ML20006G997
| ML20006G997 | |
| Person / Time | |
|---|---|
| Issue date: | 01/16/1996 |
| From: | Murphy A Office of Nuclear Regulatory Research |
| To: | |
| References | |
| NRC-2021-000179 NUDOCS 9601220236 | |
| Download: ML20006G997 (84) | |
Text
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t January 16, 1996-MEMORANDUM T0:
Appendix A Revision Team i
(See Attached List) l FROM:
Andrew J. Murphy, Chief Original Structural and Geological Engineering Branch Andrew J.
i Division of Engineering Technology, RES l
SUBJECT:
REVIEW 0F DRAFT SECY PAPER, FEDERAL REGISTER NOTICE, AND l
EARTHQUAKE ENGINEERING REGULATORY GUIDES ASSOCIATED WITH THE APPENDIX A TO PART 100 REPLACEMENT Attached for your review and comment are drafts of-the SECY Paper, Federal Register Notice, and three earthquake engineering regulatory guides associated with the Appendix A to Part 100 replacement.
I request that you j
provide your comments before January 22, 1996, so that they can be discussed i
along with the earth sciences related comments in a meeting during the week of January 22, 1996. This is a request for input from the Revision Team and not for a formal Division or Office level review.
Attachments:
l 1.
SECY Paper 2.
Federal Register Notice 3.
Regulatory Guide 1.12, Revision 2 (Seismic Instrumentation) 4.
Regulatory Guide 1.166 (Plant Shutdown) 5.
Regulatory Guide 1.167 (Plant Restart) 1 Distribution:
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9601220236 960116
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DOCUMENT NAME: g:\\rmk\\ review. reg Ta enceive a copy of this document, indicate in the box: *C" = Copy without attachment! enclosure
- E' = Copy with attachmentienclosure
- N* = fio copy 0FFICE SGEB/RES E
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NAME RKenneally m /
AMurphy (AJ M,.-
DATE 1/st/96
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u Appendix A Revision Team G. Bagchi R. Rothman D. Terao C. Munson N. Chokshi i
E. Zurflueh R. McMullen P. Sobel A.B. Ibrahim
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FOR:
The Commissica rs JQH:
James M. Taylor Executive Director for Operations
SUBJECT:
REVISIONS TO 10 rFR PART 100 AND 10 CFR PART 50, AND NEW APPENDIX S T0 in iTR PART 50 PURPOSE.
To obtain Commission approval to publish a final rule with revisions to the regulatory requirements for reactor siting in 10 CFR Part 100 and 10 CFR Part 50, including a new Appendix S to 10 CFR Part 50, for use by future applicants.
SUWARY:
This paper seeks Commission approval to publish a final rule package on reactor siting to amead its regulations for future light-water cooled nuclear power plant applicants to describe basic reactor site criteria and reflect advancements in the earth sciences and earthquake engineering. The revised Part 100 consists of two subparts. To preserve the licensing basis for existing plants, Subpart A ar.d Appenaix A to Part 100 would be identical to the present rule. Subpart B, applicable to future plants, would contain basic nonsensic site criteria, without numerical values, ir a new Section 100.21, "Nonseismic Siting Criteria." Seismic criteria would appear in a new Section 100.23, " Geologic and Seismic Siting Factors." Revisions to 10 CFR Part b0 would contain source term and dose criteria (Section 50.44) and earthquake engineering criteria (new Appendix 5).
BACKGROUND:
On April 12, 1962, the Atomic Energy Commission (AEC) 1:, sued 10 CFR Part 100,
" Reactor Site Criteria" (27 fB 3509). On Nrumber 13, 1973, the AEC issued Appendix A to 10 CFR Part 100, " Seismic and Geologic Siting Criteria for Nue; ear Power Plants," (38 ER 31279).
Contact:
Leonkrd Soffer, EDO 415-1722 Dr. Andrew J. Murphy, RCS 415-6010
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Part 50 was published for comment on OctoberA proposed
, and sections of proposed rule change combined two separate initiatives dealing with 20, 1992 (51 fB 47802)
The seismic and seismic issues, and included a minimum distance to t n-area boundary of 0.4 miles, guideline limits for population density c usion required both probabilistic and deterministic seismic hazard evaluat e
, and comment period, extended twice, expired on June 1, 1993.
both domestic and international, were received.
- s. The Extensive comments, rule and the nature of the comments receivedThe Commiss r
e proposed 1993, the Commission raised several concern:.
In an SRM dated August 12, aspects of t In response,he proposed revisions to Part 100 as well as its form an regarding the prescriptive the staff prepared an options paper, SECY-94-017 26, 1994 recommendations; however, due to the substantiv made to the rule the Commission stated that both parts were to s to be for Commission review and reissued for public comment prior to th rulemaking.
um ed section were to be submitted to the Commission for reviewO e final the basic site criteria are to be implemented.
an
, to de m trate how receiving Commission approval of the outlines.and standard re after The second proposed revision to these regulations was published f comment on October 17, 1994 (59 FR 52255).
or public stated (60 FR 7467) that it intended to extend the comment period interested persons adequate time to provide comments on staff guid documents.
ow On February 28, 1995, the availability of the five draftnce regulatory guides and three draft standard review plan sections that w developed to provide guidance on meeting the proposed regulation published (60 FR 10880) and the comraent period for the proposed ere s was extcoded to May 12, 1995 (60 FR 10810).
(Attachment 1), the resolution of public comments o rul.
ACRS letter en the rulemaking (Attachment 4), a dra or site criteria
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, the (Attachment 5), and the draft congressioaal letters (Attachment 6) ncement DISCUSSION:
NON-SEISMIC ASPICJ1; SEISMIC ASPECTS:
Because the criteria presented in the regulation will not be appli d t existing plants, the licensing bases for existing nuclear power plants e
o s
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i remain part of the regulations. Therefore, the criteria on seismic and geologic siting is designated as a new Section 100.23 to 10 CFR Part 100 and i
added to the existing body of regulations.
In addition, earthquake engineering criteria is located in 10 CFR Part 50, in a new Appendix S.
Since Appendix S is not self executing, applicable sections of Part 50 (s50.8 and 550.34) are revised to reference Appendix S.
Conforming amendments to 10 CFR j
Parts 52 and 100 are also made. Sections 52.17(a)(1), 52.17(a)(1)(vi), 100.8, 1
and 100.20(c)(1) and (3) are amended to note Section 100.23 to Pert 100 or Appendix S to Part 50.
Geoloaic and Seismic Sitina The regulations and guidance documents reflect new information and research results, and comments from the public.
In response to the August 12, 1993, SRM pertaining to the prescriptive aspects of the first proposed revisions to Part 100 as well as its form and content, the final regulation only contains the basic requirements, the detailed guidance which is contained in Appendix A to 10 CFR Part 100 has been removed to guidance documents. Thus, the regulation (Section 100.23 to Part 100) contains: (a) required definitions, (b) a requirement to determine the geological, seismological, and engineering characteristics of the proposed site, and (c) requirements to determine the Safe Shutdown Earthquake Ground Motion (SSE), to determine the potential for surface deformation, and to determine the design bases for seismically induced floods and water waves. Detailed guidance, that is, procedures acceptable to the NRC staff for meeting the requirements, will be contained in Regulatory Guide 1.165, " Identification and Characterization of Seismic Sources and Determination of Safe Shutdown Earthquake Ground Motion," (Draft was OG-1032).
NRC staff review guidelines will be provided in Standard Review Plan (SRP)
Section 2.5.2, " Vibratory Ground Motion," Revision 3.
Two other SRP sections, 2.5.1, " Basic Geologic and Seismic Information," and 2.5.3, " Surface Faulting," will also be revised to assure consistency among the rule, SRP Section 2.5.2, and Regulatory Guide 1.165.
The existing approach for determining a Safe Shutdown Earthquake Ground Motion (SSE) for a nuclear reactor site, embodied in Appendix A to 10 CFR Part 100, relies on a " deterministic" approach. Using this deterministic approach, an applicant develops a single set of earthquake sources, develops for each source a postulated earthquake to be used as the source of ground motion that can affect the site, locates the postulated earthquake according to prescribed rules, and then calculates ground motions at the site.
Although this approach has worked reasonably well for the past two decades, in i
the sense that SSEs for plants sited with this approach are judged to Se suttably conservative, the approach has not explicitly recognized uncertainties in geosciences parameters.
Because of the uncertainty about earthquake phenomena (especially in the eastern United States), there have often been differences of opinion and differing interpretations among experts as to the la.1est earthquakes to be considered and ground-motion models to be used, thus often making the licensing process relatively cumbersome.
Over the past decade, analysis methods for incorporating these different interpretations have been developed and used. These "probabilistic" methods
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hue been designed to allow explicit incorporation of different nodels for zoi.<ation, earthquake size, ground motion, and other parameters. The advantage i
of using these probabilistic methods is their ability to not only incorporate different models and different data sets, but also to weight them using judg-ments as to the validity of the different models and data sets, and thereby providing an explicit expression for the uncertainty in the ground motion estimates and a means of assessing sensitivity to varic,us input parameters.
Another advantage of the probabilistic method is the target exceedance i
probability is set by examining the design bases of more recently licensed nuclear power plants resulting in a more uniform level of safety from site to site.
The revision to the regulation now explicitly rece;Mzes that there are inherent uncertainties in establishing the seismic and geologic design parameters and allows for the option of using a probabilistic scismic hazard I
methodology capable of propagating uncerta*nties as a means to address these uncertainties. The rule further recognizes that the nature of uncertainty and the appropriate approach to account for it depend greatly on the tectonic i
regime and parameters, such as, the knowledge of seismic sources, the i
existence of histedcal and recorded data, and the understanding of tectonics.
i I
Therefore, methods other than the probabilistic methods, such as sensitivity analyses, may be adequate for some sites to account for uncertainties.
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The key elements of the approa;h exemplified in Regulatory Guide 1.165 and i
Standard Review Plan Section 2.5.2 are:
a.
Conduct site-specific and reoional aeoscience investiaations.
These investigations are performed to determine specific characteristics of the proposed site, such as, the presence or absence of potential seismic sources, capable faults on or near the site, characterization of the geological rock and soil st' 'a, earthquake history of site and environs, etc.
In addition to characterizing the site, these data are needed to verify that regional characteristics used in the Lawrence Livermore National Laboratory (LLNL) or the Electric Power Research Institute (EPRI) probabilistic seismic hazard assessn> nts (PSHA) are valid for the proposed site.
b.
Tarcet exceedance orobability is :et by examinina the desian bases of more recently licensed nuclear power Diants.
The target exceedance probability is the median annual probability of exceeding the Safe Shutdown Earthquake (SSE) for operating nuclear power plant that were designed to Regulatory Guide 1.60 or to a similar spectrum. This value has been determined to be IE-5/ year.
c.
Determine if information from aeoscience investiaations chance orobabilistic results.
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The applicant conducts an evaluation that demonstrates that the data obtained from the site investigations (Step a. above) do not i
provide information that would necessitate revision of the
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existing seismic sources and their characteristics or attenuation models.
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d.
Conduct probabilistic seismic hazard analysis and determine croJnd motion i
level correspondina to the taraet exceedance probability.
The applicant conducts a LLNL or EPRI PSHA for the proposed site i
to obtain a seismic hazard curve, ground acceleration vs. annual probability of exceedance. The hazard curve is deaggregated to
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determine a seismic event described by an average earthquake magnitude and distance (distance from earthquake to the nuclear 2
power plant site) which contributes most to the ground motion level corresponding to the target exceedance probability. This l
1 magnitude and distance is then used in subsequent steps to i
determine site-specific spectral shape.
e.
Determine-site-specific spectral shape and scale this shape to the around i
motion level determined above.
The applicant will use the seismic ever.t of magnitude and distance determined in Step c to develop site-specific spectral shape in accordance with SRP 2.5.2 procedures and additional guidance provided in the regulatory guide. The SRP procedures, in part, are based on use of seismic recorded motions or ground motion models appropriate for the event, region and site under consideration.
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f.
NRC staff review of around motion.
The NRC staff will review the applicants proposed SSE ground motion to assure that it takes into account all available data including insights and information gained from previous licensing experience.
g.
Update the data base and reassess probabilistic methnis at least every ten j
Years.
j To keep the regulatory guidance on the probabilistic methods and their seismic bizard data base current, the NRC would reas ess them at least ;very ten years and update them as appron sate.
Results of the regional and site-specific investigations must be considered in application of the probabilistic method. The current probabilistic methods, the NRC sponsored study conducted by LLNL or the EPRI seismic hazard study, are essentially regional studies without detailed information on any specific location. The regional and site-specific investigations provide detailed information to update the database of the hazard methodology to make the probabilistic analysis site-specific.
It is also necessary to incorporate local site geological factors such as stratigraphy and topography and to account for site-specific geotechnical properties in establishing the design basis ground motion.
'n order to incorporate local site factors and advances in ground motion attenuation models, ground motion estimates are determined using the procedures that will be outlined in Standard Review Plan Section 2.5.2.
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4 The NRC staff's review approach '.o evaluate an application is described in SRP Section 2.5.2.
This review takes into account the information base developed in licensing more than 100 plants. Although the basic premise in establishing 4
l the target exceedance probability is that the current design levels are j
adequate, a staff review further assures that there is consistency with previous licensing decisions and that the scientific basis for decisions are i
clearly understood. This review approach will also assist in assessing the fairly complex regional probabilistic modeling which incorporates multiple j
hypotheses and a multitude of parameters. Furthermore, this process should i
i provide a clear basis for the staff's decisions and facilitate communication j
with nonexperts.
Earthouake Enaineerina 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 is consistent with the location of other design requirements in 4
Part 50. The regulation is a new Appendix S, " Earthquake Engineering Criteria I
for Nuclear Power Plants," to Part 50.
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The specification in the current regulation, Appendi:: A to Part 100, that the j
Operating Basis Earthquake Ground Motion (OBE), the vibratory ground motion that will assure safe continued operation, is one-half the SSE has been j
deleted and replaced with two options: (1) applicant selection of an OBE that 1
is either one-third of the SSE or less, or (2) a value greater than one-third of the SSE. With the OBE level set at one-third or less of the SSE, only the SSE is 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, j
the OBE serves the function of an inspection and shutdown level. This change responds to one of the major criticisms with the existing regulations, that the ME controls the design of some parts of the plant.
The 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 revised accordingly. Related plant shutdown and OBE exr.edance guidelines for operatiag plants are being daveloped separate 13 by NRR.
Procedures acceptable to the NRC staff for meeting the requirements in the new regulation will be contained in three regulatory guides, (a) Regulatory Guide 1.12, " Nuclear Power Plant Instrumentation for Earthquakes," Revision 2, (Draft was DG-1033), (b) Regulatory Guide 1.166, " Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Postearthquake Actions," (Draft was DG-1034), and (c) Regulatory Guida 1.167, " Restart of a Nuclear Power Plant Shut Down by a Seismic Event," (Draft was DG-1035).
PUBLIC C0pmENTS NON-SEISMIC ASPECTS:
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SEISMIC ASPECTS:
Seven letters were received addressing either the regulations or both the regulations and the draft guidance documents. An additional five letters i
were received addressing only the guidance documents, for a total of twelve comment letters.
10 CFR 100.23 No changes were made to the regulation as a result of the public comments.
In general, the commentors were supportive of the regulation, specifically, the i
removal of prescriptive guidance from the regulation and locating it in
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l regulatory gu@s or standard review plan sections and the removal of the l
requirement from the first proposed rulemaking (57 FR 47802) that both deterministic and probabilistic evaluations must be conducted to determine j
site suitability and seismic design requirem nts for the site.
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A suggestion to codify, tte for existing sites east of approximately 105' l
west longitude (the Recky.tount.V front), a 0.39 standardized design level is acceptsble was not adopted. The GC has determined that the use of a spectral shape anchored to 0.3g neak ground acceleration as a standardized design level would be appropriate fe existing sites based on the current state of l
knowledge. However, as new information becomes available it may not be appropriate for future licensing decisions.
Pertinent information such as i
that described in Regulatory Cuide 1.165 (Draft was DG-1032) is needed to make that assessment. Therefore, it is not appropriate to codify the request.
l The suggestion to change the applicability of the regulation to enable an l
applicant for an operating license already holding a construction permit to apply the amended methodology and criteria in Subpart B to Part 100 was not incorporated. The NRC will sddress this request on a case-by-case basis rather than through a generic change to the regulations. This situation l
pertains to a limited number of facilities in various stages of construction.
Some of the issues that must be addressed by the applicant and NRC during the operating license review include differences between the design bases derived i
I from the current and amended regulations (Appendix A to Part 100 and Section i
100.23, respectively), and earthquake engineering criteria such as, OBE design requirements and OBE shutdown requirements.
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An explicit statement whether or not Section 100.23 to Part 100 applies to the Mined Geologic Disposal System (MGDS) and a Monitored Retrievable Storage (MRS) facility was not added to the regulation or Supplemental Information Section of the rule.
Presently, NUREG-1451, " Staff Technical Position on Investigations to Identify Fault Displacement Hazards and Seismic Hazards at a Geologic Repository," notes that Appendix A to 10 CFR Part 100 does not apply to a geologic repository. S?ction 72.102(b) requires that, for a MRS located west of the Rocky Mountain front or in areas of known potential seismic activity in the east, the seismicity be evaluated by the techniques of Appendix A to 10 CFR Part 100. The applicability of Section 100.23 to other than power reactors, if considered appropriate by the NRC, would be a separate rulemaking. That rulemaking would clearly state the applicability of Section 100.23 to a MRS or other facility.
In addition, NUREG-1451 will remain the l
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J NRC staff technical position on seismic siting issues pertain r4 t o e 6905 until it is superseded through a rulemaking, revision of NURIG-iO, cr Pher i
appropriate mechanism.
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Appendix S to 10 CFR Part 50 i
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Support for the NRC position pertaining to the elimination tv C;v v en Basis Earthquake Ground Motion (OBE) response analyses has u
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various NRC publications such as SECY-79-300, SECY-90-016, SEL M +$1, o l
3 NUREG-1061. The final safety evaluation reports related to the cert??% ion of the System 80+ and the Advanced Boiling Water Reactor design (NUREG-1462 and NUREG-1503, respectively) has already adopted the single earthquake design philosophy.
In addition, similar activities are being done in foreign countries, for instance, Germany. However, one negative comment expressed concern. bout the elimination of OBE response analyses of pressure-retaining components designed to the ASME Boiler and Pressure Vessel Section III rules.
Positions pertaining to the elimination of the Operating Basis Earthquake were proposed in SECY-93-087. Commission approval is documented in a memorandum from Samuel J. Chilk to James M. Taylor,
Subject:
SECY-93-087 - Policy, Technical and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Reactor (ALWR) Designs, dated July 21, 1993.
Item V(B)(5), "Value of the Operating Basis Earthquake Ground Motion (OBE) and Required OBE Analysis,"
to the supplemental information to the regulations was slightly modified to address the noted concerns.
The regulation was not changed to incorporate by reference the American Society of Civil Engineers (ASCE) Standard 4, " Seismic Analysis of Safety-Related Nuclear Structures and Commentary on Standard for Seismic Analysis of Safety-Related Nuclear Structures."
In response to the August 12, 1993, SRM pertaining to the prescriptive aspects of the first proposed revisions to Part 100 as well as its form and content, the final regulation only contains the basic requirements, the detailed guidance is provided in regulatory guides and standard review plan sectior.s. ASCE Standard 4 is cited in the 1989 revision of Standard Review Plan Sections 3.7.1, 3.7.2, and 3.7.3.
The reference to aftershocks in Paragraph IV(b), Surface Deformation was deleted. Paragraphs VI(a)(1), " Safe Shutdown Earthquake," and VI(b)(3) of Appendix A to Part 100 contain the phrase " including aftershocks."
In the proposed regulation the " including aftershocks" phrase was only removed from the Safe Shutdown Earthquake Ground Motion requirements (Paragraph IV(a)(1) of Appendix S to Part 50).
Guidance Documents Many of the commentors have 7rovided editorial and technical suggestions that would clarify the documents. A few commentors provided more substantive comments requiring a careful assessment of their implications.
For example, the Staff will clarify the procedure in SRP Section 2.5.2 used to assess the adequacy of an applicants submittal. Also, Regulatory Guide 1.165 (Draft was DG-1032) will be expanded to discuss how uncertainties in the SSE can be addressed throagh a suitable sensitivity analysis.
In general, no technical J
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changes will be made to the staff positions described in the draft guidance i
documents.
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It is anticipated that the availability of the related regulatory guidance and standard review plan sections will be published in the Federal Reaister coincident with the effective date of the final regulations.
REC 0mfMDATIONS:
That'the Commission:
l.
Anorove publication of the Revisions to the Regulatory Requirements for Reactor Siting (Seismic and Nonseismic) and Earthquake Engineering Criteria in 10 CFR Parts 100 and 50 (Attachment 1) as a final rule.
2.
Certify that this rule will not have a significant economic effect on a substantial number of small entities pursuant to the Regulatory Flexibility Act of 1980 (5 U.S.C. 605(b)).
3.
N_q11:
a.
The final rule will be published in the Federal Reaister and become effective 30 days after publication.
b.
The reporting and recordkeeping requirements contained in this regulation have been approved by the Office of Management and i
Budget, OMB approval Numbers. 3150-0093 and 3150-0011.
c.
A public announcement (Attachment 5) will be issued when the notice of rulemaking are sent to the Office of the Federal Register.
d.
The appropriate Congressional committees will be informed (Attachment 6).
f.
Copies of the Federal Reaistgt notice will be distributed to all power reactor licensees. The notices will be sent to other i
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 this rule will not have a significant economic effect on a substantial number of small entities.
h.
The availability of the final regulatory guides and standard review plan sections will be published in the Federal Reaister I
subsequent to the effective date of the final rule.
i.
Copies of " Resolution of Public Comments on the Proposed Reactor Site Criteria," (Attachment 2), and " Resolution of Public Comments on the Proposed Seismic and Earthquake Engineering Criteria for l
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10-The Commissioners Nuclear Power Plants," (Attachment 3), will be placed in the Public Document Room and sent to interested parties upon request.
Attachments:
Federal Register Notice of Rulemaking Resolution of Public Comments on the Proposed Reactor Site Criteria 1.
Resolution of Public Comments on the Proposed Seismic and Earthquake 2.
3.
Engineering Criteria for Nuclear Power Plants 4.
ACRS Letter 5.
Draft Public Announcement 6.
Draft Congressional Letters 7.
Regulatory Analysis 8.
Environmental Assessment I
s 1
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[7590-01-P]
NUCLEAR REGULATORY COMISSION
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10 CFR Parts 50, 52 and 100 i
RIN 3150-AD93 Reactor Site Criteria Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants and Proposed Dental of Petition from Free Environment, Inc. et. al.
1 AGENCY:
Nr' ear Regulatory Commission.
Final rule and denial of petition from Free Environment, Inc.
ACTION:
et.al.
l The Nuclear Regulatory Commission (NRC) is amenng its regulations to update the critiria used in decisions regarding powe SUMARY:
l The rule would allow NRC to benefit from i
future nuclear power plants.
experience gained in the application of the procedures and methods se l
in the current regulation and to incorporate the rapid advancements in earth sciences and earthquake engineering.two separate change The seismic and earthquake engineering considerations of reactor sitia.p Commission is also denying the re.
by Free Environment, Inc. et. al.
(30 days after publication in the Federal Register).
l EFFECTIVE DATE:
Dr. Andrew J. Murphy, Office of Nuclear FOR FURTHER INFORMATION CONTACT:
Regulatory Research, U.S. Nuclear Regulatory Commission, Washing concerning the seismic and earthquake engineering telephone (301) 415-6010, aspects and Mr. Leonard Soffer, Office of the Executive Director for Operations, U.S. Nuclear Regulatory Commission, Washington, DC 20555 telephone (301) 415-1722, concerning other siting aspects.
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SUPPLEMENTARY INFORMATION:
I.
Background.
II.
Objectives.
III.
. Genesis.
IV.
Alternatives.
Major Changes.
V.
Reactor Siting Criteria (Nonseismic).
A.
Seismic and Earthquake Engineering Criteria.
Related Regulatory Guides and Standard Review Plan Sections.
B.
VI.-
VII.
Future Regulatory Action.
VIII.
Referenced Documents.
Availability.
Findin[of No Significant Environmental Impact:
X.
Paperwork Reduction Art Statement.
XI.
Regulatory Analysis.
XII.
Regulatory Flexibility Certification.
XIII.
XIV.
Backfit Analysis.
1. Background
The present regulation regarding reactor site criteria (10 CFR Pa 12, 1962 (27 FR 3509). NRC staff guidance on exclusion area and low population zone sizes as well as population density was was promulgated April Regulatory Guide 4.7, " General Site Suitability Criteria for Nuclear Po Stations," published for comment in September 1974.On June 1, 1976, th was issued in November 1975.
Group (PIRG) filed a petition for_ rulemaking (P On April 28, 1977, Free population density limits into the regulations.
The Environment, Inc. et. al., filed a petition for rulemaking (PRM-50-20).
remaining issue of this petition requests that the central Iowa nuclear project and other reactors be sited at least 40 mile 9eneral policy statement on nuclear power reactor sit centers.
In the recommendations regarding siting of future nuclear p On July siting from design and to specify demographic criteria for siting.
29, 1980 (45 FR 50350), the NRC issued an Advance Notice of Propo Rulemaking (ANPRM) regarding revision of the reactor site criteria, discussed the reconniendations of the Siting Policy T December 1981 to await development of a Safety Goal and improved public comments.
4, 1986 (51 FR 23044), the NRC issued its accident source terms. On August Policy Statement on Safety Goals that stated quantitative health ob On December 14, with regard to both prompt and latent cancer fatality 2
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- unnecessarily restrict,NRC's regulatory siting policies and would not resu in a substantial increase in the overall protection of-th The Commission proposes to safety.
is proceeding with a rulemaking in this area.
address the remaining issue in PRM-50-20 as part of this rulemaking 10 CF". rt 1^^,f4J9~D was originally issued as a proposed Appendix A i:
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Nuclear Power Plants," Rii[WM5@"(N FN 22601), published as a final regulation 25,~1971 13, 1973.
regulation on November 13, 1973 (38 FR 31279), and became effective on Decem on November The first There have been two amendments to 10 CFR Part 100, Appendix The second amendment amendment, issued November regulation by adding the legend under the diagram.
resulted from a petition for rulemaking (PRM 100-1) requesting that a be issued that would interpret and clarify Appendix A with respect to the detemination of the Safe Shutdown Earthquake.
14, 1975 (40 FR 20983). The substance of thei petition was published on Maypetitioner's proposal was accepted an 10, 1977 (42 FR 2052).
finF regulation on JanuaryThe first proposed revision to these regulations was 20, 1992, (57 FR 47802).
public comment on Octoberfive draft regulatory guides and the standard re developed to provide guidance on meeting the proposed regulations wasThe c 25, 1992, (57 FR 55601).First, the NRC staff initiated published on November proposed regulations was extended two times.
17, 1993 to March 24, 1993, to be d d an extension (58 FR 271) from February consistent with the commer.t period on the draft regulatory guides and sta Second, in response to a request from the public, the review plan section.
riod was extended to June 1 1993 (58 FR 1637 I
commen at
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II. Objectives The objectives of this regulatory action are to --State basic i
experience and importance to risk, have been shown as key to 1.
ProvMe a stable regulatory basis for seismic and geologic siting health and safety; and applicable earthquake engineering design of future nuclear pow 2.
that will update and clarify regulatory requirements and provide a fle structure to permit consideration of new technical un 3.
plant design inte 10 CFR Part 50.
3
i III. Genesis The regulatory action reflects changes that are intended to (1) from the experience gained in applying the existing regulation and from research; (2) resolve interpretive questions; (3) provide needed regu flexibility to incorporate state-of-the-art improvements in the geoscien and earthquake engireering; and (4) simplify the language to a m The regulatory action would apply to applicants who apply for a English" text.
construction permit, operating license, preliminary design approva design approval, manufacturing license, early site l
r ulations.
a e site or estab shment C ifI a not assoc at li e se on of the Safe Shutdown Earthquake Ground Motion (SSE) have b This action is consistent with the location of other design CFR Part 50.
requirements in 10 CFR Part 50.Because the revised criteria prese applied to existing plants, the licensing bases for existing nuclea plants must remain part of the regulaticas. seismic reactor s t
A and Appendix A to 10 CFR Part 100, respectively.
criteria would be added as Subpart B in 10 CFR Part 100 and wou site applications received on or after the effective date of the fin 5100.21 to Non-seismic site criteria would be added as a newThe c regulations.
The dose Subpart B in 10 CFR Part 100.
would be ao'ded as a new s100.23 to Subpart B Because Appendix S is CFR Part '50 (550.34(a) and Appendix S, respectively?.
not self executing, applicable sections of Part 50 (s50.34 a flect revised to reference Appendix S. amendments to 10 CFR Part changes in 50.34(a)(1) and 10 CFR Part 100.
IV. Alternatives The first alternative considered by the Commission was to This is not current regulations for site suitability determinations. Accident source term considered an acceptable alternative.
calculations currently primarily influence plant design requirem It is desirable to state basic site criteria which, through importance to risk, have been shown to be key to assurin than siting.
Further, significant advances in understanding severe acciden behavior, including fission product release and transport, as safety.
4
l i
-.n earth-sciences and in earthquake engineering have taken place since the promulgation of the present regulation and deserve to be reflected inl regulations.The second alternative considered was replacement of ti,e existing i
+
This is not an acceptable 1i regulation with an entirely new regulation.
alternative because the provisions of the existing regulations fom part o the licensing bases for many of the operating nuclear power plants and Therefore, these l
that are in various stages of obtaining operating licenses.
ffect.
. provisions should remain in force and e 10 CFR Part 100 and relocating plant design re while retaining the existing regulation was chosen as the best alternati The public will benefit from a clearer, more uniform, and more consi licencing process that incorporates updated infomation and implementation (both technical and legal), fewer interpretiv fewer %terpretations.
-?
increased regulatory flexibility. addition to avoiding licensing delay requirements.
V. MAJOR CHANGES Reactor Siting Criteria (Nonselsnic).
A.
Seismic and Earthquake Engineering Criteria.
B.
" Seismic and Geologic Siting The following major changes to Appendi.x A.. fPart 100, are associated with Criteria for Nuclear Power Plants," to
'er 'a rule making. These changes the seismic and earthquake engineering cr reflect new information and research results, and incorporate the int of this regulatory action as defined in Section III of this rul; inl:'i ;
c:;;1:ti=:. A fr;; th: ;dli: :: th: fir:t ;=;=:d r=i:i= cf th:
- p::tfi: d = r::t ::;1:inin; th: "E :t:ff': di:; = iti:
f :rtin::t : n nt:
- u::t:
f'=1 =l:riir will " --
" :inid=t z'th th:
3.
Separate Sitina from % sion.
Criteria not associated with site suitability or establishment of th Safe Shutdown Earthquake Ground Motion (SSE) have been p This action is consistent with the location of other design requi Because the revised criteria presented in the regulation 50.
will not be applied to existing plants, the licensing basis for existing in 10 CFR Part 50.
The criteria on nuclear power plants must remain part of the regulations.
seismic and geologic siting would be designated as a new Subpart B in 10 CFR Part 100.
5
designated as a new Appendix S, " Earthquake Engineering Criteria for Nuclear Power Plants.. to 10 CFR Part 50.
2.
"-ave Detailed Gui@=ce from the Reaulation.
4en-contains both requiremen s a gu' ance on ow o satisfy the requirements. For example,Section IV, " Required Investigations," of Appendix A, states that investigations are required for vibratory ground motion, surface faulting, and seismically induced floods and water waves. Appendix A then provides detailed guidance on what constitutes an acceptable investigation. A similar situation j
exists in Section V, " Seismic and Geologic Design Bases," of Appendix A.
Geoscience assessments require considerable latitude in judgment. This latitude in judgment is needed because of limitations in data and the state-of-the-art of geologic and seismic analyses and because of the rapid evolution taking place in the geosciences in terms of accumulating knowledge and in l
modifying concepts. This need appears to have been recognized when the existing regulation was developed. The existing regulation states that it is c
based on limited geophysical and geological information and will be revised as necessary when more. complete information becomes available.
However, having geoscience assessments detailed and cast in a regulation has created difficulty for applicants and the staff in terms of inhibiting the use of needed latitude in judgment. Also, it has inhibited flexibility in applying basic principles to new situations and the use of evolving methods of analyses (for instance robabilistic) in the licensing process.
0CFRPart100rathe$ streamlined,becominganew egulation ;xid k The r Ihan a new a>
ndix to Part section in Subpart to 100. Also the level of detail presented in the
- regulation reduced considerably. Th u ;; n =h ref1 =t: /.: phi 1=:;hy :f th: fint en;=:d
=hi= tht th n;;hti= =1; = h'= th: h:i:
n;;irmu rd tht th d:hthd ;;id:=:, d.i:t h :=ui=d in th =cr=t n;;uti=, f
.n;;du ;;idrx drnau.
Thus, the pW;;xdh A u 10 CFR. ;rt I^^, 5: a.) required definitions, b.) a Eregulationcontains:
l requirement to deters' ne the geological, seismological, and engineering l
characteristicsoftheproposedsite,andc.)e-requirement [inty,todetermine to determine the l
Safe Shutdown Earthquake Ground Motion (SSE) =d it := = u the potential for surface deformation, and to determine the design bases for seismically induced floods and water waves. The guidance documents describe how to carry out these required determinations. The key elements of the be4 =::d approach to determine the SSE are presented in the following section.
The elements are the guidance that will 5: fully %= described in th: ;;id=::
dx= :3. Th: pn;n:d n;;hti= h : := =:1.
in P=t 1^^ nth = thz
= 1;; xdin t:. =t !^^. Th: pn;n:d n;;hti= cxid id=tify xd =ublSh h:i: n;;i; rau.
0:hihd ;;id==, thi h, th: ;=::d =: =;;;uth i:
th ""O f= =: ting th: n;;irsau, ;xid k :=ut=d in : d=ft n;;hury id: u k S:::d f= ; blic rat = S :ft Regulatory Guide, OC 1022,
" Identification and Characterization of Seismic Sources and termination of Safe Shutdown Earthquake Ground Motions."
3.
Uncertainties and Probabilistic Methods 6
j i
1
i a
i 3
4 j
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 j
' 100, relies on'a " deterministic" approach. Using this deterministic approach, l
an applicant develops a single set of earthquake sources, develops for each i
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 l
rules, and then calculates ground motions at the site.
Although this approach has worked reasonably well for the past two l
i decades, in the sense that SSEs for plants sited with this approach are judged to be suitably conservative, the approach has not explicitly recognized uncertainties.in.geosciences parameters. Because so little is known about earthquake phenomena (especially in the eastern United States), there have often been differences of opinion and differing interpretations among experts as to the largest earthquakes to be considered and ground-motion models to be
+
1 used, thus often making the licensing process relatively unstable.
Over the past decade, analysis methods for incorporating these different l
interpretations have been developed and used. These "probabilistic" methods j
have been designed to allow explicit incorporation of different models for j
zonation, earthquake size, ground motion, and other parameters. The advantage of using these probabilistic methods is their ability to not only incorporate different models and different data sets, but also to weight them using judg-ments as to the validity of the different models and data sets, and thereby providing an explicit expression for the uncertainty in the ground motion estimates and a means of assessing sensitivity to various input parameters.
Another advantage of the probabilistic method is the target exceedance probability is set by examining the design bases of more recently licensed nuclear power plants.
that there are inherent uncertainti'e~~~W regulation now-explicitly recognizes The pn; =:d ra h h t: th: M.s in establishing the seismic and geologic design parameters and allows for the option of using a probabilistic seismic hazard methodology capable of propagating uncertainties as a means to address these uncertainties. The rule further recognizes that the nature of uncertainty and the appropriate approach to account for it depend greatly on che tectonic regime and parameters, such as, the knowledge of seismic sources, the existence of historical and recorded data, and the understanding of J
tectonics. Therefore, methods other than the probabilistic methods, such as i
sensitivity analyses, may be adequate for some sites to account for uncertainties.
Th ~ :t:ff h= =hin:d = 1;;=Fht: hh::: 5:t== it=:hhtic
=d Fidiluth ni=h h=:-d =:1= tin; t: 5: =:d 5 th: nyhS3 f th:
>+"*R p,y,g m,p,w;y g***g*yy7"y:.x 33 A;n 3-y p3
. rpp - ypy
- w..
,+v,p?. y ygyn.;, p fg; yj s % gy + he key s% g &s m;).LJ__z N
<r, gj., 4. ; e my &
.Qpwg._
4 1-
+ w m
.._~s-approac are:
e enen s o Conduct site-specific and regional geoscience investigations, Target exceedance probability is set by examining the design bases of more recently licensed nuclear power plants, Conduct probabilistic seismic hazard analysis and determine ground motion level corresponding to the target exceedance probability 7
_ _.. _._._. m _,
4 Determine if information from Nifi[c resu"M geoscience
'i investigations change probabills ts, 4
Determine site-specific spectral shape and scale this shape to the ground motion level determined above.
NRC staff review using all available data including insights and information from previous licensing experience, and Update the data base and reassess probabilistic methods at least every ten years.
Thus, the pn;=:d approach requires thorough regional and site-specific geoscience investigations. The p n; =:d ;;r=:5 n fh :t: ::= :f th:
- t: Of th: ".S. utility ind=try. The ".S. 2 :1 ;i =1 ! r =; p =;id d :
=rin :f ::
- .t
- =d n;:-- ::d:ti: : th:t 1:d i: =d := 5: =0 by the d=:
ht:; = t:d :;; n = h.
1 1
Results of the regional and site-specific investig tions must be a
considered in application of the probabilistic method. The current probabilistic methods, the NRC sponsored study conducted by Lawrence Livermore National Laboratory (LLNL) or the Electric Poter Research Institute (EPRI) seismic hazard study, are essentially regional studies without detailed information on any specific location. The regional and site-specific investigations provide detailed information to update the database of the hazard methodology to make the probabilistic analysis site-specific.
It is also necessary to incorporate local site geological factors such as stratigraphy and topography and to account for site-specific geotechnical properties in establishing the design basis ground motion.
In order to incorporate local site factors and advances in ground motion attenuation models, ground motion - '-"
~3@d Revi]ew Plan Section 2 5 2 procedures outlined in th: Onft'$fancar
.., Geeend Pn;=:d 5;bi= 3, " Vibratory Ground Motion,"_ :=:
tth:d; =::;t:,bh t: th: "O :t:ff f= i
..:; the pn ; =:d g;hti= =: d==it:d h Onft 5;;ht=y Outd: 00 1932, "!d=tifiuti=
=d Chr =t=inti= Of ki= k S = =: =d 0:t: mi=ti= cf S.f: Shtd=:
Srth;=h C==d Sti=:."
The NRC staff's review a>
information base developed [@n~E@p, roach to evaluate a in 9Wi-SRP Section 2.5.2 ci@niGg more than 100 plants.$. This review I
Thh :t:ff
=vi= b = = ht =t with th: ht=t :f : "SCS r=:x=d:ti=.
"lthough the basic premise in establishing the target exceedance probability is that the current design levels are adequate, a staff review further assures that there is consistency with previous licensing decisions and that the scientific basis for decisions are clearly understood. This review approach will also assist in assessing the fairly complex regional probabilistic modeling which incorporates multiple hypotheses and a multitude of parameters.
Furthermore, this process should provide a clear. basis for the staff's decisions and facilitate communication with nonexperts.
4.
Safe Shutdown Earthouake.
The existing regulation (10 CFR Part 100, Appendix A, Section V(a)(1)(iv)) states "The maximum vibratory accelerations of the Safe Shutdown Earthuake at each of the various foundation locations of the nuclear power plant structures at a given site shall be determined..." The location of the 8
pj seismic input motion control point as stated in the existing regulation has led to confrontations with many ap'p'11 cants that believe this stipulation is inconsistent with good engineering fundamentals.
Thepropeoed-j ulation c::1d =sve the location of the seisraic input motion contro @ point from the foundation [-level to the free-field 3
1' free ground surface.
The 1975 version of the Standard Review Plan placed the control motion in the free-field. The peepened-RMregulation is also consistent with the resolution of Unresolved Safeuy Issue (USI) A-40, " Seismic Design Criteria" (August 1989), that resulted in the revision of Standard Review Plan Sections 2.5.2, 3.7.1, 3.7.2, and 3.7.3.
However, the F:;:::d regulation requires that the horizontal component of the Safe Shutdown ake Ground Motion in the free-field at the foundation level of the ar structures must be an appropriate response spectrum considering the site 9ectechnical properties, with a peak ground acceleration of at least 0.1g.
5.
Value of the Operatino Basis Earthouake Ground Motion (OBE) and
~
Reauired OBE Analy1gi.
The existing regulation (10 CFRI$N MiBD states that the maximum vibratory ground motion, Appendix A,Section V(a)(2))
of the OBE is at least one half the maximum vibratory ground motion of the Safe Shutdown Section VI(a)(2)) states that the engineering method [gg M ground motion. Also, the existir.g regulation (10 CFR Appendix A, usedIo'n,surethat 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 mu'.ti-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 ested that or a s ng even nspec; on ar ion for earthquakes in excess of some spec ed limit ev W
HQlaaybethemostsoundregulatoryapproach.
f,.
% por. o e J.5.' Nuclear Regulatory Commission Piping Review Committee,"
Vol.5, April 1985, (Table 10.1) ranked a decoupling of the OBE and SSE as third out of six high priority changes.
In SECY-90-016, " Evolutionary Light Water Reactor (LWR) Certification Issues and Their Relationship to Current Regulatory Requirements," the NRC staff states that it a rees that the OBE should not control the desi n of safet stems.
3 es equ to coup ng are a so 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 F:;:::d RFregulation ;::1d allowsf the value of the OBE to be set at (i) one-thirif or iess of the SSE, where OBE requirements are satisfied without an explicit response or design analyses being performed, or (ii) a 9
l
.- - -~ -- --
- ~ ~ ~ " ~ ~ ~ ~
l 4
g i'
value greater than one-third of the SSE, where analysis and design are required. There are two issues the'appli~ cant 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 ij Shutdown), and second, the amount of analyses associated with the OBE.
i applicant may determine that at one-third of the SSE level, the probability of An j
exceeding the OBE vibratory ground motion is too high, and the cost associated with plant shutdown for inspections and testin prior to restarting the plant is unacceptable.g of equipment and structures voluntarily select an OBE value at some hicher fraction of the SSE to avoid j
plant shutdowns.
However, if an applicant selects an OBE value at a fraction 1
of the SSE higher than one-third, a suitable analysis shall be performed to demoastrate that the requirements associated with the OBE are satisfied.
design shall take into account soil-structure interaction effects and the The expected duration of the vibratory ground motion.
The requirement associated with the OBE is that all structures, systems, and components of the nuclear j
power plant necessary for continued operation without undue risk to the health j
and safety of the public shall remain functional and within applicable stress, strain and deformation limits when subjected to the effects of the DBE in i
combination with normal operating loads.
As stated above, it is determined that if an OBE of one-third of the SSE is used, the requirements of the OBE can be satisfied without the applicant performing any explicit response analyses.
function of an inspection and shutdown earthquake.In this case, the OBE serves the Some minimal design checks i
and the applicability of this position to seismic base isolation of buildings are discussed below.
There is high confidence that, at this ground-motion level with other postulated concurrent loads, most critical structures, i
systems, and components will not exceed currently ~used design limits.
I This is ensured, in part, because PRA insights will be used to support a margins-type assessment of seismic events. ~ A PRA-based seismic margins analysis will i
consider sequence-level High Confidence, Low Probability of Failures (HCLPFs) and fragilities for all sequences leading to core damage or containment i
failures up to approximately one and te thirds the ground motion acceleration l
of the design basis SSE (
Reference:
Itt 1,N, Site-Specific Probabilistic Risk Assessment and Analysis of External tvents, memorandum from Samuel J.
Chilk to James M. Taylor,
Subject:
SECY-93-087 - Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advance Light-Water Reactor j
(ALWR) Designs, dated July 21, 1993.
There are situations associated with current analyses where only OBE is associated with the design requirements, for example, the ultimate heat sink (see Regulatory Guide 1.27, " Ultimate Heat Sink for Nuclear Power Plants").
In these situations, a value expressed as a fraction of $ $
the SSE response would be used in the analyses.
existing guides that would be revised technically to malintain the design philosophy.
In SECY-93-087, "P:licy, Technical, and Licensing Issues Pertaining to Evolutionary and Advance Light-Water Reactor (ALWR) Designs," the NRC staff requested Commission approval on 42 technical and policy issues pertaining to either evolution:,:y LWRs, passive LWRs, or both.
The issue pertaining to the elimination of the OBE is designated I.M.
The NRC staff identified actions necessary for the design of structures, systems, and components when the OBE design requirement is eliminated.
The staff clarified that guidelines should 10
o 7 ;
k k
be maintained to ensure the functionality of components, equipment, and.their supports. -In addition, the staff clarified how certain design requirements
'are to be considered for buildings and structures that are currently oesigned for the CBE, but not the SSE. Also, the NRC staff has evaluated the effect on
^
safety of eliminating the OBE from the design lod combinations for selected structures, systems, and components and has developed proposed criteria for an analysis usino only.the SSE.
Commission approval is documented in the Chilk to Taylor memorandum dated July 21, 1993, cited above.
More than one earthquake response analysis for a seismic base isolated 4
nuclear power plant design may be necessary to ensure adequate performance at all earthquake levels. Decisions pertaining to the response analyses associated with base isolated facilities w111 be handled on a case by case basis.
6.
Raauired P1 ant Shut %.
motion exceeding that of the OBE occurs, s)hutdown of the nucle 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 550.36(c)(2) of 10 CFR Part 50 to assure that each power plant is aware of the limiting condition of operation which is imposed under Section V(2) of Appendix A to 10 CFR Part 100.
This limitation requires that if vibratory ground motion exceeding that of the OBE occurs, shutdown of the nuclear power plant will be required.
Prior to resuming operations, the licensee will be required to demonstrate to the Commission that no functional damage has occurred to those features necessary for continued operation without undue risk to the health and safety of the public." At that time, it was the intention of the Commission to treat the Operating Basis Earthquake as a limiting condition of operation.
From the statement in the Supplementary Information, the Commission directed applicants to specifically review 10 CFR Part 100 to be aware of this intention in complying with the requirements of 10 CFR 50.36. Thus, the requirement to shut down if an OBE occurs was expected to be implemented by being includ.d nong the technical sr:cifications submitted by applicants after the adoption of Appendix A.
In fact, applicants did not include OBE shutdown requirements in their technical specifications.
The W _
regulation ;::!d treat plant shutdown associated with vibratory ground motion. exceeding the DBE {or significant plant d a condition in every operating license. A new 550.54(ff) ;::ld k ii added to the regulations to require a process leading to plant shutdown for bcensees of nuclear power plants that comply with the earthquake engineering criteria in Paragraph IV(a)(3) of N:;:::d Appendix S. " Earthquake Engineering Criteria for Nuclear Power Plants," to 10 CFR Part 50.
i Immediate shutdown could be required until it is determined that structures, systems, and components needed for safe shutdown are still functional.
DeeMegulatory Guide DG6, " Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Post-Earthquake Actions," h MS; i;1;:d t: provide [ guida'nce acceptable to the NRC staff for determining j
whether or not vibratory ground motion exceeding the OBE ground motion or i
significant plant damage had occurred and the timing of nuclear power plant shutdown.
The guidance is based on criteria developed by the Electric Power 11
- - ~ '
~
^-~
^ ^ ~
^
Research' Institute (EPRI). The decision to shut down the plant should be made
. jthin eight hours after the earthquake. _The data.from the seismic winstrumentation, coupled with information'obtained from a plant walk dow.,' are used to make the determination of when the plant should be shut down, if. it has not already been shut down by operational perturbations resulting from the seismic event. The guidance big i.;1;:f in Deem-Regulatory Guide DG-44M M is based on two assumptions, first, that the nuclear power plant has operable seismic instrumentation, including the equipment and software required to process the data within four hours after an earthquake, and second, that the operator walk' down inspections can be performed in'approxi-mately four to eight hours depending on the number of personnel conducting the inspection. The regulation also includes a provision that requires the licensee to consult with the' Commission and to propose a plan for the timely, safe shtdown of the nuclear power plant if systems, structures, or components necessary for a safe shutdown or to maintain a safe shutdown are not available.
(This unavailability may be due to earthquake related damage.)
Dren-Regulatory Guide
" Restart of a Nuclear Power Plant Shut Down by a Seismic Event," i: hig
- m p
- d t: provide [d tests ofguidelines that are acceptable to the NRC staff for performing inspections an nuclear power pla'it equipment and structures prior to plant restart. This guidance is also based on EPRI reports.
Prior to resuming operations, the licensee must demonstrate to the Commission that no functional damage has occurred to those features necessary for continued operation without undue risk to the health and safety of the public. The results of post-shutdown inspections, operability checks, and surveillance tests must be documented in written reports and submitted to the Director, Office of Nuclear Reactor Regulation. The licensee shall not resume operation until authorized to do so by the Director, Office of Nuclear Reactor Regulation.
7.
Clarify interoretations.
In Section 100.23 to 10 CFR Part 100, changes have been made to resolve
,definitionsandreq[uired questions of interpretation. As an ex regulation ;;;1d h investigations stated in the significantly changed to eliminate or ify phrases that were more applicable to only the western part of the Unitad States.
The institutional definition for " safety-related structures, systems, and components
- is drawn from Appendix A to Part 100 under III(c) and VI(a).
With the pr:;:::d relocation of the earthquake engineering criteria to Appendix S to Part 50 and the pr:;:::d rel,ocation and modification to dose guidelines in 550.34(a)(1), the definition of safety-related structures, systems, and components is included in Part 50 definitions with reference to both the Part 100 and Part 50 dose guidelines.
VI. Related Regulatory Guides and Standard Review Plan Section The NRC is developing the following dr:ft regulatory guides and standard review plan sections to provide prospective licensees with the necessary guidanca for implementing the pr;;:::d M regulation. The notice of 12
later issue of the availability for these naterials will be published in a
. " Identification and Characterization Endsral Raaister.
E E.
ers.nat on of Shutdown Earthquake Ground Motions."
describes 1.
The 2:ft guide provides general guidance and recommendations of Seismic Sources a that present acceptable procedures and provides a list of references ble tectonic sources acceptable methodologies to identify and characterize capaSectio and seismogenic sources.
Plant
- 2. = M??, Third 7.;;;;;d Regulatory Guide 1.12, " Nuclear P key elements.
The &;ft suide describes Instrumentation for Earthquakes," Revision 2.
h teristics, seismic instrumentation type and location, operability, c arac and maintenance that are acceptable to the NRC
" Pre-Earthquake Planning and installation, actuation std arthquake Actions."C staff for a timely The deem 3.
ra or guide provides guidelines that are acceptable to r
Immediate Nuclear d to determine uired. " Restart of a Nuclear Power Plant whether or not plant shutdown is' r
.. guide provides guidelines that are d tests of nuclear 4.
acceptable to the NRC staff for performing inspections anf a plant th ven.
Shut Down by a Sei power plant equipment and structures prior to restart o" :::f Rev shut down because of a seismic event.5. k:ft -Stan Kdescribes se se c information Geologic and Seismic Infomation."
the suitability of c an procedures to. assess the adequacy of the geolcite i
R:::d Tc:;:::d Revision 3 the plant site.6. Deem-Standard Review Plan Section.2.S.2he-dee J
t the site and to
" Vibratory Ground Motion." assess the ground motion potentia suic sources a 3,
assess the adequacy of the SSE.7. R;ft-Standard Review Pl
~
Wl.$
re aled to the existenco of a
" Surface Faulting."
the adequacy of the applicant's sutinatla potential for surface faulting affecting the site.
ite
- 8. Z T?, C;; f Tc:;:::d Regulatory Guide 4.7, " General S This guide Suitability Criteria for Nuclear Power Plants," Revision 2.
lth and safety d'iscusses the major site characteristics related to public hea in detemining the and environmental issues that the NRC staff considers 4
suitability of sites.
VII. Future Regulatory Action t
Several existing regulatory guides will be revised to inco desi or anal sis hilosophy.
subsequent editorial changes or maintain the existi
~'
These guides will be issued ~
regu a lons to the publication of the fina
- tir.
- x.-
13
The following regulatory guides will be revised to incorporate editorial changes, for example to reference new sections to Part 100 or Appendix S to Part 50. No technical changes will be made'in these regulatory guides.
l 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 Tube:;."
5.
1.92, " Combining Modal Responses and Spatial Components in Seismic Response Analysis."
6.
1.102, " Flood Protection for Nuclear Power Plants."
7.
1.121, " bases for Plugging Degraded PWR Steam Generator Tubes."
8.
1.122, " Development of Floor Design Response Spectra for Seismic Design of Floop-Supported Equipment or Components."
The following regulatory guides will be revised to update the design or analysis philosophy, for example, to change DBE to a fraction of the SSE:
1.
1.27, " Ultimate Heat Sink for Nuclear Power Plants."
2.
1.100, " Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power Plants."
3.
1.124, " Service Limits and Loading Combinations for Class 1 Linear-Type Component Supports."
4.
1.130, " Service Limits and Loading Combinations for Class 1 Plate-and-Shell-Type Component Supports."
5.
1.132, " Site Investigations for Foundations of Nuclear Power Plants."
6.
1.138, " Laboratory Investigations of Soils for Engineering Analysis and Design of Nuclear Power Plants."
7.
1.142, " Safety-Related Concrete Structures for Nuclear Power Plants (Other than Reactor Vessels and Containments)."
8.
1.143, " Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Waten-Cooled Nuclear Power Plants."
Minor and conforming changes to other Regulatory Guides and standard review plan sections as a result of F:;:::d changes in the nonseismic criteria are also planned.
If substantive changes.re made during the revisions, the applicable guides will be issued for public comment as draft guides.
VIII. Referenced Documents An interested person may examine or obtain copies of the documents referenced in this F:;;;:d rule as set out below.
14
SuperinIi6dist(fli2}'
7dWRE es of NUREG-0625, M{be purchased from the{C3451
$1]NUREG-ll50,JiUR of C
Co U.S. Government Printing Office, Mail S
'~
and NUREG/CR-2239 may Riiiids 22161. A copy is also Service, 5285 Port Royal Road, Springfield, VA 20402-9328.
RC Public Document available for inspection and copying for a fee in the N Room, 2120 L Street, NW. (Lower Level), Washin nt Printing Office (GPO) at the current GP0 price.d t of Documents, U.S.
prices may be obtained by contacting the Superinten en 2
20402-9328.
Government Printing Office, Mail Stop SSOP, W l Information Service on a standing order basis.
22161.
writing NTIS, 5826 Port Royal Road, Springfield, VA l ble for inspection and copying for a fee at the _.....
E* PubSECY 79-300, SECY 90-016, SECY 93-087, and W lic Documer t Room,
~^
2120 L Street, NW. (Lower Level), Washington, DC.
IX. Summary of Comments on the Proposed Regulations.
Reactor Siting Criteria (Honseismic).
A.
Seismic and Earthquake Engineering Criteria.
both Seven letters were received addressing either the regulati B.
d in Section VI the regulations and the draft guidance documents identifie An additional five letters were received addressing A document, t
the guidance documents, for a total of twelve comment let ers.
(except DG-4003).
thquake
" Resolution of-Public Comments on the P l i ing the Copies of this NRC's disposition of the comments received L Street NW.
l Regulatory (Lower Level), Washington, DC.
Murphy, Office of Nuclear Regulatory Research, U.S. Nuc ear A second Commission, Washington, DC 20555, telephone (301) 415-601 ides and document, " Resolution of Public Coments i
nd ll explain the Earthquake Engineering Criteria for Nuclear Power Plants,"
documents. The NRC's disposition of the comments received o idance documents will also discuss how to obtain copies of the commen follows.
Section 100.23 to 10 CFR Part 100 15
i I
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!-d l
The Nuclear Energy Institute (NEI) congratulates the NRC staff'for i
carefully considering and responding to the voluminous and complex comments that were provided on the earlier proposed rulemaking package (57 FR 47802
{
and considers that the seismic portion of the proposed rulemaking package )j j
nearing maturity and with the inclusion of industry's comments (which werei principally on the guidance documents), has the potential to satisfy the i
objectives of predictable licensing and stable regulations, i
j Both NEI and Westinghouse Electric Corporation support the regulation j
format, that is, prescriptive guidance is located in regulatory guides or j
standard review plan sections not the regulation.
i k
{
NEI and Westinghouse Electric Corporation support the removal of the j
requirement from the.first proposed rulemaking (57 FR 47802) that both j
deterministic and probabilistic evaluations must be conducted to determine site suitability and seismic design requirements for the site.
[ Note: the commentors do not agree with the NRC staff's deterministic check of the hazard analyses (Regulatory Guide 1.165, draft was DG-1 i
not support the NRC staff's deterministic check of the applicants submittalAlto, the (SRP Section 2.5.2).
These items are addressed in the document pertaining to comment resolution of the draft regulatory ouldes and standard review plan sections.]
The Department of Energy (Office of Civilian Radioactive Waste Management), requests an explicit statement whether or not Section 100.23 to Part 100 applies to the Mined Geologic Disposal System (NGDS) and a Monitored Retrievable Storage (MRS) facility.
The NRC has noted in NUREG-1451, " Staff Technical Position on Investigations to Identify Fault Displacement Hazards and Seismic Hazards at a Geologic Respository," that Appendix A to 10 CFR Part 100 does not apply to a geologic repository. NUREG-1451 also notes that the contemplated revisions to Part 100 would also not be applicable to a geologic repository. Section 72.102(b) requires that, for a MRS located west of the Rocky Mountain front or in areas of known poter.tial sMsmic activity in the east, the seismicity be evaluated by the techniques of Appendix A to 10 CFR Part 100.
Although Appendix A to 10 CFR Part 100 is titled " Seismic and Geologic Siting Criteria for Nuclear Power Plants," it is also referenced in two other parts of the regulation.
They are (1) Part 40, " Domestic Licensing of Source Material " Appendix A " Criteria Relating to the Operation of Uranium Mills and the Disposition of Tailings or Waste Produced by the Extraction or Concentration of Source Material from Ores Processed Primarily for Their Source Material Content,"Section I, Criterton 4(e), and (2) Part 72,
" Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste," Paragraphs (a)(2), (b) and (f)(1) of $72.102.
The referenced applicability of Section 100.23 to other than power reactors, if considered appropriate by the NRC, would be a separate rulemaking. That rulemaking would clearly state the applicability of Section 100.23 to a MRS or other facility.
In addition, NUREG-1451 will remain the NRC staff technical position on seismic siting issues pertaining to a MGDS until it is superceded through a rulemaking, revision of NUREG-145I, or other appropriate mechanism.
16.
_ q
\\
NEI, Westinghouse Electric Corporation,-and Yankee Atomic-Electric Corporation recommend that the regslation should state that for' existing sites east of the Rocky Mountain Front (east of approximately 105* west longitude),
a 0.3g standardized design level is acceptable at these sites given A
confirmatory foundations evaluations [ Regulatory Guide 1.132, but not the geologic, geophysical, seismological investigations in Regulatory Guide 1.165].
The NRC has determined that the use of a spectral shape anchored to 0.3g l
j peak ground acceleration as a standardized design level would be appropriate i
l for existing sites based on the current state of knowledge.
However, as new information becomes available it may not be appropriate for future licensing decisions.
Pertinent information such as that described in Regulatory Guide
.l.165 (Draft was DG-1032) is needed to make that assessment. Therefore, it is not appropriate to codify the request.
4 NEI recommended a rewording of. Paragraph (a), Applicability. Although unlikely, an applicant for an operating license already holding a construction permit may elect to apply the amended methodology and criteria in Subpart B to Part 100.
The NRC will address this request on a case-by-case basis rather than through a generic change to the regulations. This situation pertains to a limited number of facilities in.various stages of construction.
Some of the issues that must be addressed by the applicant and NRC during the operating license review include differences between the design bases derived from the current and amended regulations (Appendix A to Part 100 and Section 100.23, respectvely), and earthquake engineering criteria such as, OBE design requirements and OBE shutdown requirements.
Appendix S to 10 CFR Part 50 Support for the NRC position pertaining to the elimination of the Operating Basis Earthquake Ground Motion (OBE).esponse analyses has been documented in various NRC publications such as SECY-79-300, SECY-90-016, SECY-93-087, and NUREG-1061. The final safety evaluation reports related to the cer'.ification of the System 80+ and the Advanced Boiling Water Reactor design (NOREG-1462 and NUREG-1503, respectively) has already adopted the. single earthquake design philosophy.
in foreign countries, for instance, Germany.In addition, similar activities are being d (Additional discussion is provided in Section V(B)(5) of this rule).
l One commentor, ABB Combustion Engineering Nuclear Systems, specifically stated that they agree with the NRC's proposal to not require explicit design analysis of the Operating Basis Earthquake Ground Motion (OBE) if its peak acceleration is less than one-third of the Safe Shutdown Earthquake Ground Motion (SSE).
The only negative comments, from G.C. Slagis Associates, stated that the proposed rule in the area of required OBE analysis is not sound, not technically justified, and not appropriate for the design of pressure-retainning components. The following are specific comments limited to the design of pressure-retaining components to the ASME Boiler an(d Pressure Vessel Section III rules) that pertain to the supplemental information to the proposed regulations, item V(B)(5), "Value of the Operating Basis Earthquake Ground Motion (OBE) and Required OBE Analysis."
i 17 i
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.C O
1 1
(1) Disagrees with the statement in SECY-79-300 that design for a single limiting event and inspection and evaluation for earthquakes in excess of some specified limit may be the most sound rehdiatory approach.
It is not feasible to inspect for cyclic damage to all the pressure-retaining components.
Visually inspecting for permanent deformation, or leakage, or failed component supports is certainly not adequate to determine cyclic damage.
The NRC agrees. Postearthquake inspection and evaluation guidance is described in Regulatory Guide 1.167 (Draft was DG-1035), " Restart of a Nuclear Power Plant Shut Down by an Seismic Event." The guidance is not limited to visual inspections, it includes inspections, tests, and analyses including -
fatigue analysis.
(2) Disagrees with the NRC statement in SECY--090-016 that the OBE should not control design. There is a problen with the present requiremerts.
Requiring design for five OBE events at 4 SSE is unrealistic for most (all?)
sites and requires an excessive and unnecessary number of seismic supports.
The solution is to. properly define the OBE magnitude and the number of events expected during the life of the plant. And'to require design for that loading. OBE may or uy not control the design. But you cannot assume, before you have the seismicity defined and before you have a component design, that OBE will not govetn the design.
The NRC has concluded that design requirements based on an estimated OBE magnitude at the plant site and the number of events expected during the plant life will lead to low design values that will not control the design thus resulting in unnecessary analyses.
(3) It is not technically justified to assume that Section III components will remain within applicable stress limits (Level B limits) at one-third the SSE. The Section III acceptance criteria for Level D (for an SSE) is completely different than that for Level B (for an OBE). The Level D criteria is based on surviving the extremely-low probability SSE load. Gross structural deformations are possible, and it is expected that the component will have to be replaced. Cyclic effects are not considered. The cyclic effects of the repeated earthquakes have to be considered in the design of the component to ensure pressure boundary integrity throughout the life, especially if the SSE can occur after the lower level earthquakes.
In SECY-93-087, Issue I.N, " Elimination of Operating-Basis Earthquake,"
the NRC recognizes that a designer of piping systems considers the effects of primary and secondary stresses and evaluates fatigue caused by repeated cycles of loading.
Primary stresses are induced by the inertial effects of vibratory i
l motion. The relative motion of anchor points induces secondary stresses. The repeating seismic stress cycles induce cyclic effects (fatigue). However, after reviewing these aspects, the NRC concludes that, foi primary stresses, a
if the OBE is established at one-third the SSE, the SSE load combinations control the piping design when the earthquake contribution dominate:; the load combination. Therefore, the EC concludes that eliminating the OBE piping stress. load combination for primary stresses in piping systems will not significantly reduce existing safety margins.
Eliminating the OBE will, however, directly affect the current methods used to evaluate the adequacy of cyclic and secondary stress effects in the piping design.
Eliminating the OBE from the load combination could cause uncertainty in c.aluating the cyclic (fatigue) effects of earthquake-induced motions in piping systems and the relative motion effects of piping anchored to equipment and structures at various elevations because both of these 18
effects are currently evaluated only for OBE loadings. Accordingly, to
. acco' int for earthquake cycles in the fatigde analysis of piping systems, the staff proposes to develop guideli.ies for selecting a number of SSE cycles at a fraction of the peak amplitude of the SSE. These guidelines will provide a level of fatigue design for the piping equivalent to that currently provided in Standard Review Plan Section 3.9.2.
Pos',tions pertaining to the eliminulon of the Operating Basis Earthquake were proposed in SECY-93-087. Commission approval is documented in a memorandum from Semuel J. Chilk to Jases M. Taylor,
Subject:
Policy, Technical and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Reactor (ALWR) Designs, dated July 21, 1993.
(4) There is one major flaw in the "SSE only" design approach. The equipment designed for SSE is limited to the equipment necessary to assure the integrity of the reactor coolant pressure boundary, to shutdown the reactor, and to prevent or mitigate accident consequences. The equipment designed for SSE is only part of the equipment "necessary for continued operation without undue risk to the health and safety of the public." Hence, by thir. rule, it is possible that some equipment necessary for continued operation will not be designed for SSE or OBE effects.
The NRC does not ag m st the design approach is flawed.
It is not possible that some equipme,.>
assary for continued aft cperation will not be designed for SSE or OBE et:
ts. General Design Criterion 2, " Design Bases for Protection Against Natural Phenomena," of f.ppendix A, " General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50 requires that nuclear power plant structures, systems, and components important to safety be designed to withstand the efft ts of earthquakes without loss of capability to perform their safety functions. The criteria in Appendix S to 10 CFR Part 50 implement Ger.aral Design Criterion 2 insofar as it requires structures, systems, and components important to safety to withstand the effects of earthquakes.
Regulatory Guide 1.29, " Seismic Design Clar sification,"
describes a method acceptable to the NRC for identifying and classifying those features of light-water-cooled r.uclear power plants that should he designed to withstand the effects of the SSE.
The American Society of Civil Enginec: 5 lASCE) recommended that the seismic design and engineering criteria of ASCE Standard 4, " Seismic An& sis of Safety-Related Nuclear Stru:tures and Cosmientary on Standard for Seis;. c Analysis of Safety-Related Nuclear Structures," be incorporated by reference into Appendix S to 10 CFR Part 50.
The Commission nas determined that new regulations will be more streamlined containing only basic requirements with guidance being provided in regulatory guides and, to some extent, in standard review plan sections. Both the NRC tnd industry have experienced difficultias in applying prescriptive l
regulations such as Appendix A to 10 CFR Part 100 because they inhibit the use of needed latitude in judgement. Therefore, it is cornon NRC practice not 'o l
reference publications such as ASCE Standard 4 (an ualysis, not design standard) in its regulations. Rather, publications such as ASCE Standard 4 are cited ira regulatory guides and standard review plan section:. ASCE Standard 4 is cited in the 1989 revision of Standard Review Plan Sections 3.7.1, 3.7.2, and 3.7.3.
19
The Department of Energy stated that the required consideration of aftershocks in Paragraph IV(B), Surface Deformation, is confusing and recome~ led that it be deleted.
T 4 NRC agrees. The reference to aftershocks in Paragraph IV(b) has been deleted.
Paragraphs VI(a), Safe Shutdown Earthquake, and VI(B)(3) of Appendix A to Part 100 contain the phrase " including aftershocks." The
" including aftershocks" phrase was removed from the Safe Shutdown Earthquake Ground Motion requirements in the proposed regulation. The recommended change will make Paragraphs IV(a)(1), " Safe Shutdown Earthquake Ground Motion," and IV(b), " Surface Deformation, of Appendix S to 10 CFR Part 50 consistent.
X. Finding of No Significant Environmental Impact: Availability The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in Subpart A of 10 CFR Part 51, that this pr:;:::d regulation, if :d:;ted, 2:ld :nt 5: j'sMa major Federal action significantly affecting the quality of the human environment and therefore an environmental impact statement is not required.
The revisions asscciated with the reactor siting criteria in 10 CFR Part 100 and the relocation of the plant design requirements from 10 CFR Part 100 to 10 CFR Part 50 have been evaluated against the current requirements. The Commission has concluded that relocating the requirement for a dose calculation to Part 50 and adding more specific site criteria to Part 100 does not decrease the protection of the public health and safety over the current regulations. The pr:p:::d amendments do not affect nonradiological plant effluents and have no other environmental impact.
The addition of s100.23 to 10 CFR Part 100, and the addition of Appendix i
S to 10 CFR Pirt 50, will not change the radiological environmental impact offsite. Onsite occupational radiation exposure associated with inspection and maintenance will not change. These activities are principally associated with base line inspections of structures, equipment, and piping, and with maintenance c'
- sic in:,trumentation. Base line inspections are needed to differentiate... een pre-existing conditions at the nuclear power plant and earthquake related damage.
The structures, equipment and piping selected for these in:,pections are those routinely examined by plant perators during nomal plant walkdowns and inspections. Routine maintenance of seismic instrumentation ensures its operability during earthquakes. The location of th'e seismic instrumentation is similar to that in the existing nuclear power plants. The pr:p:::d amendments do not affect nonradiological plant effluents and have no other environmental impact.
The environmental assessment and finding of no significant impact on wt.ich this determination is based are available for inspection at the NRC Peblic Document Room, 2120 L Street NW. (Lower Level), Washington, DC. Single copies of the environmental assessment and finding of no significant impact HisTEiiiiERhiiEWW6cfiiffliif~fM~~((TsiiQ~, Office of Mu are available from Mr. Leonard Soffer U.S. Nuclear Regulatory Commission, Tsh'i%tonTDC~~2b55CGlipbKi 301) 415-1722, or Dr. Andrew J. Murphy, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 415-6010.
XI. Paperwork Reduction Act Statement 20
_n.,
This final rule amends inforration collection requirements that are subject to the Paperwork ReductioF Act of These requirements were approved by the Office of Management and approval numbers 3150-0011 and 3150-0093.
The public reporting burden for this colle
. estimated to average 800,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> per response,ction of information is including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information.
Send comments on any aspect of this collection of information, including suggestions for reducing the burden, to the Information and Records Management Branch (T-6 F33)..U.S. Nuclear Regulatory Commission, Washingto DC 20555-0001, Affairs, NE06-10202, (3150-0011 and' 3150-0093),and to the Desk O Budget, Washington, DC 20503.
Office of Management and Public Protection Notification 7
The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number.
XII. Regulatory Analysis l
The Commission has prepared a dr:ft regulatory analysis on this pr:;;; d l
regulation.
The analysis examines the costs and benefits of the alternatives considered by the Commission..Th: dr:ft :::l";i: i: ;":i':ble f:r in:;:: tic.
Dot.
copies of the analysis are available from Mr. Leonard Soff n
Office of E :1::r " ;;1:t:ry "::::rch Nuclear Regulatory Commissio'W,ijT T T E ~^ "RIM'er
.., U.S.
asnington,
, telephone (,3'01) 415-1722, n
or Dr. Andrew J. Murph Regulatory Commission,y, Office of Nuclear Regulatory Research, U.S. Nuclear Washington, DC 20555, telephone (301) 415-6010.
The C:n i::i r:;;;;t: ;;hli: :: n nt :n th dr:ft r;;;1:t:ry :::1y:i:.
Cr.t: :: the dr:ft :::ly:i: :y 5: :: bitted t the ""C :: i-di::ted ::d:r th: ??a"iSEEE h:: ding.
XIII. Regulatory Flexibility Certification
!= ::::rd:::: rith "~~~~
jedIjfjtheRegulatoryFlexibilit 6, 5 U.S.C. 605(b), thN"ggjiijp~if~oYiertifies that this pr:;:::
1980 isimis d regulation
. ; jnot, if prml;:ted, have a significant economic impact on a substantial number of small entities.
the licensing and o This pr:;:::d regulation affects only
- 11
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ration of nuclear r lants. " :le:r ;=e ;1 :t :ite MiB o not fall within the definition II ---
- ' * -- ' ' ' '- "- - " "" - ' ness
^:t (15 '!.S.C. S??), th: e :ll "m:in::: Si : St :d:rd: Of th: * :11 " :in ::
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i XIV. Backfit Analysis The NRC has determined that the backfit rule, 10 CFR 50.109, does not apply to this pr:;:::d-regulation, 'and therefore, a backfit analysis is not required for this pr:;:::d regulation because these amendments do not involve i
l any provisions that would impose backfits as defined in 10 CFR 50.109(a)(1).
The pr:;:::f regulation would apply only tc applicants for future nuclear power plant construe. tion permits, preliminary design approval, final design approval, manufacturing licenses, early site reviews, operating licenses, and combined operating licenses.
~
List of Subjects s
10 CFR Part 50 - Antitrust, Classified information, Criminal penalty, Fire protection, Intergovernmental relations, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements.
10 CFR Part 52 - Administrative practice and procedure, Antitrust, Backfitting, Combined license, Early site permit, Emergency planning, Fees, Inspection, Limited work authorization, Nuclear power plants and reactors, Probabilistic risk assessment, Prototype, Reactor siting criteria, Redress of site, Reporting and recordkeeping requirements, Standard design, Standard design certification.
10 CFR Part 100 - Nuclear power plants and reactors, Reactor siting criteria.
For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 553, the PRC is proposing to adopt the following amendments to 10 CFR Parts 50, 52 and 100.
PfRT 50 - DONESTIC LICENSING 0F PRODUCTION Alm LTTILIZATION FACILITIES 1.
The authority citation for Part 50 continues to read as follows:
AUTHORITY: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68 Stat.
936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 1244, as amended 2282); secs.(42 U.S.C. 2132, 2133, 2134, 2135, 2201, 2232, 2233, 2236, 2239, 201, as amended, 202, 206, 88 Stat. 1242, as amended, 1244, 1246, (42 U.S.C. 5841, 5842, 5846).
22
Sec. tion 50.7 also issued under Pub. L.95-601, sec. 10, 92 Stat. 2951 as
. amended by Pub. L.- 102-486, sec. 2902,106 Stat.- 3123, (42 U.S.C. 5851).
Section 50.10 also issued ureder secs. 101, 185, 68 Stat. 936, 955 as amended I
(42 U.S.C. 2131, 2235), sec. 102, Pub. L.91-190, 83 Stat. 853 (42 U.S.C.
4332). Sections 50.13, 50.54(dd) and 50.103 also issued under sec. 108, 68 Stat. 939, as amended (42 U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56 also issued under sec.185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a, 50.55a and Appendix Q also issued under sec. 102, Pub. L.91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 50.58, 50.91 and 50.92 also issued under Pub. L.97-415, 96 Stat. 2073 (42 U.S.C. 2239). Section 50.78 also issued under sec.122, 68 Stat. 939 (42 U.S.C. 2152). Sections 50.80 -
50.81 also issued under sec.184, 68 Stat. 954, as amended (42 U.S.C. 2234).
Appendix F also issued under sec.187, 68 Stat. 955 (42 U.S.C. 2237).
2.
Section 550.2 is revised by adding in alphabetical order the definitions for Committed dose eauivalent, Committed effective dose eauivalent, Deen-dose eauivalent, Exclusion area, low population zone, Safety-related structures. systems. and components and Total effective dose eauivalent to read as follows:
s 50.2 Definitions.
Committed dose eauivalent means the dose equivalent to organs or tissues of reference that will be received from~an intake of radioactive material by an individual during the 50-year period following the intake.
Committed effective dose eauivalent is the sum of the products of the weighting factors applicable to each of the body organs or tissues that are irradiated and the committed dose equivalent to these organs or tissues.
Deeo-dose eauivalent, which applies to external whole-body exposure, is the dose equivalent at a tissue depth of I cm (1000mg/cm ).
e 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 th? public health and safety. Residence within the exclusion area shall normally be prohibited.
In any event, residents shall be subject to ready removal in case of necessity. Activities unrelated to operation of the reactor may be permitted in an exclusion area under appropriate limitations, provided that no significant hazards to the public health and safety will result.
l'ow oooulation 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 23
- -.. _ ~.
w measures could be taken in their behalf in the event of a ser I
These guides do not specify a permissible population density or total population within this zone beca;:se the situation may vary free case to Whether a specific number of people can,.for example, be evacuated from a many factors such as location, number and size of advance planning, and actual distribution of residents within the area.
Safety-related Structures Systems and Componenh means those i
structures, systems, and components that are relied on to remain fun during and following design basis (postulated) events to assure The capability to shutdown the reactor and maintain it in a (1)
(2) safe shutdown condition, andThe capability to prevent or mitigate the conseq accidents which could result in potential offsite (3) chapter.
Total effective dose eauivalent (TEDE) means the sum of the deep-dose equivalent (for external exposures) and the committed effective d i
equivalent (for internal exposures).
l In 550.8, paragraph (b) is revised to read as follows:
3.
OMB approval.
s 50.8 Information collection requirements:
The approved information collection requirements contained i 5550.30, 50.33, 50.33a, 50.34, 50.34a, 50.35, 50.36, 50.36a, (b) 50.48, 50.49, 50.54, 50.55, 50.55a, 50.59, 50.60, 50.61, 50.6 part appear in 50.71, 50.72, 50.80, 50.82, 50.90, 50.91, K, M, N, 0, Q, R, and S.
9 and 10 and paragraph (a)(1) is revised a i
4.
(b)(10), and (b)(11) are added to read as follows:
s 50.34 Contents of applications; technical information.
Stationary power reactor applicants for a construction permit (a) t pursuant to this part, or a design certification or (1)
E
~~
24
l e.
i i
L a FINAL RULE), shall comply with paragraph (a)(1)(ii) of this section. All othe applicants for a construction permit pursuant to this part or a design certification or combined license pursuant to Part 52 of this chapter, shall comply with paragraph-(a)(1)(1) of this section.
I (1) A description and safety assessment of the site on which the facility is to be located, with appropriate attention to featu; i
affecting facility design.
evaluation factors identified in Part 100 of this chapter. The assessment m l
contain an analysis and evaluatitei of the major stru th site under the site evaluation factors identified in Part 100 l cha ter, assuming that the facility will be operated at the ultimate powerWi lev ' which is contemplated by the applicant.
rejected initial power level, the applicant is requi the w
the information required by this paragraph, in support of the inft well
+ for a construction persit, or a design, approval.
appli A description and safety assessment of the site and a (ii)
It is expected that reactors will reflect ant of the facility.
lesign, construction and operation an extremely low probability safety as.
for accidents that could result in the release of significant quantities of through th A radioactive fission products. The following power reac
. Intended use of the reactor includi q the proposed maximum power Commission:
level and the nature and inventory of containe, radioac (A)
~
(B) applied to the design of the reactor;The extent to which the reactor inco enhanced safety features having a significant bearing on the probability o (C) consequences of accidental release of radioactive materia and those barriers that must be breached as a result of an (D) release of radioactive material to the environment radiological crinsequences of accidents. In performing this assessment, a applicant shall assume a fission product release
- from the core into containment assuming that the facility is operated at the ultimate p
~.
The applicant shall perform an evaluation and analysis of the postulated fission product release, using the expected demonstrable contemplated.
containment leak rate and any fission product cleanup systems intended mitigate the consequences of'the accidents, together with applicabl i
characteristics, including site meteorology, to evaluate the offsite
- The tission product release assmed for this evaluation should be based upon l events.
hypothestred for purposes of site analysis or postuisted f b
t i
J release into the contaisuunnt of approctable quantitles of fission products.
- _
- NG 25
1 1
a Site charatteristics must comply with Part 100 of radiological consequences.
The evaluation must detamine that:
this chapter.
(1) An individual located at any point on the boundary of the exclusion area for iny 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period following the onset of the postolated i
fission product release, would not receive a radiation dose in excess of 25 ren' total effective dose equivalent (TEDE).
(2) An individual located at any point on the outer boundary of i
the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a radiation dose in excess of 25 rem total effective dose equivalent (TEDE).
(E)
With respect to operation at the projected initial power level, the applicant is reouired to submit infomation prescribed in paragraphs
)
(a)(2) through (a)(8) of this section, as well as the infomation required by l
this paragraph, in support of the application for a construction permit, or a design approval.
(12) On or after []W$ER EFFECTIVE DATE.OF THE FINAL RULE], stationary 1
power reactor applicants who a(pply for a construction permit pursua part, or a design certification or combined license v uant to Part 52 of this chapter, as partial confomance to General Design s.rtterion 2 of Appendix j
A to this part, shall. comply with the earthquake engineering criteria in Appendix S of this part.
(b)
(10) On or after [plEEE EFFECTIVE DATE OF THE FINAL RULE], stationary power reactor applicants who apply for an operating license pursuant to this part, or a design certification or combined license pursuant to Part 52 of this chapter, as partial conformance to General Design Criterion 2 of Appendix A to this part, shall comply with the earthquake engineering criteria of However, if the construction permit was issued prior Appendix S to this part.
NSE N EFFECTIVE DATE OF THE FINAL RULE], the stationary power reactor to [I,icani. shall comply with the earthquake engineering criteria in Section VI appl of Appendix A to Part 100 of this chapter.
(11) On or after [plBEM EFFECTIVE DATE OF THE FINAL RULE), stationary power reactor applicants who apply for an operating license pursuant to this Part, or a combined license pursuant to Part 52 of this chapter, shall provide
' A whole body dose of 25 rom has been stated to correspond nwnerically io the once in a lifettav accidental or emergency dose fo radiation workers which, according to NCRP recomunendations at the the could be disregarded in the determination of their radiation exposure status (see NBS Handbook 69 dated June However, its use is not intended to imply that this number constitutes an acceptable limit for an l
Rather, this dose value has been set forth in this
- 5. 1959).
emergency dose to the public under accident conditions, section as a reference value, which can be used in the evaluation of plant design features with respect to postulated reactor accidents, in order to assure that such designs provide assurance of low risk of pub exposure to radiation, in the event of such accidents.
j 26 i
-. ~. -
=-
h facility as in l
'a description and safety assessment of the site and of t e 550.34(a)(1)(ii) of this part.*
In $50.54, paragraph (ff) is added to read as follows:
5.
550.54 Conditions of licenses.
l e ted the (ff) For licensees of nuclear power plants that have imp earthquake engineering criteria in Appendix S of t dix S. Prior to resuming required as provided in Paragraph IV(a)(3) of Appen i
ion that no operations, the licensee shall demonstrate to th for continued operation without undue risk to the health and safety o licensing basis is maintained.
Appendix S to Part 50 is added to read as follows:
6.
OR NUCLEAR POWER I
APPENDIX S TO PART 50 - EARTHQUAKE ENG PLANTS General Information tion or This appendix applies to applicants for a design certifica combined license pursuant to Part 52 of this chapter or t r on or after [111$ERT or operating license pursuant to Part 50 of this c ap eHowever h
the operating EFFECTIVE DATE OF THE FINAL RULE].
issued prior to [lIISERI EFFECTIVE DATE OF TSIE FINAL i
iteria in license applicant" shill comply with the earthquake engine Section VI of Appendix A to 10 CFR Part 100.
l I.
Introduction design Each applicant for a construction permit, operating licens certification, or combined license is required by 5 design nuclear power General Design Criterion 2 of Appendix A to this Part toto safe plant structures, systans, and components im ithout loss of capability i
i eering criteria to perform their safety functions.
power plants that have implemented the earth h IV(a)(3) of this it requires structures, systems, and com appendix are exceeded.
the effects of earthquakes.
27 I
a j
i II. Scope Th6 evaluations described in this appendix are within the scope of investigations permitted by 550.10(c)(1).
III. Definitions As used in these criteria:
4 Combined license means a combined construction permit and operating license with conditions for a nuclear power facility issued pursuant to Subpart C of Part 52 of this chapter.
]
Desian Certification means a Commission approval, issued pursuant to i
Subpart B of Part 52 of this chapter, of a standard design for a nuclear power facility. A design so approved may be referred to as a " certified standard design."
The Doeratina Basis Earthauake Ground Motion (OBE1 is the vibratory ground motion for which those features of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the i
public will remain functional. The Operating Basis Earthquake Ground Motion
{
is only associated with plant shutdown and inspection unless specifically selected by the applicant as a design input.
j
)
A response spectrum is a plot of the maximum responses (acceleration, i
velocity, or displacement) of idealized single-degree-of-freedom oscillators as a function of the natural frequencies of the oscillators for a given damping value. The response spectrum is calculated for a specified vibratory motion input at the oscillators' supports.
The Safe Shutdown Earthauake Ground Motion (SSE) is the vibratory ground 1
motica for which certain structures, systems, and components must be designed to remain functional.
The structures. systems. and components reauired to withstand the effects of the Safe Shutdown Earthauake Ground Motion or surface deformation are those necessary to assure:
(1).The integrity of the reactor coolant pressure boundary, (2) The capability to shut down the reactor and maintain it in a safe shutdown condition, or (3) The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guideline exposures of 550.34(a)(1)(ii).
Surface deformation is distortion of geologic strata at.or near the ground surface by the processes of folding or faulting as a result of va_r.ious earth forces, Tectonic surface deformation is associated with earthquake processes.
28
't IV. Application To Engineering Design, The following are pursuant to the seismic and geologic design basis requirements of s100.23 of this chapter:
(a) Vibratory Ground Motion.
(1) safe Shutdown Earthquake Ground Motion. The Safe Shutdown Earthquake Ground Motion me-t te characterized by free-field ground motion response spectra at th' N-o und surface.
In view of the limited data available oc vibratory 4
W wtions of strong earthquakes, it usually will be appropriate thr.t cM M p e wponse spectra be smoothed spectra. The horizontal ~
w ", t
- t w W e Shutdown Earthquake Ground Motion in the frea-fie~
<, 4 :..adation level of the structures must be an appropriate respony scrum with a peak ground acceleration of at least 0.19 4eclear power plant must be designed so that, if the Safe Shutdown Earthquake Ground Motion occurs, certain structures, systems, and components will remain functional and within applicable stress, strain, and deformation jg i
limits. In addition to seismi:: loads, applicable concurrent normal operating, functional, and accident-induced loads must be taken into account in the design of thase safety-related structures, systems, and components. The desig of the nuclear power f ant 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, arJ 1andsliding, as required in s100.23 to Part 100 of this chapter.
The required safety functions of structures, systems, and components must be assured during and after the vibratory ground motion associated with the Safe Shutdown Earthquake Ground Motion through design, testing, or qualification methods.
The evaluation must take into account soil-structure interaction effect 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.
by resp (on)se spectra.The Operating Basis Earthquake Ground Motion must be i
Motion must be set to one of the following choices:The value of the Opera e'
(A) One-third or less of the Safe Shutdown Earthquake Ground Not %
design response spectra.
Basis Earthquake Ground Motion in Paragraph (a)(2The requirements a without the applicant performing explicit response)(or)(design analyses, or 1 B)(I) can be satisfieo (B) A value greater than one-third of the Safe Shutdown Earthquake Ground Motion design response spectra.
Analysis and design must be performed i
to demonstrate that the requirements associated with this Operating Basis Earthquake Ground Motion in Paragraph (a)(2)(1)(B)(I) are satisfied.
i The design must take into account soil-structure interaction effects and the duration of vibatory ground motion.
(I) When subjected to the effects of the Operating Basis Earthquake Ground Motion in combination with normal operating loads, all structures, 29 4
l systems, and components of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the public-must remain functional and within applicable stress, strain, and deformation limits.
(3) Required Plant Shutdown.
If vibratory ground motion exceeding that of the Operating Basis Earthquake Ground Motion or if significant plant damage occurs, the licensee must shut down the nuclear power plant.
If systems, structures, or components necessary for the safe shutdown of the nuclear power ant are not available after the occurrence of the GBE-BM- ' ' M
, the licensee must consult with tie Comm'ss' on and aus propose a p an the timely, safe shutdown of the nuclear power plant.
Prior to resuming operations, the licensee must demonstrate to the Commission that no functional damage has occurred to those features necessary for i
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 tale into I
account seismic loads, inchdin;; cfter:h::k:, and applicable conc urrent functional and accident-induced loads. The design provisions for surface deformation must be based on its postulated occurrence in any direction and i
azimuth and under any part of the nuclear power plant, unless evidence indicates this assumption is not appropriate, and must take into account the 1
estimated rate at which the surface deformation may occur.
l (c) Seismically Induced Floods and Water Waves and Other Design l
Conditions. Seismically induced floods and water waves from either locally or distantly generated seismic activity and other design coMit. ions determined pursuant to s100.23 of this chapter must be taken into acuunt 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 l
LICENSES FOR NUCLEAR' POWER PLANTS 1
7.
The authority citation for Part 52 continues to read as follows:
AUTHORITY: Secs. 103, 104, 161, 182, 183, 186, 189, 68 Stat. 936, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 1244, as amended (42 0.S.C.
2133, 2201, 2232, 2233, 2236, 2239, 2282;; secs. 201, 202, 206, 88 Stat
- 1242, 1244, 1246, as amended (42 U.S.C. 5841, 5842, 5846).
8.
In 552.17, the introductory text of paragraph (a)(1) and paragraph (a)(1)(vi) are revised to read as follows:
1 30 l
557.17 Contents of applications.
~
(a)(1) The. application must contain the information required by s 50.33(a)-(d), the information required by s 50.34 (a)(12) and (b)(10), and to the extent approval of emergency plans is sought under paragraph (b)(2)(ii) of i
this section, the information required by s 50.33 (g) and (j), and s 50.34 (b)(6)(v). The' application must also contain a description and safety assessment of the site on which the facility is to be located. The assessment must contain an analysis and evaluation of the major structures, systems, and components of the facility that bear significantly on the acceptability of the site under the radiological consequence evaluation factors identified in s j
50.34(a)(1) of this chapter. Site characteristics must comply with Part 100 of this chapter.
In addition, the application should describe the following:
(vi) Tha seismic, meteorological, hydrologic, and geologic characteristics of the proposed site;
\\
PART 100 - REACTOR SITE CRITERIA 9.
The authority citation fr Part 100 continues to read as follows:
i 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).
10.
The table of contents for Part 100 is revised to read as follows:
PART 100 - REACTOR SITE CRITERIA Sec. 100.1 Purnose.
1 100.2 Scope.
100.3 Definitions.
100.4 Commur.ications.
100.8 Information collection requirements: OMB approval.
Subpart. A - Evaluation Factors for Stationary Power Reactor Site Applications Before [ EFFECTIVE DATE OF THE FINAL RilLE] and for Testing Reactors.
100.10 Factors to be considered when evaluating sites.
100.11 Determination of exclusion area, low population zone, and population center distance.
Subpart B -Evaluation Factors for Stationary Power Reactor Site Applications on or After [ EFFECTIVE DATE OF-THE FINAL RULE).
100.20 Factors to be considered when evaluating sites.
100.21 Non-seismic site criteria.
100.23 Geologic and seismic siting criteria.
31
APPENDIX A-Seismic and Geologic Sitir3 Criteria for Nuclear Power Plants.
11.
Section100.1isrevised50readasfollows:
s 100.1 Purpose.
(a) The purpose of this part is to establish approval requirements for proposed sites for stationary power and testing reactors subject to Part 50 or Part 52 of this chapter.
(b) There exists a substantial base of knowledge regarding power reactor siting, design, construction and operation. This base reflects that the primary factors that determine public health and safety are the reactor design, construction and operation.
(c) Siting factors and criteria are important in assuring that radiological doses from nomal operation and postulated accidents will be acceptably low, that natural phenomena and potential man-made hazards will be appropriately accounted for in the design of the plant, and that the site characteristics are amenable to the development of adequate emergency plans to protect the public and adequate security measures to protect the plant.
(d) This approach incorporates the appropriate standards and criteria for approval of stationary power and testing reactor sites. The Commission intends to carry out a traditional defense-in-depth approach with regard to reactor l
siting to ensure public safety. Siting away from densely populated centers has been and will continue to be an important factor in evaluating applications for l
site approval.
12.
Section 100.2 is revised to read as follows:
s 100.2 Scope.
The siting requirements contained in this part apply to applications for site approval for the purpose of constructing and operating stationary power and testing reactors pursuant to the provisions of Parts 50 or 52 of this chapter.
13.
Section 100.3 is revised to read as follows:
s 100.3 Definitions.
As used in this part:
Combined license means a combined construction permit and operating license with conditions fo* a nuclear power facility issued pursuant to Subpart C of Part 52 of this chapter.
Early Site Permit means a Commission approval, issued pursuant to subpart A of Part 52 of this chapter, for a site or sites for one or more nuclear power facilities.
Exclusion area means that area surrounding the reactor, in which the veactor 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 32 l
=
the highway, railroad, or waterway, in case of emergency, to protect the public health and safety.
Residence within the exclusion area shall normally be prohibited.
In any event, residents shall be subject to ready removal in case of necessity. Activities unrelated to operation of the reactor may be permitted in an exclusion area under appropriate limitations, provided that no significant hazards to the public health and safety will result.
Low nooulation 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 I
specify a pemissible population density or total population within this zone l
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 l
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 nuciear reactor of a type described in 5550.21(b)or 50.22 of this chapter designed to produce electrical or heat energy.
A Response spectrum is a plot of the maximum responses (acceleration, velocity, or displacement) of idealized single-degree-of-freedom oscillators as a function of the natural fr equencies of the oscillators for a given damping value.
The response spectrum is calculated for a specified vibratory motion input at the oscillators' supports.
The Safe Shutdown Earthouake Ground Motion ta the vibratory ground motion for which certain structures, systems, and components must be designed pursuant to Appendix S to Part 50 of this chapter to remain functional.
Surface deformation is distortion of geologic strata at or near the ground surface by the processes of folding or faulting as a result of various earth forces. Tectonic surface deformation is associated with earthquake processes.
Testino reactor means a testino facility as defined in 550.2 of this chapter.
- 14. Section 100.4 is added to read as fol1%
s100.4 Communications.
Except where otherwise specified in this part, all correspondence, reports, applications, and other written communications submitted pursuant to 10 CFR 100 should be addressed to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555, and copies sent to the appropriate Regional Office and Resident Inspector. Communications and reports may be delivered in person at the Commission's offices at 2120 L Street, NW., Washington, DC, or at 11555 Rockville Pike, Rockville, Maryland.
15.
Section 100.8 is revised to read as follows:
33
O q 1
ONB approval.
s 100.8 Information collection requirements:
l (a) The Nuclear Regulatory Commission has submitted' the infol collection requirements contained in this part to the Office of Manage 1980 (44 3
Budget (02) for approval as required by the Paperwork Red
^'
U.S.C. 3501 et seq.).
3150-0093.
contained in this part under control number (b) The approve
/
part appear in s100.23 and Appendix A.
A heading for Subpart A isi added directly before $100.10 to rea 16.
i follows:
Subpart A - Evaluation Factors for Stationary Power-Reactor S Applications before [ EFFECTIVE DATE OF THIS REGULATIO lD Reactors.
Section 100.10 is revised to read as follows:
17.
be considered when evaluating sites.
Factors t._asidered in the evaluation of sites include those relati s100.10 Factors It to the proposed reactor design and the characteristics peculiar to the is expected that reactors will reflect through their design, constru operation an extremely low probability for accidents that could r of significant quantities of radfoactive fission products. In addition, t location and the engineered featurcs included as safeguards again consequences of an accident, should <.ne occur, should insure a
~~
In particular, the t. omission will take the following factors into 8
consideration in determining the acceptability of a site for a power o exposure.
(a) Intended use of the reactor including the pr reactor:
and the nature and inventory of contained radioac (1) applied to the design of the reactor;
~
having a significant bearing on the probability m
1 release of radioactive materials; (4) The safety features that are to be engineered into the l
those barriers that must be breached as a result of an of radioactive material to the environment can occur.
(b) Population der.sity and use characteristics of the site env including the exclusior, area, iow population zone, and the popula 1
(c)
Physical characteristics of the
- site, including seismology, distance.
meteorology, geology, and hydrology.
(1) Appendix A to Part 100, " Seismic and Geologic Siting Nuclear Power Plants," describes the nature of investigations req l
the geologic and seismic data necessary to determine site suit provide reasonable assurance ~that a nuclear power plan
. my. ~
34 r
,ww,
-y.
r--.c~--,--r w
. - - - - -,,,. - - -, - - - +
=.
1 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 I
designed to withstand the effects of surface faulting.
)
(2) Meteorological conditions at the site and in the surrounding area j
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 j
facility. Special precautions should be planned if a reactor is to be located at a site where a significant quantity of radioactive effluent might accidentally flow into nearby streams or rivers or might find ready access to underground water tables.
(d) Where unfavorable physical characteristics of the site exist, the proposed site may nevertheless be found to be acceptable if the design of the I
facility includes appropriate and adequate compensating engineering safeguards.
18.
Section 100.11 is revised to read as follows:
5100.11 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 the containment and the meteorological conditions pertinent to his site to derive an exclusion area, a low population zone and population center distance. For the purpose of this analysis, which shall set forth the basis for the numerical values used, the applicant should determine the following:
(1) An exclusion area of such size that an individual located at any point on its boundary for two hours immediately following onset of the postulated fission product release would not receive a total radiation dose to the whole body in excess of 25 res* or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
(2) A low population zone of such size that an individual located at any point on its outer boundary who is exposed to the radio ::tive 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
'The fission product release assumed for these calculations should be based upon a major accident, hypothesired 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 nwnerically to the once in a lifetime accidental or emergency dose for radiation workers which, according to NCRP reconsuendations 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 while 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 rSattor sites with respect to potential reactor accidents of exceedingly low probability of occurrence, and low risk of pr.
exposure to radiation.
35
W -
25 ren or a total.r<ation dose in excess of 300 rem to the thyroid from iodine exposure.
(3) A pr 21ation center distance of at least one and one-third times the distance fror.he reactor to the outer boundary of the low population zone. In applying this guide, the boundary of the population center shall be determined i
i upon consideration of population distribution.
Political boundaries are not i
controlling in the application of this guide.
Where very large cities are involved, a greater distance may be necessary because of total integrated 1
j population dose consideration.
i (b) For sites for multiple reactor facilities consideration should be given i
to the following:
(1) If the reactors are independent to the extent that an accident in one 4
i reactor would not initiate an accident in another, the size of the exclusion area, low population zone and population center di. stance shall be fulfilled with respect to each reactor individually. The envelopes of the plan overlay of the areas so calculated shall then be taken as their respective boundaries.
(2) If the reactors are interconnected to the extent that an accident in one reactor could affect the safety of operation of any other, the size of the exclusion area, low population zone and population center distance shall be based upon the assumption that all interconnected reactors emit their postulated fission product releases simultaneously.
This requirement may be reduced in relation to the degree of coupling between reactors, the probability of concomitant accidents and the probability that an individual would not be exposed to the radiation effects from simultaneous releases.
The applicant would be expected to justify to the satisfaction of the Commission the basis for such a reduction in the source ters.
(3) The applicant is expected to show that the simultaneous operation o.f multiple reactors at a site will not result in total radioactive effluent releases beyond the allowable limits of applicable regulations.
NOTE: For further guidance in developing the exclusion area, the low population zone, and the population center distance, reference is made to Technical Information Document 14844, dated March 23,1%2, 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 Comission, Washington, DC 20555.
19.
Subpart B (ss100.20 - 100.23) is added to read as follows:
Subpart B - Evaluation Factors for Stationary Power Reactor Site Applications on or After [ EFFECTIVE DATE OF THE FINAL RULE).
s100.20 Factors to be considered when evaluating sites.
36
/
The Coinnission will take the following factors into consideration in determining the acceptability of a site for a stationary power reactor:
4 (a) Population density and use chaFacteristics of the ' site eratrons, including the exclusion area, the population distribution, and site-related characteristics must be evaluated to determine whether individual as well as j
societal risk of potential plant accidents is low, and that site-related characteristics would not prevent the development of a plan to carry out suitable protective actions for members of the public in the event of emergency.
(b) The nature and proximity of man-related hazards (e.g., airports, dams, transportation routes, military and chemical facilities) must be evaluated to establish site parameters for use in determining whether a plant design can accommodate commonly occurring hazards, and whether the risk of other hazards is i
very low.(c) Physical characteristics of the site, including seismology, meteorology, geology, and hydrology.
(i) s100.23, " Geologic and seismic siting factors," of this part describes the criteria and nature of investigations required to obtain the i
geologic and seismic data necessary to determine the suitability of the proposed
~
site and the plant design bases.
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) must be identified and characterized.
I (3) Factors important to hydrological radionuclide transport (such as soil, sediment, and rock characteristics, adsorption and retention coefficients, ground water velocity, and distances to the nearest surface body of water) must be obtained from on-site measurements. The maximum probable flood along with the 2
potential for seismically induced floods discussed in s100.23 (d)(3) of this part must be estimated using historical data.
4 1
s 100.21 Non-seismic siting criteria.
l Applications for site. appr3 val for comercial power reactors shall demonstrate that the proposed site meets the follating criteria:
l (a) Every site must have an exclusion area and a low population zone, as defined in $100.3; (b) The population center distance, as defined in $100.3, must be at least one and one-third times the distance from the reactor to the outer boundary of the low population zone. In applying this guide, the boundary of the population center shall be determined upon consideration of population distribution.
Political boundaries are not controlling in the application of this guide; (c) Site atmospheric dispersion characteristics must be evaluated and dispersion parameters established such that:
i (1) Radiological effluent release limits associated with normal operation from the type of facility proposed-to be located at the site can be met for any individual located offsite; and (2) Radiological dose consequences of postulated accidents shall meet the criteria set forth in 550.34(a)(1) of this chapter for the type of facility proposed to be located at the site; 37 t
J
'(d) The physical characteristics of the site, including meteorology,
, geology, seismology, and hydrology must be evaluated and site. parameters established such that potential threats from such physical characteristics will pose no undue risk to the type of facility proposed to be located at the site; (e) Potential hazards associated with nearby. transportation routes, industrial and military facilities must be evaluated and site parameters t
established such that potential hazards from such routes and facilities will pose
{
no undue risk to the type of facility proposed to be located at the site; (f) Site characteristics must be such that adequate security plans and measures can be developed, (g) Site characteristics must be such that adequate plans to take protective actions for members of the public in the event of emergency can be developed:
{
i (h) Reactor sites should be located away from very densely populated centers. Areas of low population density are, generally, preferred. However, in determining the acceptability of a particular site located away from a very densely populated center but not in an area of low density, consideration will be given to safety, environmental, economic, or other factors, which may result in the site being found acceptable *.
I I
s 100.23 Geologic and seismic siting factors.
f This section sets forth the principal geologic and seismic considerations I
that guide the Commission in its evaluation of the suitability of a proposed site
}
and adequacy of the design bases established in consideration of the geologic and i
seismic characteristics of the proposed site, such that, there is a reasonable j
assurance that a nuclear power plant can be constructed and operated at the l
proposed site without undue risk to the health and safety of the public.
Ar;;11 cations to engineering design are contair.ec in Appendix S to Part 50 of this i
chapter.(a) Applicability. The requirements in paragraphs (c) and (d) of this section apply to applicants for an early site permit or combined license pursuant l
to Part 52 of this chapter, or a construction permit or operating license for a nuclear power plant pursuant to Part 50 of this chapter on or after [W EFFECTIVE DATE OF. _THE. FINAL RULE].
However, if the construction permit was issued prior to [...
.. EFFECTIVE DATE OF THE FINAL RULE], the operating license applicant shall coup y with the seismic and geologic siting criteria in Appendix A to Part 100 of this chapter.
e
- Examples of these factors include, but are not limited to, such factors as the higher population density site having superior seismic characteristics, better access to skilled labor for construction, better rail and highway access, shorter transmission line requirements, or less envirornmental impact on undeveloped areas, wetlands or endangered species, etc. some of these factors are included in, or impact, the other criteria included in this section.
38
~-
- _w_,
.-.r,,.c
(b) causencement of construction. The investigations requi (c) of this section are within the scope of investigations pemitt l
The and engineering characteristics.
50.10(c)(1) of this chapter.
d its (c) Geological, seismological, geological, seismological, and engineering characteristics of a si environs must be investigated in sufficient scope and detail to pel adequate evaluation of the proposed site, t l
Earthquake Ground Motion, and to permit adequate enginee The size of the or potential geologic and seismic effects at the proposed site.
l region to be investigated Lnd the type of data pertinent to th d
t aust be detemined based on the nature of the region surrounding t Data on the vibratory ground motion, tect.,nic surface defoma d slip l
nontectonic deformation, earthquake recurrence site.
field l
must be obtained by reviewing pertinent literature and s"
l d
seismic factors (for example, volcanic activity) that may affect the investigations.
h operation of the proposed nuclear power plant irrespective of i
i factors are explicitly included in this section.
(d) Geologic and seismic siting factors. The geologic and factors considered for design must include a d
d nontectonic defomations, the design bases for :cismically induced f this water waves, and other design conditions as stated in paragraph (d The Safe (1) Determination of the Safe Shutdown Earthquake Ground Mo section.
for the site hor'zontal and vertical free-field ground motion response spec Shutdown Earthquake Ground Motion The Safe Shutdown Earthquake Ground Motion for the si determined considering the results of the investigations requ ground surface.
These Uncertainties are inherent in such estimates.
h as a uncertainties must be addressed through an appropriate analysis, su (c) of this section.
probabilistic seismic hazard analysis or suitable sensitivity I
Paragraph IV(a)(1) of Appendix S to Part 50 of tnis chapter Safe Shutdown Earthquake Ground Motion for design.
(2) Determination of the potential for surface be provided to clearly establish whether there is a potentia defomations.
(3) Determination of design bases for seismically induce deformation.
d affect waves. The size of seismically induced floods and water waves t t be a site from either locally or distantly generated seismic activity mu Sitinn (4) Determination of siting factors for other design conditions determined.
factors for other design conditions that must be evaluated and artificial slope stability, 1
stability, liquefaction potential, natural licant cooling water supply, and remote safety-related structure siting h as, the shall evaluate all siting factors and potential c tion, and d
4 w.
i 39 1
t i
the effects of vibratory ground motion that may affect the design and operation of the proposed nuclear power plant.
f Dated at Rockville, Maryland, this day of i
For the Nuclear Regulatory Commission.
1 John C. Hoyle, Acting Secretary of the Commission.
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Revision 2 1
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2 REGULATORY GUIDE 1.12 l
3 (Draft was DG-1033) 4 NUCLEAR POWER PLANT INSTRUNENTATION FOR EARTHQUAKES i
i 5
A.
INTRODUCTION 6
In 10 CFR Part 20, " Standards for Protection Against Radiation," licens-7 ees are required to make every reasonable effort to maintain radiation 8
exposures as low as is reasonably achievable.
Paragraph IV(a)(4) of Peepesed
)
9 Appendix S, " Earthquake Engineering Criteria for Nuclear Power Plants," to 10 10 CFR Part 50, " Domestic Licensing of Production and Utilization Facilities,"
11
- 1d requires that suitable instrumentation Esisjbe provided so that the 12 seismic response of nuclear power plant features important to safety can be l
13 evaluated promptly-[ag rianload@iiika,. Paragraph IV(a)(3) of "r:p;;cd e
14 Appendix S to 10 CFR Part 50 ;;;1d requires shutdowa of the nuclear power 15 plant if vibratory ground motion exceeding that of the operating basis 16 earthquake ground motion (OBE) occurs.'
17 This guide 1: bein; devel:p:d te describes seismic instrumentation 18 acceptable to the NRC staff for satisfying the requirements of Parts 20 and 19 50 :nd th: Pr:p;; d Appendix S to Part 50.
20 Regul:t:ry ;;id:: r i::::d i d:: crib :nd ::he :v:il:ble to the 21
- 5li
- ch infer::ti:n :: ::thed: ::: pt:ble t: the "RC :t:ff for 22 i=p1;;;nting :p ific p:rt: Of the C::;i::i
': re;;1:ticn:, techni;;;; :: d 23 by :t:ff in :v:1;; ting :pecific proble:: Or p :tel:ted :: ident, :nd guid:nce j
24 t: :p;110:nt:. R ;;1:tery ;;id:: r: net ::t:titute: for r ;;1: tion:, :nd 25
- n;11:nce with r ;;1:tery ;;id:: i: n:t r:;uired. R: gel:tery ;;id:: Or:
26 i::::d ir dr:ft fer; f:r ;;tlic ::r :nt t: involve the public in th: ::rly 27
- t:; : Of dev:10 pia; th: 7:;;1:tery p::iti:n:. Or:ft rt;;1:tery ;;id:: h:ve 28 not rectie:d ;;;1et: :i:ff revier :nd de :t repre: rt Offici:1 "RC :t:ff 29 p :iticn:.
30
'Cuid:::: i: being dev ! p:d in Dr:ft Regulatory Guide DC 1031 (({6(,
31
" Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator 32 Postearthquake Actions," en-pfs[idisicriteria for plant shutdown.
j i
i i
=-
j j
1 Any information collection activities mentioned in this dr:ft regulatory 2
guide are contained as requirements in the pr:pmed :=nd=nt: to 10 CFR 3
Part 50ZMKp[roifdii th:t ;;;1d pr vide the reguhtory basis for this 4
4 guide. The pr:p=d.=nd=nt: h=: ben ::b-itted i: JifssisfijQsjJssf@
5 EsIIFs._Ei_st_s_YliT10TC.Fh_iPiiE50EG.is4_isi. !ijijiWisidiliilthe 0ffice of Management
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6 and Budget f:r cle==ce th:t =y S :p;repri t =d:r th: Np rnrk " d :ti=
7 Act.
O ch 01: = = :0, if Ott:in:d,.;=ld :h: :pply 10 =y ins =t4en 8
eclh ti= =tivitie: = ti=:d in thi: guid:RAjii@jia1[MMM00fl.
5 9
B.
DISCUSSION 10 When an earthquake occurs, it is important to take prompt action to 11 assess the effects of the earthquake at the nuclear power plant. This 12 assessment includes both an evaluation of the seismic instrumentation data and 13 a plant walkdown. Solid-state digital time-history accelerographs installed 14 at appropriate locations will provide time-history data on the seismic 15 response of the free-field, containment structure, and other pijMilCategory 16 I structures. The instrumentation should be located so that a comparison and 17 evaluation of such response may be made with the design basis and so that 18 occupational radiation exposures associated with their location, installation, 19 and maintenance are maintained as low as reasonably achievable (ALARA).
T_ii_iru_me_nfa_ff_e~i_ wit ~7_Fd_i_dT_iii_llii3_Fi_i_4fE_TF._IE_3?f_66E_dif_f3_E_7137e 20 M
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21 I._Fi_li_iif_fdiff_hT_Sii_ss_Eti__lig_oryj_l?_st_Fs._Eis_F_e_WFree-field instrumentation
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22 data would be used to compare measured response to the engineering evaluations 23 used to determine the design input motion to the structures and to determine 24 whether the OBE has been exceeded (see Dr:ft Regulatory Guide DC 1034 @).
25 B=d ti= 1:=1 intr==nt:ti= =;1d pr=ide d:t: = th: =t=1 sci =ic 26 input t th: :=t:ir.=nt =d Other building: =d.:=ld q =tify differ = ::
27 bet== the vibr:t:ry gr=nd =ti= :t the fre field =d :t the f:=d:ti=
28 1 ::1. The instruments located at the foundation level and at elevation in 29 the structures measure responses that are the input to the equipment or piping 30 and would be used in long-term evaluations (see Dr:ft Regulatory Guide DC 1035 31 Q(1, " Restart of a Nuclear Power Plant Shut Down by a Seismic Event").
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35 si3hi]fisidillRlisilj Instrumentation is not located on equipment, piping, 2
1 or supports since experience has shown that dri.a obtained at these locations 2
are obscured by vibratory motion associated with normal plant operation.
3 The guidance being d: :1 p:d in Dr:ft Regulatory Guide 00 1034 Q 66]is 4
based on the assumption that the nuclear power plant has operable seismic 5
instrumentation, including the equipment and software needed to process the 6
data within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after an earthquake. This is necessary to determine 7
whether plant shut down is required. This determination will be made by 8
comparing the recorded data against OBE exceedance criteria and the results of 9
the plant walkdown inspections that take place within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of the event.
10 It may not be necessary for identical nuclear power units on a given 11 site to each be provided with seismic instrumentation if essentially the same 12 seismic response at each of the units is expected from a given earthquake.
13 An evaluation of seismic instrumentation noted that instruments have 14 been out of service during plant shutdown and sometimes during plant 15 operation. The instrumentation system should be operable and operated at all 16 times.
If the seismic instrumentation or data processing hardware and 17 software necessary to determine whether the OBE has been exceeded is 18 inoperable, the guidelines in Appendix A to Or:ft Regulatory Guide DG-WM 19 Q E ::ld Q ))]be used.
20 The characteristics, installation, activation, remote indication, and 21 maintenance of the instrumentation are described in this guide to help ensure 22 (1) that the data provided are comparable with the data used in the design of 23 the nuclear power plant, (2) that exceedance of the OBE can be determined, and 24 (3) that the equipment will perform as required.
25
$stJQlf{t@j!Kjysii{Eii(MiliisBjg!E@]iladiG}MMQ}s6ii 26 Sij{@ R E R HfMiliR @ @ @]{@ Qa(sf @ g ( M {fiti3]M H E jfj 27 F {sjj @ M rXp s ls Q R$ece@ Mil {@ H $ @ {s@E]ji{ M { @ M }&
28
@me,}eljlMEWigMjjiQliifssij(ld!))s]Qi[@@jisjsil!EflM5Eis 29 I@ibilj}ljMM7@{@jiigjQEM@l6ilEf@[a3ME@iMMlM2 Q0]Indon(@@MKdEal ghjpfli[EI@3f(@MKili{rT;is6}Ml6(jf i
30 31
[cjihi@j{fi@@f((fiRegcei1FMa@{i{j{Q]{@jfjfsQ 32 The appendix to this guide provides definitions to be used with this 33 guidance.
I 34 Sold:r: Of :n Op;r: ting licen;; er ::::tructi n pe;-it i::: d pri;r 10 35 the i p1;= nt:tica date 10 bc :pecified in the ::tiv: guid: = y volunt:rily 36 i:pl =nt th =thed; i be d:: rth:d in th: ::tive gaid: :nd th: =thed:
37 being devel;;;d in Or:ft Regul:t:ry Cuide: DC 1034, Tre E:rth:;;;ke Pl:nning 3
I cnd IT : dict: M;;10:r P;;;r Pl:nt Oper t0r P;;te:rthq::k: Acti:n:," :nd DC 2
1035, "R::t:rt :f : M :le:r P;;;r P1:nt Shut C;;r. by : Sci;:i: Event."
3 C. REGULA10RY POSITION 4
The type, locations, operability, characteristics, installation, 5
actuation, remote indication, and maintenance of seismic instrumentation j
\\
6 described below are acceptable to the NRC staff for satisfying the require-7 ments in 10 CFR Part 20, 10 CFR S0.55(b)(2), and Paragraph IV(a)(4) of 8
Pr:;;; d Appendix S to 10 CFR Part 50 for ensuring the safety of nuclear power 9
plants.
10 1.
SEISMIC INSTRUMENTATION TYPE AND tOCATION 11 121 Solid-state digital instrumentation that will enable the 12 processing of data at the plant site within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of the seismic event l
13 should be used.
l 14 112 A triaxial time-history accelerograph should be provided at each 15 of the following locations:
16 1.
Free-field.
17 2.
Containment foundation.
l 18 3.
Two elevations (excluding the foundation) on a structure 19 internal to the containment.
l 20 4.
An independent @[jIp(ij[TiCategory I structure foundation where 21 the response is different from that of the containment 22 structure.
23 5.
An elevation (excluding the foundation) on the independent 24 Si]ji[1]} Category I structures selected in 4 above.
25 6.
If seismic isolators are used, instrumentation should be l
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placed on both the rigid and isolated portions of the same 26 4
i 1
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or an adjacent structure, as appropriate, at approximately 2
the same elevations.
1 1
3 121 The specific locations for instrumentation should be determined by 4
the nuclear plant designer to obtain the most pertinent information consistent 5
with maintaining occupational radiation exposures ALARA for the location, 6
installation, ar.d maintenance of seismic instrumentation.
In general:
7 1.3.1 The free-field sensors should be located and installed so that @KMRTEMEMyJEMR@f@]the effects that 8
9 are associated with ::rt:f: i@((]@[] features, buildings, and components will 10 be absent from the recorded ground motion.
11 1 '.Z The instrumentation should be placed at locations that have 12 been modeled as mass points in the building dynamic analysis so that the 13 measured motion can be directly compared with the design spectra. The 14 instrumentation should not be located on a secondary structural frame member 15 that is net modeled as a mass point in the building dynamic model.
16 1.3.3 A design review of the location, installation, and 17 maintenance of proposed instrumentation for maintaining exposures ALARA should 18 be performed by the facility in the planning stage in accordance with 19 Regulatory Guide 8.8, "Information Relevant to Ensuring that Occupational 20 Reiiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably 21 Achievable."
1 22 1.3.4 Instrumentation should be placed in a location with as low a 23 dose rate as is practical, consistent with other requirements.
24 1.3.5 Instruments should be selected to require minimal 25 maintenance and in-service inspection, as well as minimal time and numbers of 26 personnel to conduct installation and maintenance.
5
I 2.
INSTRUMENTATION AT MULTI-UNIT SITES 2
Instrumentation in addition to that installed for a single unit will not 3
be required if essentially the same seismic response is expected at the other 4
units based on the seismic analysis ust.d in the seismic design of the plant.
5 However, if there are separate control rooms, annunciation should be provided 6
to both control rooms as specified in Regulatory Position 7.1 7
3.
SEISMIC INSTRUMENTATION OPERABILITY 8
The seismic instrumentation should operate during all modes of plant 9
operation, including periods of plant shutdown. The maintenance and repair 10 procedures should provide for keeping the maximum number of instruments in 11 service during plant operation and shutdown.
12 4.
INSTRUMENTATION CHARACTERISTICS 13 M
The design should include provisions for in-service testing. The 14 instruments should be capable of periodic channel checks during normal plant 15 operation.
16 M
The instruments should have the capability for in-place functional 17 testing.
18 M
Instrumentation that has sensors located in inaccessible areas 19 should contain provisions for data recording in an accessible location, and 20 the instrumentation should provide an external remote abrm to indicate 21 actuation.
22 4.4 After::tu:ticr.,thepjQinstrumentationshouldrecordEst 23 iii!sTM,$ ne-3 seconds of low amplitude motion prior to seismic trigger 24 actuation, continue to record the motion during the period in which thm 25 earthquake motion exceeds the seismic trigger threshold, and continue to 26 record low amplitude motion for a minimum of 5 seconds beyond the last 27 exceedance of the seismic trigger threshold.
6
1 M
The instrumentation should be capable of recording 25 minutes of 2
sensed motion.
3 M
The battery should be of sufficient capacity to power the 4
instrumentation end-jijsense and record (see Regulatory Position 4., 25 5
minutes of motiot,-eith r.: b:ttery ch:rger, over a period of pe+
as than the 2-6 channel check test interval (Regulatory Position 8.2). 57o^ M EsiiiilfMEEEUKEauliEMWFME!EfifXsRhoNifBfil 7
jlf Qijnii!M{@] ' $3]EWJi23]h3Efji3[@{ill& Tut]MNf&;{$
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10 ljiis^tMii@$[li5iifs@)l@[jsiiblis}Tii@QsWFs}tsiWEi@iDNiffiliiji 11 12
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13 6.7 Acceleration Sensors 14 4.7,1 The dynamic range should be 1000:1 zero to peak, or greater; 15 for example, 0.00lg to 1.0g.
16 4,7,2 The frequency range should be 0.20 Hz to 50 Hz or an 17 equivalent demonstrated to be adequate by computational techniques applied to 18 the resultant accelerogram.
U 19 M
Recorder 20 4,8.1 The sample rate should be at least 200 sampies per second in 21 each of the three directions.
22 4,8,2 The bandwidth should be at least from 0.20 Hz to 50 Hz.
23 4e8.3 The dynamic range should be 1000:1 or greater and be able to 24 record at least 1.0g 0 to peak.
25 a.9 Seismic Trigger. The actuating level should be adjustable and 26 within the range of 0.00lg to 0.02.
9 27 5.
INSTRUMENTATION INSTALLATION 7
1 M
The instrumentation should be designed and installed so that the 2
mounting is rigid.
3 M
The instrumentation should be oriented so that the horizontal axes 4
are parallel to the orthogonal horizontal axes assumed in the seismic 5
analysis.
6 M
Protection against accidental impacts shauld be provided.
7 6.
INSTRUMENTATION ACTUATION 8
f_d Both vertical and horizontal input vibratory ground motion should 9
actuate the same time-history accelerograph. One or more seismic triggers may 10 be used to accomplish this.
11 52 Spurious triggering should be avoided.
12 M
The seismic trigger mechanisms of the time-history accelerograph 13 should be set for a threshold ground acceleration of not more than 0.02g.
14 7.
REMOTE INDICATION 15 Activation of the free-field or any foundation-level time-history 16 accelerograph should be anr.anciated in the control room.
If there is more 17 than one control room at the site, annunciation should be provided to each 18 control room.
19 8.
MAINTENANCE 20 M
The purpose of the maintenance program is to ensure that the 21 equipment will perform as required. As stated in Regulatory Position 3, the 22 maintenance and repair procedures should provide for keeping the maximum 23 number of instruments in service during plant operation and shutdown.
24 M
Systems are to be given channel checks every 2 weeks for the first 25 3 months of service after startup. Failures of devices normally occur during 25 initial operation. After the initial 3-n.onth period and 3 consecutive 8
-. -.. - =. -. _ - -. -.. -.. -.
1 I
successful checks, monthly channel checks are sufficient. The monthlv channel i
2 check is to include checking the batteries. The channel functicnal test
]
3 should be perfomed every 6 months. Channel calibration should be performed
]
4 during refueling.
i l
5 D.
IMPLEMENTATION i
6 The purpose of this section is to provide guidance to applicants and j
7 licensees regarding the NRC staff's plans for using this regulatory guide.
8 Thi: pn; n:d n;i:i : 5:: k n :1::::d t: ;;::en ;; ; bli 9
p eticip:ti : in it: d = i ; xt.
Except in those cases in which the i
10 applicant prcposes an acceptable alternative method for complying with the i
11 specified portions of the Commission's regulations, the method i: be described l
f in th: ::tiv: gi guide nf1 : tin; ;;ili: : ;.xt: =ill be used in the 12' 13 evaluation of applications for construction permits, operating licenses, 14 combined licem as, or' design certification submitted after th: ig h x;t: tic:
15 d:t: t: b; :p::ift:d in th: ::tiv: ;;id: QFEQ I E M E ( 111 R E M alCE.
16 This guide ruld $1Hnot be used in the evaluation of an application for an 17 operating license submitted after th: ig h nat:tica d:t: t: 5: :p::ified in 18 th: ::tiv: ;;id: a_rd_i!_WEID_AT_EI_C_ET_E_TFI_INE_R_ILE.alif the construction permit 19 was issued prior to that date.
20 EINN!Mf5T(Is%jMQssTei2r Mi[e]s JesiiNWIIMMEM 21 E_F_FM_ 19E'~94Kf9_1G110EEf!IN_E, NI.Lihiiii[p?iis_ lib _isiRY~Tf(liiisiffiliiis~as~tSiiili L-
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APPENDIX f
i 2-DEFINITIONS 1
l-3 Acceleration Sensor. An instrument capable of sensing absolute acceleration i
4 and transmitting the data to a recorder.
f I
l.
5'
' Accessible Instr m.ts.
Instruments or sensors whose locations pemit ready i
I 6
access during plant operation without' violation of applicable safety 7
regulations, such as-0ccupational Safety and Health Administration (OSHA), or f
[
j 8
regulations dealing with plant security or radiation protection safety.
)
k-j 9
Channel Calibration (Primary Calibration). The determination and, if l
e i
l 10 required, adjustment of an instrument, sensor, or system such that it responds 9
1 i
11 within a specific range and accuracy to an acceleration, velocity, or 12 displacement input, as applicable, or responds to an acceptable physical 13 constant.
- t 14 Channel Check. The qualitative verification of the functional status of the 15 instrument sensor. This check is an 'in-situ" test and may be the same as a 16 channel functional test.
17 Channel Functional Test (Secondary Calibration). The determination without 18 adjustment that an instrument, sensor, or system responds to a known input of 19 such character that it w 11 verify the instrument, sensor, or system is 20 functioning in a manner that can be calibrated.
El Containment - See Primary Containment and Secondary Containment.
22 Nonaccessible Instruments.
Instruments or sensors in a location that does not 23 pemit ready access during plant operation because of a risk of violating 24 applicable plant operating safety regulations, such as OSHA, or regulations 25 dealing with plant security or radiation protection safety.
26 Operatina Basis Earthouake Ground Motion (OBE). The vibratory ground motion 27 for which those features of the nuclear power plant necessary for continued 28 operation without undue risk to the health and safety of the public will 29 remain functional. The value of the OBE is set by the applicant.
10
1 Primary Containment. The principal structure of a unit that acts as the 2
barrier, after the fuel cladding and reactor pressure boundary, to control the 3
release of radioactive material. The primary containment includes (1) the 4
containment structure and its access openings, penetrations, and appurte-5 nances, (2) the valves, pipes, closed systems, and other components used to 6
isolate the containment atmosphere from the environment, and (3) those systems 7
or portions of systems that, by their system functions, extend the containment 8
structure boundary (e.g., the connecting steam and feedwater piping) and 9
provide effective isolation.
10 Recorder. An instrument capable of simultaneously recording the data versus 11 time from an acceleration sensor or sensors.
12 Secondary Containmeqt. The structure surrounding the primary containment that 13 acts as a further barrier to control the release of radioactive material.
14 Seismic Isolator. A device (for instance, laminated elastomer and steel) 15 installed between the structure and its foundation to reduce the acceleration 16 of the isolated structure, as well as the attached equipment and components.
17 Seismic Triacer. A device that starts the time-history accelerograph.
18 Time-History Accelerocraoh. An instrument capable of sensing and permanently 19 recording the absolu'.s acceleration versus time. The components of the time-20 history accelerograph (acceleration sensor, recorder, seismic trigger) may be 21 assembled in a self-contained unit or may be separately located.
22 Triaxial. Describes the function of an instrument or group of instruments in 23 three mutually orthogonal directions, one of which is vertical.
11 l
I 1
REGULATORY ANALYSIS 2
A separate regulatory analysis was not prepared for this regulatory 3
guide. The dr:ft regulatory analysis, "."r;;;;;d Revision of 10 CFR Part 100 4
and 10 CFR Part 50," was prepared for the pr:p:: d amendments, and it provides 5
the regulatory basis for this guide and examines the costs and benefits of the i
6 rule as implemented by the guide. A copy of the dr:ft regulatory analysis is 7
available for inspection and copying for a fee at the NRC Public Document 8
Room, 2120 L Street NW. (Lower Level), Washington, DC, as Er.:1::gr: 2 t:
9 S::y ?t I?t
- f.
}
1 12
w
-4 1
REGULATORY GUIDE 1.166 2
(Draft was DG-1034) 3 PRE-EARTHQUAKE PLANNING AND Ilt9EDIATE NUCLEAR POWER 4
PLANT OPERATOR POSTEARTHQUAKE ACTIONS 5
A.
INTRODUCTION 6
Paragraph IV(a)(4) of In;ned Appendix S " Earthquake Engineering 7
Criteria for Nuclear Power Plants," to 10 CFR Part 50, " Domestic Licensing of 8
ProductionandUtilizationFacilities,"weeM-requirejthatsuMnleinstru-9 mentation' be provided so that the seismic response of nuclear power plant 10 features important to safety can be evaluated promptly.
Paragraph IV(a)(3) of 11 Pn;n d Appendix S to 10 CFR Part 50 ;;;1d requirej shutdown of the nuclear 12-power plant if vibratory ground motion exceeding that of the operating basis 13 earthquake ground motion (OBE) or significant plant damage occurs.
If 14 systems, structures, or components necessary for the safe shutdown of the 15 nuclear power plant ;xid y not be-available after occurrence of the OBE, 16 the licensee crld be n;;ind t: 3 consult with the NRC and @] propose a 17 plan for the timely, safe shutdown of the nuclear power plant. Peoposed 18 Paragraph 50.54(ff) to 10 CFR Part 50 = 1d requirej licensees of nuclear 19 power plants that have adopted the earthquake engineering criteria in Peepesed 20 Appendix S to 10 CFR Part 50 to shut down the plant if the criteria in Para-21 graph IV(a)(3) of "n; ned Appendix S are exceeded.
22 This guide i: 5-ing d:=1:;;d t providej guidance acceptable to the 23 NRC staff for a timely evaluation after an earthquake of the recorded 24 instrumentation data ano ior determining whether plant shutdown would be 25 required by th: pn ; n ed.n nd= t: t: 10 CFR Part 50.
26
":;;ht ry guida :r: i n ;;d t: d u rib: =d ch: : =i hb h h the
'27 ym see :h inf: :ti= = =th:d; ne:pt:th to the "".C :t:ff f:r i :; hunt 28 4my v. ific p:rt: Of th: C:rr.i n i = ': n;;hti=:, itchni; n ned by the 29
- t:ff b :ninting :pnific pntkn Or pntchied :: id;;b, =d ;;id=:: 1:
30
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" ;;hury ;;ida :r: ::t ::i:tituts hr n;;hti:n, =d 31
- -;1i=n zith n;;hur; ;;id= h =t n;;ind. ":g; h e r; ;;id u :r:
32 innd 5 d=ft h= f= ;;ilic :rr::t t h=h: th: ;;blic h th: : =1; 33 0 id=n h 5:in; d:=h;;d S Onft " ;;hury Cuid: 00 1033, th: Thid 2
34 "n;ned "=hin 2 i: Regulatory Guide 1.12 " Nuclear Power Plant Instru-35 mentation for Earthquakes," RiiHsW2Q4+-describei seismic instrumentation 36 acceptable to the NRC staff.~~
1
- t;;;; Of d::: hpia; th: 7;;;ht:ry p::iti =:.
Dr:ft r;;;htry ;;id:: h::
2
=t n::ixd ::;-ht: :t:ff rc;ic: =d d: ::t r:pr;;;;t Offici:1
""0 :t:ff 3
p=it in:.
4 Any information collection activities mentioned in this dn ft regulatory 5
guide are contained as requirements in th: pn; n:d x : ' t: t: 10 CFR Part 6
50 tht crld pr::id:,1Tg[WM]the regulatory basis for this guide.
7 The pn ; x:d x t:-t: h = h = & itted t: ] g ( g g g fg
]
8 M[ffg838 MIEMg}ggagiggMthe Office of Management 9
and Budget fr :1:rr= tht n; b; :;;n;rict: xdr th: ":p:=rt "-d=ti=
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Erh chr:r, if citind, ;xid h: ;;1y t n; thmtin 11
- "::ti= nthitic: nati=:d h thh ;;id:QMije3]gf!!IQM.
12 B.
DISCUSSION 13 When an earthquake occurs, ground motion data are recorded by the 14 seismic instrumentation.* These data are used to make a rapid determination 15 of the degree of severity of the seismic event. The data from the ( M Q Q 16 seismic instrumentation, coupled with information obtained from a plant 17 walkdown, are used to make the initial determination of whether the plant must 18 be shut down, if it has not already been shut down by operational 19 perturbations resulting from the seismic event.
If on the basis of these 20 initial evaluations (instrumentation data and walkdown) it is concluded that 21 the plant shutdown criteria have not been exceeded, it is presumed that the 22 p1 ant will not be sNt downi@[@Qiji@lfsMipjgg@{@R(1sif@@]fj 23 MIWlR$ffj]Miiiiiiiiilsf[tWKOM.
Guidance S b:h; d =h;;d 24 on postshutdown inspections and plant restartt EE5jOl@3E::: Onft 25 Regulatory Guide 001035, ll14[,3" Restart of a Nuclear Power Plant Shut Down 26 by a Seismic Event." The Electric Power Research Institute has developed 27 guidelines that will enable licensees to quickly identify and assess 28 earthquake effects on nuclear power plants. These guidelines are in EPRI NP-29 5930, "A Criterion for Determining Exceedance of the Operating Basis 30 Earthquake," July 1988'; EPRI NP-6695, " Guidelines for Nuclear Plant Response 31 to an Earthquake," December 1989'; and EPRI TR-100082, " Standardization of 32 Cumulative Absolute Velocity," December 1991.*
33
'EPRI reports may be obtained from the Electric Power Research Institute, 34 Research Reports Center, P.O. Box 50490, Palo Alto, CA 94303 2
m
-r-t
- -,i -
V
l l
1 This regulatory guide is based on the assumption that the nuclear power 2
plant has operable seismic instrumentation, including the equipment and soft-3 ware required to process the data withi.t 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after an errthquake. This is 4
necessary because the decision to shut down the plant will be made, in part, 5
by comparing the recorded data against OBE exceedance criteria The decision 6
to shut down the plant is also based on the results of the plant walkdown 7
inspections that take place within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of the event.
If the seismic 8
instrumentation or data processing equipment is inoperable, the guidelines in 9
Appendix A to this guide would be used to determine whether the OBE has been 10 exceeded.
11 MER(Mf@@ilisij]hif{sii@ lit]Kdif{KE@[i{MRijy 12 i))fdiiisMiijsijfatisiiq[QQujift]ishtit3[dsiisiifFitiiijgthi3Rc3[iffluist; 13 EffialHiW613siHilDaWesIisdissitsiWEEEfGEfiaEMIfdisHi12!
14
]hi'@fi@H@fiifpsisiliM[M}@jjsii)j{j@iSisl5fii]@[f[(QV{3hb5]d 15 EEilMif@[ifMQMtlil))js[ilTif[gi@]jiEh3@iijli((fEgeijfjjly 16 $st~psiiisfafl3EAftsCWijfthiidakiistithi]pliitgifi',{tSjfsijishiiQji@tfUii 17 lasdICAgils5Gl[R{iMiljfidiss EthiffiiiiiissDi@tishdith(Hisi ti3&iiE~ arid 18 Gilt @ % l if fsl 3 Af @ l Q i 3 t 3 @ j [in}} 19 Because earthquake-induced vibration of the reactor vessel could lead to 20 changes in neutron fluxes, a prompt check of the neutron flux monitoring 21 sensors would provide an indication that the reactor is stable. 22 Shutdown of the nuclear power plant would be required if the vibratory 23 ground motion experienced exceeds that of the OBE. Tw; criteri: @ @fi[}66 24 fo. determining exceedance of the OBE ", based or, data recorded in the free-25 field) are provided in EPRI NP-5930: a threshold response spectrum ordinate 26 criterica EhsEQand a cumulative absolute velocity (CAV) criterica Mihil. 27 Seismic Category I structures at the nuclear power plant site may be designed 28 using different ground motion response spectra; for example, one used for the 29 certified standard design and another for site-specific applications. The 30 spectrum ordinate criterion is based on the lowest spectrum used in the design l 31 of the Seismic Category I structures. A procedure to standardize the 32 calculation of the CAV is provided in EPRI TR-100082. A spectral velocity ~ 33 threshold has also been recommended by EPRI since some structures have 34 fundamental frequencies below the range specified in EPRI NP-5930. The NRC 35 staff now recommends 1.0 to 2.0 Hz for the range of the spectral velocity 36 limit since some structuret have fundamental frequencies below 1.5 Hz. The 37 former range was 1.5 to 2.0 Hz. j 1 3
1 l 1 Since the containment isolation valves may have malfunctioned during an 2 earthquake, inspection of the containment isolation system is necessary to 3 ensure continued containment integrity. 4 The NRC staff does not endorse the philosophy discussed in EPRI NP-6695, 5 Section 4.3.4 (first paragraph, last sentence), pertaining to plant shutdown 6 considerations following an earthquake based on the need for continued power 7 generation in the region. If the licensee determines that plant shutdown is 8 required by the NRC's regulations, but the licensee does not consider it 9 prudent to do so, the li ensee would be required to consult with the NRC and 10 propose a plan for the timely, safe shutdown of the nuclear power plant. 11 Appendix B to this guide provides definitions to be used with this 12 guidance. 13 Sold:r; cf :n :per: ting licen:: cr con:truction permit i::ued prior t; 14 the i:pl;; nt: tion d:t to bc :p ified in th: ::tive guid: ::y volunt rily 4cf ;;;nt th: ::th:d: 10 be de: crit:d in th ::tive guid :nd 'h
- thed:
l 15 16 being d;;;1:p:d in Dr:ft Regulatcry Cuid:: DC !033, "N; le:r Power Pl nt 17 In:trument:tien for E:rthq :ke:," :nd DC 1035, "R :t:rt of ; Nucle:r Power 18 Pl:nt Shut 0;un by : Sci;;ic Event."- 19 C. REGULATORY POSITION 20 1. BASE-LINE DATA 21 i.1 Information Related to Seismic Instrumentation 22 A file containing information on all the seismic instrumentation should 23 be kept at the plant. The file should include: 24 1. Information on each instrument type such as make, model, and 25 serial number; manufacturers' data sheet; list of special matures or options; 1 l 26 performance characteristics; examples of typical instrumentation readings and 27 interpretations; operations and maintenance manuals; repair procedures (manu-28 facturers' recommendations for repairing common problems); and a list of any l 29 special requirements, e.g., maintenance, operational, installation. l l 4 t 1
i 1 2. Plan views and vertical sections showing the location of each 2 seismic instrument and the orientation of the instrument axis with respect to l 3 a plant reference axis. 4 3. A complete servica history of each seismic instrument. The 5 service history should include information such as dates of servicing, j 6 description of completed work, and calibration records and data (where 7 appl i cabl e).[QKeldiisjyitiKjMsisil]@'snjish](@Qits]s[IOREe] t ~ijinTalisi@]lMK@liffidilliifMiig5sTfaiis[Q 8 c 9 4. A suitable earthquake time-history (e.g., the October 1987 I 10 Whittier, California, earthquake) or manufacture's calibration standard and 11 the corresponding response spectrum ana cumulative absolute velocity (CAV) 12 (seeRegulatoryPositioni4DlkisfM2). The resp n:: :pectru: : d CAY cheuld 13 bc ::lculated :fter M{sHthe initial installation and each servicing of the ) 14 free-field instrumentation the response spectrum and CAV should be co'culated j ssd))ll~Qssi3sislits$lP6DM6G5 15 e 16 1.2 Plannino for Postearthauake Inspections i 17 The @MgrtisilisEsMM36sBhit3i$hs{seleetion of equipment and 18 structuresforinspections)andthecontentofthebaselineinspectionsas ) 19 described in Sections 5.3.1 :nd 5.3.2.1 of EPRI NP-6695, " Guidelines for 20 Nuclear Plant Response to an Earthquake," are acceptable to the NRC staff for 21 satisfying the prop : d requirements in Paragraph IV(a)(3) of Prop;;;d 22 Appendix S to 10 CFR Part 50 for ensuring the safety of nuclear power plants. 23 2. IMMEDIATE POSTEARTH00AKE ACTIONS l 24 The guidelines for immediate postearthquake actions specified in 25 Sections 4.3.1 (with the exception specified below) and 4.3.2 (including 26 Section 5.3.2.1 :nd ite:: 7 :nd S cf T:ble 5 I) of EPRI NP-6695 are acceptable 27 to the NRC staff for satisfying the requirements prope: d in Paragraph 28 IV(a)(3) of Pr p:::d Appendix S to 10 CFR Part 50. 29 In Section 4.3.1, a check of the neutron flux monitoring sensors for 30 changes should be added to the specific control room board checks. 5
l 1 3. EVALUATION OF GROUND MOTION RECORDS t 2 3.1 Data Identification 3 A record collection log should be maintained at the plant, and all data 4 should be identifiable and traceable with respect to: f r 5 1. The date and time of collection, 1 6 2. The make, model, serial number, location, and orientation of the 7 instrument (sensor) from which the record was collected. 8 3.2 Data Collection I I 9 3.2.1 Only personnel trained in the operation of the instrument should l 10 collect the data. 11 3.2.2 The steps for removing and storing records from each seismic 12 instrument thould be planned and performed in accordance with established 13 procedures. 14 3.2.3 Extreme caution should be exercised to prevent accidental damage 15 to the recording media and instruments during data collection and subsequent 16 handling. 17 3.2.4 As data are collected and the instrumentation is inspected, notes 18 should be made regarding the condition of the instrument and its installation, 19 for example,, instrument flooded, mounting surface tilted, fallen objects that 20 struck the instrument or the instrument mounting surface. 21 3.2,5 For validation of the collected data, the information described 22 in Regulatory Position 1.l(4) should be added to the record without affecting 23 the previously recorded data. 24 3.2,6 If the instrument's operation appears to have been normal, the 25 instrument should remain in service without readjustment or change that would l 26 defeat attempts to obtain postevent calibration. 6
. ~. -. ~.. - -...- -. i l l 1 3.3 Record Evaluat18 9 I f l 2 Records should be analyzed according to the manufacturer's specift:a-3 tions and the results of the analysis should be evaluated. Any record 4 anomalies, invalid data, and nonpertinent signals' should be noted, along with 5 any known causes. L 6 4. DETERMINING OBE EXCEEDANCE
- 7 The esaluation to determine whether the OBE was exceeded should be 8
performed using data obtained from the three components of the free-field 9 ground motiori (i.e., two horizontal and one vertical). The evaluation may be 10 performed on uncorrected earttquake records. It was found in a study of 11 uncorrected versus corrected earthquake records (see EPRI NP-5930) that the 12 use of uncorrected records is conservative. The evaluation should consist of a check of the response spectrua -sij(CAV li it, and the operability of the 33 r 14 instrumentation. This evaluation should take place within 4 hours of the i 15 ' earthquake. 16 4.1 Response Spectrum Check l 17 4.1.1 18 The OBE response spectrum check is performed using the lover of: 19 1. The spectrum used in the certified standard design, or 20 2. A spectrum other than (1) used in the design of any Seismic 21 Category I structure. 22 4.1.2 23 The OBE response spectrum is exceeded if any one of the three components 24 (two horizontal and one vertical) of the 5 percent damped free-field ground 25 motion response spectra is larger than: 7
~ l 1. The corresponding design response spectral acceleration (OBE 2 spectrum if used, otherwise 1/3 of the safe shutdown earthquake 3 (SSE) spectrum) or 0.29, whichever is greater, for frequencies 4 between 2 to 10 Hz, or 5 2. The corresponding design response spectral velocity (OBE spectrum 6 if used, otherwise 1/3 of the SSE spectrum) or a spectral velocity 7 of 6 inches per second (15.24 centimeters per second), whichever 8 is greater, for frequencies between I and 2 Hz. 9 4.2 Cumulative Absolute Velocity (CAV) Li it "E 10 For each component of the free-field ground motion, the CAV should be 11 calculated as follows: (1) the absolute acceleration (g units) time-history l 12 is divided into 1-second intervals, (2) each 1-second interval that has at 13 least I exceedance of 0.025g is integrated over time, (3) all the integrated 14 values are summed together to arrive at the CAV. The CAV limit M i@is S 15 exceeded if any CAV calculation is greater than 0.16 g-second. Additional 16 information on how to determine the CAV.is provided in EPRI TR-100082. I I 17 4.3 Instrument Operability Check 18 After an earthquake at the plant site, the response spectrum and CAV 19 should be calculated using gNslisislijijiii@3hipyisegth: ::libr:ti:n e 20
- t: d:rd (::: Regulatory Position 1.l(4)t siW3@MiiGl{ilsbiiEQj)ijlifT a~rlid 21 iiij}IDRjh]e Q@3 sis to demonstrate that the time-history analysis 22 hardware and software were functioning properly. { @ M y$1]s3 Qtjjj 23
@QMibEl(6sHiips@@{$1tQilflC) 24 4.4 Inonerable Instrumentation or Data Processina Hardware or Software 25 If the response spectrum and the CAV (Regulatory Positions 4.1 and 4.2) i 26 can not be obtained because the seismic instrumentation is inoperable, data 27 from the instrumentation are destroyed, or the data processing hardware or j 28 software is inoperable, the criteria in Appendix A to this guide should be 29 used to determine whether the OBE has been exceeded. i 8 Iu -
1 5. CRITERIA FOR PLANT SHUTDOWN f 4 2 If the OBE is exceeded or significant plant damage occurs, the plant 3 must be shut down unless a plan for the timely, safe shutdown of the nuclear 4 power plant has been proposed by the licensee and accepted by the NRC staff. 5 5.1 OBE Exceedance 1 6 IftheresponsespectrumcheckandtheCAVlimit(([M(performedor i I 7 calculated in accordance with Regulatory Positions 4.1 and 4.2) were exceeded, 8 the OBE was exceeded and plant shutdown is required. If either li=it j@ j 9 does not exceed the criterion, the earthquake motion did not exceed the OBE. 10 If only one li=it jidMcan be ch::ked Msfiiiiid, the other li=it j@]is 11 assumed to be exceeded (({Q@[c6icHsii[Q@6NiisfM[Jt@}My 12 yisjjjlMB. The determination of whether or not the OBE has been exceeded 13 should be performed even if the plant automatically trips off-line as a result 14 of the earthquake. 15 5.2 Damaae 16 The plant should be shut down if the walkdown inspections performed in 17 accordance with Regulatory Position 2 discover damage. This evaluation should 18 take place within 8 hours of the earthquake occurrence. 19 5.3 Continued Operation 20 If the OBE was not exceeded and the walkdown inspection indicates no j 21 damage to the nuclear power plant, shutdown of the plant is not required. The 22 plant may continue to operate (or restart following a post-trip review, if it 23 tripped off-line because of the earthquake). 24 6. PRE-SHUTDOWN INSPECTIONS 25 The pre-shutdown inspections described in Section 4.3.4 (inchding all 26
- 2;ccti:n:) of EPRI NP-6695, " Guidelines for Nuclear Plant Response to an 27 Earthquake," with the exceptions specified below are acceptable to the NRC 28 staff for satisfying the requirements pr p:::d in Paragraph IV(a)(3) of 9
i l l l 1 Pr:;;; d Appendix S to 10 CFR Part 50 for ensuring the safety of nuclear power 2 plants. 3 6.1 Shutdown Timina 1 4 Delete the last senten:e in the first paragraph of Section 4.3.4. 5 6.2 Safe Shutdown Eautoment 6 In Section 4.3.4.1, a check of the containment isolation system should l 7 be added to the minimum list of equipment to be inspected. 8 6.3 Orderly Plant Shutdown i 9 The following paragraph in Section 4.3.4 of EPRI NP-6695 is printed here 10 to emphasize that the plant should shut down in an orderly manner. 11 " Prior to initiating plant shutdown following an earthquake, l 12 visual inspections and co,ntrol board checks of safe shutdown i 13 systems should be performed by plant operations personnel, and the 14 availability of off-site and emergency power sources should be f 15 determined. The purpose of these inspections is to determine the 16 effect of the earthquake on essential safe shutdown equipment 17 which is not normally in use during power operation so that any 18 resets or repairs required as a result of the earthquake can be l 19 performed, or alternate equipment can be readied, prior to 20 initiating shutdown activities. In order to ascertain possible 21 fuel and reactor internal damage, the following checks should be 22 made, if possible, before plant shutdown is initiated.... " l 23 D. IMPLEMENTATION 24 The purpose of this section is to provide guidance to applicants and 25 licensees regarding the NRC staff's plans for using this regulatory guide. 26 Thi: pr:pned r=i ica hn been relened t en ::r:g p blic 27 p rticip:ti n in it; d=chinnt. Except in those cases in which the 28 applicant proposes an acceptable alternative method for complying with the 1 10 i
I specified portions of the Commission's regulations, the method t: : described 2 in the niiv Ms} guide reficcting public mat: will be used in the 3 evaluation of applications for construction permits, operating licenses, 4 combined licenses, or design certification submitted after the i :pl =nt:tc 5 d:t te bc :pn ified in th: =tiveguid:EFFECTIVEIDATEIDFiTHEIFINALTREE. .v sA,vua....w.- w.v -v e.au.w. w.,V.-c ~.< .w~ e... w.wm o -- w e,.*swe..--- l 6 This guide c=1d iiDi?not be used in the evaluation of an application for an l MwMa4& 7 operating license submitted after the i=pl:=ntati:n d:tc i be spuf fied in 8 th: =tiv: guid: E_FFE.,q~W~EIWYE_%Q. _HE_FI.NA.v.CIR. R.._Egif the construction permit i 9 was issued prior to that date. i s suea~um<v~ ^w.>,;-y 4p_rio. mat _o
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1 APPENDIX A 2 INTERIM OPERATING BASIS EARTHQUAKE EXCEEDANCE GUIDELINES I 3 This regulatory guide is based on the assumption that the nuclear power 4 plant has operable seismic instrumentation and equipment (hardware and 5 software) to process the data. If the seismic instrumentation or data 6 processing equipment is inoperable, the following should be used to determine 7 whether the operating basis earthquake ground motion (OBE) has been exceeded: 8 1. For plants at which instrumentally determined data are available only 9 from an instrument installed on a foudation, tha cumulative absolute 10 velocity (CAV) li=it )$siik}(see Regulatory Position 4.2 of this guide) 11 is not applicable. In this case, the determination of OBE exceedance is 12 based on a response spectrum check similar to that described in 13 Regulatory Position 4.1 of this regulatory guide. A comparison is made 14 between the foundatien-level design response spectra and data obtained 15 from the foundation-level instruments. If the response spectrum check 16 at any foundation is exceeded, the OBE is exceeded and the plant must be 17 shut down. At this instrument location it is inappropriate to use the 18 0.2g spectral acceleration limit or the 6 inches per second (15.24 19 centimeters per second) spectral velocity limit stated in Regulatory 20 Position 4.1.2. l 21 2. For plants at which no free-field or foundatior.-1c.el instrumental data 22 are available, 2t the data processing equipment is inoperable and the 23 responsespectrumcheckandtheCAVlimit56sik[cannotbedetermined 24 (Regulatory Positions 4.1 and 4.2), the OBE will be considered to have 25 been exceeded and the plant must be shut down if one of the following 26 applies: 27 1. The earthquake resulted in Modified Mercalli Intensity (MMI) VI or 28 greater within 5 km of the plant, i 29 2. The earthquake was felt within the plant and was of magnitude 6.0 30 or greater, or l 31 A-1
~ i i e l 1 3. The earthquake was of magnitude 5.0 or greater and occurred within i j 2 200 km of the plant. l \\ 3 A postearthquake plant walkdown should be conducted (see Regulatory l 4 Position 2 of this guide). 5 If plant shutdown is warranted under the above guidelines, the plant 6 should be shut down in an orderly manner (see Regulatory Position 6 of this 7 guide). i 1 8 Hole: The determinations of epicentral location, magnitude, and j 9 intensity by the U.S. Geological Survey, National Earthquake 10 Information Center, will usually take precedence over other estimates; 11 however, regional and local determinations will be used if they are I 12 considered to be more accurate. Also, higher quality damage reports or 13 a lack of damage reports from the nuclear power plant site or its 14 immediate vicinity will take precedence over more distant reports. I I ) I A-2
- - -..-. ~.~.- l 1 APPENDIX B 2 DEFINITIONS r 3 Certified Standard Desion. A Commission approval, issued pursuant to Subpart j 4 B of 10 CFR Part 52, of a standard design for a nuclear power facility. I 5 Desian Response Spectra. Response spectra used to design Seismic Category I j 6 structures, systems, and components. ) l l 7 Doeratina Basis Earthauake Ground Motion (OBE). The vibratory ground motion 8 for which those features of the nuclear power plant necessary for continued ] 9 operation without undue risk to the health and safety of the public will { 10 remain functional. The value of the OBE.is set by the applicant. I 11 Spectral Acceleration. The acceleration response of a linear oscillator with 12 prescribed frequency and damping. 13 SDectral Velocity. The velocity response of a linear oscillator with pre-14 scribed frequency and damping. 1 4 i 1 B-1
l l 1 REGULATORY ANALYSIS 2 A separate regulatory analysis was not prepared for this regulatory 3 guide. The dr:ft regulatory analysis, "Pr;;;;;d Revisions of 10 CFR Part 100 i i 4 and 10 CFR Part 50," was prepared for the pr:;:::d amendments, and it provides l I i 5 the regulatory basis for this guide and examines the 1sts and benefits of the 6 rule as implemented by the guide. A copy of the dr:ft regulatory analysis is 7 available for inspection and copying for a fee at the NRC Public Document 8 Room, 2120 L Street NW. (Lower Level), Washington, DC, as Eccle: r: 2 to 9 S y Si 104 (({G$. l l l RA-1
i 1 REGULATORY GUIDE 1.167 2 (Draft was DG-1035) 3 RESTART OF A NUCLEAR POWER PLANT 4 SHUT DOWN BY A SEISMIC EVENT 5 A. INTRODUCTION i i 1 6 Paragraph IV(a)(3) of.ar:p::cd Appendix S, " Earthquake Engineering l 7 Criteria for Nuclear Power Plants," to 10 CFR Part 50, " Domestic Licensing of l 8 Production and Utilization Facilities," ;;uld requirej shutdown of the nuclear 9 power plant if vibratory ground motion exceeding that of the operating basis l 10 earthquake ground motion (OBE) occurs or if significant plant damage occurs.' 11 Prior to resuming operations, the licensee must demonstrate to the NRC that no 12 functional damage has occurred to those features necessary for continued 13 operation without undue risk to the health and safety of the public. 14 This guide i: being devel:p:d 10 providej guidance acceptable to the NRC i 15 staff for performing inspections and tests of nuclear power plant equipment 16 and structures prior to restart of a plant that has been shut down by a 17 seismic event. 1 18 Regulatory guid:: Or i::u-d 10 describe :nd ::ke : vail:ble t; the j l 19 public uch inf;rmation :: =cthed: Occept:ble t; the "RC st:ff for 20 impl :: ting :pecific p:rt: Of the C:::i :icn': regul: tion:, techr.iqu:: :cd 21 by-tb :t:ff in ev;lu ting :pecific pr blem: er p :tel ted accident:, :nd 22 guid:nce t: :pplic;nt:. Regulatory guide: :r not :ab:titute: f;r 23 regul: tion:, : d :::pli:nce with regul:tery guide; i: n t required. 24 Regul:t:ry guide: cre i::ued in dr:ft form f r public cc:::nt t; inv;1ve the 25 public in th ::rly :t:g : Of developing the regul:tery p :ition:. Dr:ft 26 regulatory guide: h;v n-t received cc pletc :t:ff r^ view :nd do n:t r^prc:ent 27
- r. e. s.. z.,. une... c. c....s. z....
-~... r.. .m... 28 Any information collection activities mentioned in this draft regulatory 29 guide are contained as requirements in th^ prop;;cd ::endment: t 10 CFR Part i 30 50 th:t w:uld pr: vide FM_iiicF..p_F6H._di_s, lithe regulatory basis for this guide. m. 31 The pr:p::^d : end: nt: h:vebeen:ab-ittedi;Jif@%itl56))31}ieQ66 i.,_. t. d,n,1. 0, ,KPa...r_t,t S0 haive:b_een_,;a_p.__ds.bya;the Office of Management .. FRC. - -.- 32 r~equ..re. men.s i p ve ..r.o..~ .-.- ~ _ 33 'Cuid:nce is being devel:p-d in Draft Regulatory Guide DC 1031 }{l66, 34 " Pre-Earthquake Planning and immediate Nuclear Power Plant Operator i l 35 Postearthquake Actions," te-provide) criteria for plant shutdown.
l 6 i i 1 and Budget f r char =:: th;t =y 5 :ppr:prict: = der the P: pent:rk Red;; tion 2 Act. &ch k r= 0, if Obt:ined, :: ld :h :pply t ny inf = t;: 3 0lketi= nticitie: =nti=:d in thi: guid:E M ii M 3 15 (00M. 4 B. DISCUSSION 5 Data from seismic instrumentation
- and a walkdown of the nuclear power
[ 6 plant are used to make the initial determination of whether the plant must be l 7 shut down after an earthquake, if the plant has not already shut down from 8 operational perturbations resulting from the seismic event.' 9 The Electric Power Research Institute has developed guidelines that will 10 enable licensees to quickly identify and assess earthquake effects on nuclear 11 power plants in EPRI NP-6695, " Guidelines for Nuclear Plant Response to an 12 Earthquake,"' December 1989. This regulatory guide addresses sections of 13 EPRI NP-6695 that relate to pos'.nutdown inspection and tests, inspection 14 criteria, inspection personnel, documentation, and long-term evaluations. 15 EPRI NP-6695 has been supplemented to add inspections and tests as a 16 basis for acceptance of stresses in excess of Service Level C and to recommend i 17 that engineering evaluations of components with calculated stresses in excess l 18 of service Level D focus on areas of high stress and include fatigue analyses. 19 "0lder: Of = cper: ting li =:: cr :=:tr=ti= pt:-it i==d pri;r to 20 the imph=ntati= d:tc to bc :pecified in the utie: guid =y vehntarily 21 i :pk=nt th: =thod: to be de:Cribed in th: =tiv: juid =d the =thod 22 being devchped in Or:ft Reg;htery Cuide: DC 1033, "Sch r P =r Pl=t 23 !=tr;=nt:ti= for E:rthq=kes," =d DC 1034, "Pr: E:rthq=k: Ph=ing =d i 24 !=:di;t ich;r P:=r Pl=t Oper tcr P :tarthq=k: Acti=." 25 C. REGULATORY POSITION l 26 After a plant has been shut down by an earthquake, the guidelines for 27 inspections and tests of nuclear power plant equipment and structures that are 28
- Cuid=ce i; being devchp:d in Dr:ft Regulatory Guide DC 1033 132, % e Cuide 1.12, " Nuclear Power Plant third Pr:p:: d Revist= 2 to Reguhtcr#"T66i27 that will describes seismic 29 Instrumentation for Earthquakes," Riifis 30 31 instrumentation acceptable to the NRCsfaff.
32 'EPRI reports may be obtained from the Electric Power Research Institute, 33 Research Reports Center, P.O. Box 50490, Palo Alto, CA 94303. 2
\\ ~ l 1 l l l l 1 depicted in EPRI NP-6695 in Figure 3-2 and specified in Sections 5.3.2 5.3.3 s. _3..2..... v. u, _.. e. i s,, and 5.3.4.,
- s. _,.2,__-...-, v. u....e.
i.,s, n... n __2 e 2 3 the documentation to be submitted to the NRC specified in Section in 5.3.5; 4 and the long-term evaluations that are specified in Section 6.3 (:11 ::: tion; i 5
- nd ::b:::ti:n:), with the exceptions specified below, ;;;1d be sfs] acceptable 6
to the NRC staff for satisfying the requirentnts pr:p:: d in Paragraph 7 IV(a)(3) of th: "r:p:: d Appendix S to 10 CFR Part 50. 8 1. EXCEPTIONS TO SECTION 6.3.4.10F EPRI NP-6695 1 l l 9 1.1 Item (1) should read: 10 If the calculated stresses from the actual seismic loading conditions 11 are less than the allowables for emergency conditions (e.g., ASME Code 12 Level C Service Limits or equivalent) or original design bases, the item 13 is considered acceptable, provided the results of inspections and tests j 14 (Section 5.3.2) show no damage. l 15 122 The second dashed statement of Item (3) should read: 16 -- An engineering evaluation of the effects of the calculated stresses 17 on the functionality of the item. This evaluation should address all 18 locations where stresses exceed faulted allowables and should include fatigue analysis _ rfASME_,. Code:Cl _ili_ comp _one_n._i._sean%_sy_ stems. w n-_ - _s _ss 19 a 20 111 The 1 :t p:r:gr:ph :h eld re:d: 21 "::n:ly:i: Of ::fety rel:ted piping cy:te : is not n id r:d n :::::ry 22 unles; there i: Ob;;rved d:::ge to the piping yste :. Exp;rience h:: 23
- hern that piping :ystem; d ;igned t; the AS"E C;d cre not d:::ged by 24 inerti: 10:d; re;;1 ting fr : :n ::rthqu:ke.
If d:::g: :: cur:, it will 25
- t likely : cur in the piping : pp;rt: cr :. d:::;: to the pip :t 26 fixed ::pp:rt: :::::d by rel:tiv: : pport displ:::::nt:. Th :- typ : cf 27 d:::;: ;;;1d be detected by the pl:nt ::lkd::: in:pection :nd pe:t 28
- h:td:rn in;;;; tion: d::cribed in Section: 4 :nd 5 cf this rep;rt.
In 29 gencr:1, piping re:::lysi: :h: ld be p rf r -d on : :::pling b :i t: 30
- rify th: :d:q ::y Of piping :nd 10 :::c;; the need for : pple: nt:1 31 nd :tructiv: ::::in: tion of potential high : train crc :.
32 2. LONG-TERM EVALUATIONS l 3 1
I l l f f ~ I i 1 Coincident with the long-term evaluations, the plant should be restored 2 to its current licensing basis. Exceptions to this must be approved by the j 3 Director, Office of Nuclear Reactor Regulation. l i 4 D. IMPLEMENTATION 5 The purpose of this section is to provide guidance to applicants and j 6 licensees regarding the NRC staff's plans for using this regulatory guide. l 1 7 Thi: dr:fi g id h:: 5::: r:le:::d t: ::::;r;;; ;;ili: p:rticip:ti: in l 8 it: d:::1:;-::t Except in those cases in which the applicant proposes an 9 acceptable alternative method for complying with the specified portfons of the 10 Commission's regulations, the method t be described in th: ::tive this guide 11 r:fl:: ting ;;bli ::-- ::t: will be used in the evaluation of applications for 12 construction permits, operating licenses, combined licenses, or design l 13 certification submitted after th: i pl;n nt:tica d:t: to bc :p::ified in the 14
- tive g;id: E_FF_E..C.T_IIT..._[04_TE_YW.-._IFI_NA~E_R.O..C_E.
This guide.::ald Sill.s?not be 15 used in the evaluation of an application for an operating license submitted 16 after the i pi:x nt:ti: d:t: t: b; ;;;;ified in th: ::tiv: ;;id: !FFitT he 17 n__nTife_FIMIFilinE_NKE_l.if the construction permit was issued prior to that m 18 date. ._,, iwonse ..n.- t.- _.io.niPtvu_.lt,d__,Er n_ ~ e.-,r,.s,,_,te. _n-n i M _c___ lor.t_coas_.. v ~ 19 ~ 20 EFFECTWix*anttieESMYFthilfalLETa5Tio1EEERiyHipliiiiistT.i.tisfisiiiiiiiiisi ^*nw,, whMehwr#Ah.va -Av r - --- enn'uvAw - "- -- w+mer.W" - s r
- HAW.. <andewM -
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- m. p e E
.G:ww 23 s 4gMww vg& wgh v ap _________g _m __._____ _ ______m__ 25 lipl*m~iiiififfiiiii"ifyi_ sift'iWiiiiiffiliiiii~itii@ijilit~iiFyTjil3&xiiji_niWIWlUitid - - - - - - - - _ - - - ~ - - - - ~ - ~ ~ ~ - 26 lift,i_lF_IRC_Tifi..._fff_is.i_fi_T.ss_eitiy_W._il,t..ii_s_is?_ f 4
o 1 REGULATORY ANALYSIS 2 A separate regulatory analysis was not prepared for this regulatory 3 guide. The dr:ft regulatory analysis, "."r:;;;;d Revision of 10 CFR Part 100 4 and 10 CFR Part 50," was prepared for the pr:p:: d amendments, and it provides 5 the regulatory basis for this guide and examines the costs and benefits of the 6 rule as implemented by the guide. A copy of the dr:ft regulatory analysis is 7 available for inspection and copying for a fee at the NRC Public Document 8 Room,2120LStreetNW.(LowerLevel), Washington,DC,asSecy?!105[ATEn. l l 1 5}}