for Raymond F. Fraley, Advisory Committee on Reactor Safeguards,
Subject:
Proposed Revision of 10 CFR.Part 100, Reactor Site Criteria, Revisions to 10 CFR Part 50, New Appendix B to 10 CFR Part 100 and Appendix S to 10 CFR Part 50, and Associated Regulatory Guides (Draft Predecisional)
- 2. Memorandum dated December 11, 1991 from Warren Minners, Nuclear Regulatory Commission, for Raymond F. Fraley, Advisory Committee on Reactor Safeguards,
Subject:
Revision of In-
!- Containment Accident Source. Terms for Nuclear Power Plants
[ (Draft Predecisional)
[ 3. Report dated May 17, 1991 from David A. Ward, Chairman,
} Advisory Committee on Reactor Safeguards, to Kenneth M. Carr,
- i. Chairman, Nuclear Regulatory Commission,
Subject:
. Proposed
! Criteria to Accommodate Severe Accidents in Containment Design 1
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e 79d N4ti CU U-il-CO pA K8e UNITED STATES NUCLEAR REGULATORY COMMIS$10N l ADVlaORY COMMITTEE ON REACTOR SAFEGUARDS J WASHINGTON. D, C. 20666 February 14, 1992 The Honorable Ivan Selin Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Dear chairman salins
SUBJECT:
PROPOSED REVISIONS TO 10 CFR PARTS 50 AND 100 AND PROPOSED REGUIATORY GUIDES REIATING TO SEISMIC UTING AND EARTHQUAKE ENGINEERING CRITERIA During the 382nd meeting of the Advisory Committee on Reactor Safeguards, February 6-8, 1992, we completed our initial review of those proposed revisions to reactor siting regulations that deal with seismology, geology, and earthquake engineering. These matters were considered also during meetings of our Extreme External 5, 1992.Phenomena Subcommittee on December 10, 1991, and February During these meetings, we had the benefit of discussions with representatives of the NRC staff and the industry.
The proposed revisions are to be submitted to the Commission as part of a package intended to decouple siting from plant design.
Our report of January 15, 1992, provided comments on those portions of the package relating to the nonseismic revisions to 10 CFR Parts 50 and 100 and to the source ters. The specific revisions covered by this report are those referenced at the end of this report.
The existing requirements in 10 CFR Part 100 and its Appendix A remain in effect for all plants licensed prior to the effective date of the proposed revisions. For future plants and sites, the seismic siting portion of Part 100 will be included in a new subpart B addition, a(the newexisting Appendixrequirements will become B will be referenced subpart B.
in Subpart A) . In This new appendix those :,n Appendix A.
will contain much less detailed requirements than The identification and characterization of seismic sources and procedures for the selection of a safe shutdown Earthquake and (SSE) will be covered in a new Regulatory Guide DG-1015,
- the engineering criteria for seismic design of structures,
{
systems, and components will be in a new Appendix s to 10 CFR Part 50.
These relocations of' various requirements within the body of '
regulations and guidance serve two purposes: (1) oriteria for i
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The Honorable Ivan Salin 2 February 14, 1992 l 1
seismic design of a plant, now in Part 100, are moved to part 50, where they belong. And (2), many requirements in Appendix A, that i were state of the art when the appendix was written in 1973, are being brought up to date and are being removed from the regulations and placed in a regulatory guide where they can be more easily kept up to date in the future. We commend the staff for this proposed reorganization. It should make the licensing process more rational, and perhaps simpler, and will have no adverse effect on 4 l
risk.
In addition to the proposed reorganization, two of the proposed changes to the content of the regulations deserve comment. The proposed Appendix 6 to Part 50 redefines the Operating Basis i
l Earthquake (CBE) in a voy that leads to more rational consideration of the OBE in design and operation. Studies are being made to l
ensure that the proposed changes will not lead to significant increases in risk. We believe that this change is a step in the right direction. Two new Regulatory Guides, DG-1017 and DG-1018, have been proposed to provide guidance on inspections, evaluations, shutdown, and restart, following the occurrence of an earthquake greater than the OBE at a plant.
The other change is a more significant departure from current requirements. Proposed Appendix B to Part 200 requires that the SSE ground motion be determined "using both probabilistic and deterministic approaches." The staff does not claim that this new requirement will have any significant effect on safety. The staff does be1# eve, however, that a probabilistic approach will make it easier co determine an SSE ground motion in the face of unknowns or uncertainties, and that the resulting value will be more robust and resistant to challenge. In our view, that would argue for the use of a probabilistic approach, not for the use of a dual approach.
Although we are not convinced that the proposed dual approach is either necessary or desirable, we have no objection to the staff proposing and publishing it for public comment.
In summary, we have no reservations or concerns at this time that would argue against publication for comment of the several proposed revisions considered in our review. We note, however, that the documents considered were not in final form. Some had not yet been edited, others were still being modified by the staff, and none had yet been reviewed by the Committee to Review Generic Requirements.
If substantial changes are made in any of these documents before
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w The Honorable Ivan Selin 3 February 14, 1992 they are submitted to you, we expect the staff to inform us of :
them.
Sincerely, O, l David A. Ward chairman
References:
- 1. Memorandum dated October 11, 1991, from Themis P. Speis, Nuclear Regulatory Commission, for Raymond F. Fraley, Advisory Committee on Reactor Safeguards, Subjects Proposed Revision of 10 CFR Part 100, Reactor Site Criteria, Revisions to 10 CFR Part 50, New Appendix B to 10 CFR Part 100 and Appendix S to 10 CFR Part 50, and Associated Regulatory Guides (Draft Predeoisional), enclosing:
- a. Draft Revision to 10 CFR Part 50
- b. Revised Part 100, Reactor Site Criteria
- c. Proposed Revisions to Regulatory Guide 4.7,
" General Site Suitability Criteria for Nuclear Power Stations"
- 2. Memorandum dated January 21, 1992, from Lawrence C. Shao, Nuclear Regulatory Commission, for Raymond F. Fraley, Advisory Committee on Reactor Safeguards, subject: Revision of Appendix A to 10 CFR Part 100 -- Geological and Seismological Siting Criteria for Nuclear Power Plants, enclosing:
- a. Draft 10 CFR Part 100, Appendix B, Criteria for The Seismic and Geologic Siting of Nuclear Power Plants After (Effective Date]
- b. Draft 10 CFR Part 50, AppendhX S, Earthquake Engineering Criteria for Nuclear Power Plants l
- c. Draft Regulatory Guide DG-1015, Identification ~
and Characterization of Seismic Sources, Expected Maximum Earthquakes and Ground Motion i d. Draft Regulatory Guide DG-1016, Second Proposed l Revision 2 to Regulatory Guide 1.12, Nuclear Power Plant Instrumentation for Earthquakes
- e. Draft Regulatory Guide DG-1017, Pre-Earthquake Planning and Immediate Nuclear Power Plant operator Post-Earthquake Actions
- f. Draft Regulatory Guide DG-1018,' Restart of a Nuclear Power Plant Shut Down By A Seismic Event 9 Proposed Revision 3, Standard Review Plan 2.5.2, '
Vibratory Ground Motion l
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