Regulatory Guide 1.29: Difference between revisions

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{{Adams
{{Adams
| number = ML070310052
| number = ML13350A385
| issue date = 03/15/2007
| issue date = 02/28/1976
| title = Seismic Design Classification
| title = Seismic Design Classification
| author name =  
| author name =  
| author affiliation = NRC/RES/DFERR/DDERA/MSEB
| author affiliation = NRC/OSD
| addressee name =  
| addressee name =  
| addressee affiliation =  
| addressee affiliation =  
| docket =  
| docket =  
| license number =  
| license number =  
| contact person = Istar, Ata (301) 415-6601, RES/DFERR/ERA
| contact person =  
| case reference number = DG-1156
| document report number = RG-1.029, Rev. 2
| document report number = RG-1.029, Rev 4
| package number = ML070240135
| document type = Regulatory Guide
| document type = Regulatory Guide
| page count = 6
| page count = 3
}}
}}
{{#Wiki_filter:The U.S. Nuclear Regulatory Commission (NRC) issues regulatory guides to describe and make available to the public methods that the NRC staffconsiders acceptable for use in implementing specific parts of the agency's regulations, techniques that the staff uses in evaluating specific problemsor postulated accidents, and data that the staff need in reviewing applications for permits and licenses.  Regulatory guides are not substitutesfor regulations, and compliance with them is not required.  Methods and solutions that differ from those set forth in regulatory guides will be deemedacceptable if they provide a basis for the findings required for the issuance or continuance of a permit or license by the Comm ission.This guide was issued after consideration of comments received from the public.  The NRC staff encourages and welcomes comments and suggestionsin connection with improvements to published regulatory guides, as well as items for inclusion in regulatory guides that are currently being developed. The NRC staff will revise existing guides, as appropriate, to accommodate comments and to reflect new information or experience.  Written commentsmay be submitted to the Rules and Directives Branch, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.Regulatory guides are issued in 10 broad divisions:  1, Power Reactors; 2, Research and Test Reactors; 3, Fuels and Materials Facilities;4, Environmental and Siting; 5, Materials and Plant Protection; 6, Products; 7, Transportation; 8, Occupational Health; 9, Antitrust and Financial Review;and 10, General.Requests for single copies of draft or active regulatory guides (which may be reproduced) should be made to the U.S. Nuclear Regulatory Commission,Washington, DC 20555, Attention: Reproduction and Distribution Services Section, or by fax to (301) 415-2289; or by email to Distribution@nrc.gov. Electronic copies of this guide and other recently issued guides are available thr ough the NRC's public Web site under the Regulatory Guides documentcollection of the NRC's El ectronic Reading Room at http://www.nrc.gov/reading-rm/doc-collections/
{{#Wiki_filter:U.S. NUCLEAR REGULATORY  
and through the NRC's Agencywide DocumentsAccess and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. ML070310052.U.S. NUCLEAR REGULATORY COMMISSIONMarch 2007Revision 4 REGULATORY GUIDEOFFICE OF NUCLEAR REGULATORY RESEARCHREGULATORY GUIDE 1.29(Draft was issued as DG-1156, dated October 2006)SEISMIC DESIGN CLASSIFICATION
COMMISSION
REGULATORY  
GUIDE OFFICE OF STANDARDS
DEVELOPMENT
REGULATORY  
GUIDE 129 SEISMIC DESIGN CLASSIFICATION
Revision 2 February 1976


==A. INTRODUCTION==
==A. INTRODUCTION==
General Design Criterion (GDC) 2, "Design Bases for Protection Against Natural Phenomena,"
General Design Criterion  
of Appendix A, "General Design Criteria for Nuclear Power Plants," to Title 10, Part 50, of the Code of Federal Regulations (10 CFR Part 50), "Domestic Licensing of Production and Utilization Facilities"(Ref. 1), requires that nuclear power plant structures, systems, and components (SSCs) important to safety must be designed to withstand the effects of earthquakes without loss of capability to perform their
2, "Design Bases for Protec-tion Against Natural Phenomena," of Appendix A,"General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, "Licensing of Production and Utiliza-tion Facilities," requires that nuclear power plant structures, systems, and components important to safety be designed to withstand the effects of earthquakes without loss of capability to perform their safety functions.


safety functions.
nuclear power plants that should stand the effects of the SSE. J designed to with.A hL B. After reviewing struction permits pressurized water has developed a identifying p to withstan a splqol plications for con-o ngj 'enses for boiling and c r plants, the NRC staff* "gn classification system for ures that should be designed fec5 of the SSE. Those structumes, ents that should be designed to if the 4ZqIF n-t-vc ho rp, n vt.ei .S Appendix B, "Quality Assurance Criteria for Nuclear .-.. 1.Power Plants and Fuel Reprocessing Plants," to 10 CFR as~ic Lategory Part 50 establishes quality assurance requirements for C. REGULATORY
POSITION the design, construction, and operation of nuclear power plant structures, systems, and components that prevent e following structures, systems, and compo-or mitigate the consequences of postulated acc' n ts of a nuclear power plant, including their founda-that coubldc.aTe unuertisnto theqremntsof and of tions and supports, are designated as Seismic Category I apply to all activeit ing the e safeqtu.imd and should be designed to withstand the effects of the applyof those all rctivites, affeti ng the sfen SSE and remain functional.


Toward that end, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50 establishes quality assurance requirements for the design, construction, and operation of nuclear power plant SSCs that prevent or mitigate the consequences
The pertinent quality tions of those structures, systems, and conw~nents, assurance requirements of Appendix B to 10 CFR Part Appendix A, "Seismic and Geologic iSteria 50 should be applied to all activities affecting the for Nuclear Power Plants," to 10 CFR Part 100, safety-related functions of these structures, systems, and"Reactor Site Criteria," requ that all nuclear power components.


of postulated accidents that could cause undue risk to the health and safety of the publi
plants be designed so -the Safe Shutdown Earthquake (SSE) occurs, es, systems, and a. The reactor coolant pressure boundary.components import 0 remain functional.


====c. The pertinent====
These plant featur h essary to ensure (1) b. The reactor core and reactor vessel internals.


requirements of Appendix B apply to all activities affecting the safety-related functions of those SSCs.
the integrity of th at oant pressure boundary, (2) the capab t the reactor and maintain c. Systems' or portions of systems that are it in a safe td'n ion, or (3) the capability to required for (1) emergency core cooling, (2) postacci-prevent or a. the consequences of accidents that dent containment heat removal, or (3) postaccident could result in tial offsite exposures comparable to the guideline exposures of 10 CFR Part 100. The- system boundary includes those portions of the system I~~~I4U.U
~ ~ ~ __r r.U "LAIIJUI -A ~.A~W I UI~UI This guide describes an acceptable method of identi.fying and classifying those features of light.water.cooled ter.q di~ to acopm )Ilie spmt w n~ onl aS ItIUIlconnected piping up to and including the first valve (including a safety or relief valve) that is either nornally closed or capable of automatic closure when the safety function is require


1 Appendix S to 10 CFR Part 50 applies to applicants for a design certification or combined license pursuant to10 CFR Part 52, "Early Site Permits; Standard Design Certifications; and Combined Licenses for Nuclear Power Plants,"
====d. USNRC REGULATORY ====
or a construction permit or operating license pursuant to 10 CFR Part 50 on or after January 10, 1997.  However, the earthquake engineering criteria in Section VI of Appendix A, "Seismic and Geologic Siting Criteria for Nuclear
GUIDES should be sent to the secreary of the Commission.


Power Plants," to 10 CFR Part 100, "Reactor Site Criteria" (Ref. 2), continue to apply to operating license applicants
U.S aetlest Rgulato*r, Gweiel are fted to desincribe end make evaible to th. ib~l. Regulatory Commission.


or holders whose construction permit was issued before January 10, 1997.
Washongto,, 0 C 206. Attaenton Ooceotmg and methods acceptable
1o the NRC %lell of Implementing specific pont of the Sartce Sacton Commisson'
regulations.


2 Dose values set forth in 10 CFR Part 100, "Reactor Site Criteria" (Ref. 2), continue to apply to operating license
to delineate techniques uled by the %I&" i ovoU The guides ar Issued , the following tor broad divsons at1" sglif¢c peOblerns or poouleated accidents.


applicants or holders whose construction permits were issued before January 10, 1997. However, application
or to proetsa jog.,dnce to soplfagultaorey Guides are not substitutes fat reiatraol.fs, and conpliance I Power Reactors 6 Prodvcte woth themr t not toqruied Methods and sOlutions ditferent from those tat Out on 2 Research and Tolt Reactors I 1tanspOrletDon the guidaes wil be acceptable J9 they provide a bel fot the finding$ realusilt to 2 Fuels and Materals Facilities a Occupatiorel HMeath the or conulruunce of a Permi or ocen


of 10 CFR 50.67, "Accident source term," with the alternative source terms identified in the latest edition
====t. by the Commission ====
4 fnroonmenttl aend Sli.ng I AnttIuel Review Comment. and tuggesttunt ofa ,rmproomeflls .n that* guides are encouraged S Mterial& and Plant Protection
10 General at elf troeS and g;dmi wI t be , a.led at sporopa 0o g odlat caom manl a end to *ettIct new ,injotornron or oopefence Ioweve. Comment, on Copoge of pubklshed guides may be obltme/d by wirltn request indicating Ith Ihis guide. 0t receiead Winhr.t About two months &latet ISluafnce.


of Regulatory Guide 1.183,"Alternative Radiological Source Terms for Evaluating Design-Basis Accidents
wilt be Par divisions desired to the U S Nuclear 0aegvletory Comigneitong.


at Nuclear Power Reactors" (Ref. 3), is a voluntary option to meet the new positions in this regulatory guidance.Rev. 4 of RG 1.29, Page 2 In addition, Appendix S, "Earthquake Engineering Criteria for Nuclear Power Plants,"
Washmtlon, 0 C lcumI' usefutl in evaluating the need fat arn *calI revlsion 2065. Altaenton Director.
to 10 CFR Part 50, requires that all nuclear power plants must be designed so that certain SSCs


remain functional if the safe-shutdown earthquake ground motion (SSE) occurs.
Office a9 Standl Oletevelopment containment atmosphere weanup (e.g., hydrogen re-moval system).d. Systems' or portions of systems that are requized for (1) reactor shutdown, (2) residual heat removal, or (3) cooling the spent fuel storage pool.e. Those portions of the steam systems of boiling water reactors extending from the outermost contain-ment isolation valve up to but not including the turbine stop valve, and connected piping of 2-1/2 inches or larger nominal pipe size up to and including nhe first valve that is either normally closed or capable of automatic closure during all modes of normal reactor operation.


1  These plant features are those necessary to ensure (1) the integrity of the reactor coolant pressure boundary, (2) the capability
The turbine stop valve should be designed to withstand the SSE and maintain its integrity.


to shut down the reactor and maintain it in a safe shutdown condition, or (3) the capability to prevent
f. Those portions of the steam and feedwater systems of pressurized water reactors extending from and Including the secondary side of steam generators up to and Including the outermost containment isolation vulve, and connected piping of 2-1/2 inches or larger nominal pipe size up to and including the first valve (including a safety or relief valve) that is either normally dosed or capable of automatic closure during all modes of normal reactor operation.


or mitigate the consequences of accidents that could result in potential offsite exposures comparable to
g. Cooling water, component cooling, and auxil-iaty feedwater systems' or portions of these systems, including the intake structures, that are required for (1)emzerncy core cooling, (2) postaccident containment heat removal, (3) postaccident containment atmosphere cleanup, (4) residual heat removal from the reactor, or (5) cooling the spent fuel storage pool.h. Cooling water and seal water systems' or portions of these systems that are required for function-ing of reactor coolant system components important to safety, such as reactor coolant pumps.I. Systems' or portions of systems that are re-quired to supply fuel for emergency equipment.


the guideline exposures of 10 CFR 50.34(a)(1) or 10 CFR 100.11.
j. All electric and mechanical devices and circuitry between the process and the input terminals of the actuator systems involved in gpnerating signals that initiate protective acUon.k. Systems' or portions of systems that are required for (I) monitoring of systems important to safety and (2) actuation of systems important to safety.1. The spent fuel storage pool structure, including the fuel racks.m. The reactivity control systems, e.g., control rods, control rod drives, and boron injection system.'See footnote 1, p. 1.29-1.n. The control room, including its associated vital equipment, cooling systems for vital equipment, and life support systems, and any structures or equipment inside or outside of the control room whose failure could result in incapacitating Injury to the occupants of the control room.2 o. Primary and secondary reactor containment.


2 This guide describes a method that the staff of the U.S. Nuclear Regulatory Commission (NRC)
p. Systems,'
considers acceptable for use in identifying and classifying those features of light-water-reactor (LWR)
other than radioactive waste manage-ment systems, 3 not covered by itemns l.a through 1.o above that contain or may contain radioactive material and whose postulated failure would result in consrva-tively calculated potential offsite doses (using mete-orology as prescribed by Regulatory Guide 1.3, "As-sumptions Used for Evaluating the Potential Radio-logical Consequences of a Loss of Coolant Accident for Boiling Water Reactors," and Regulatory Guide 1.4,"Assumptions Used for Evaluating the Potential Radio-logical Consequences of a Loss of Coolant Accident for Pressurized Water Reactors")
that are more than 0.5 rem to the whole, body or its equivalent to any part of the body.q. The Class IE electric systems, including the auxiliary systems for the onsite electric power supplies, that provide the emergency electric power needed for functioning of plant features included in items l.a through Lp above.2. Those portions of structures, systems, or compo-nents whose continued function is not required but whose failure could reduce the functioning of any plnat feature included in items La through l.q above to an unacceptable safety level should be designed and con-structed so that the SSE would not cause such failure.3. Seismic Category I design requuements should extend to the first seismic restraint beyond the defined boundaries.


nuclear power plants that must be designed to withstand the effects of the SSE.
Those portions of structures, systems, or components that form interfaces between Seismic Cate-gory I and non-Seismic Category I features should be designed to Seismic Category I requirements.


This regulatory guide relates to information collections that are covered by the requirements of 10 CFR Part 50 and 10 CFR Part 100, which the Office of Management and Budget (OMB) approved
4. The pertinent quality assurance requirements of Appendix B to 10 CFR Part 50 should be applied to all activities affecting the safety-related functions of those portions of structures, systems, and components covered under Regulatory Positions
2 and 3 above.*Lie indicate substantive changes from previous issue.'Wherever practical, structures and equipment whose failure could possibly cause such injuries should be relocated or separated to the extent required to eliminate this possibility.


under OMB control numbers 3150-0011 and 3150-0093, respectively. The NRC may neither conduct
'Specific guidance on seismic requirements for radioactive waste management systems is under development.


nor sponsor, and a person is not required to respond to, an information collection request or requirement
I $I 1.29-2  
 
"I  
unless the requesting document displays a currently valid OMB control number.
 
==B. DISCUSSION==
After reviewing a number of applications for construction permits and operating licenses for boiling- and pressurized-water nuclear power plants, the NRC staff developed a seismic design
 
classification system for identifying those plant features that must be designed to withstand the effects
 
of the SSE.  In so doing, the staff designated as Seismic Category I those SSCs that must be designed
 
to remain functional if the SSE occurs.
 
3 The system boundary includes those portions of the system required to accomplish the specified safety function and connected piping up to and including the first valve (including a safety or relief valve) that is either normally closed
 
or capable of automatic closure when the safety function is required.Rev. 4 of RG 1.29, Page 3
 
==C. REGULATORY POSITION==
1.The following SSCs of a nuclear power plant, including their foundations and supports, are designated as Seismic Category I and must be designed to withstand the effects of the SSE
and remain functional.  The titles and functions of these Seismic Category I SSCs for LWR designs
 
are based on existing technology from prior applications.  Certain SSCs previously considered
 
Seismic Category I may no longer have a safety-related function requiring Seismic Category I
 
classification, and certain passive SSCs in new LWR designs may be titled differently.
 
The pertinent quality assurance requirements of Appendix B to 10 CFR Part 50 shall apply
 
to all activities affecting the safety-related functions of these SSCs:a.the reactor coolant pressure boundary b.the reactor core and reactor vessel internals c.systems 3 or portions thereof that are required for (1) emergency core cooling,(2) post-accident containment heat removal, or (3) post-accident containment atmosphere cleanup (e.g., hydrogen removal system)d.systems 2 or portions thereof that are required for (1) reactor shutdown, (2) residual heat removal, or (3) cooling the spent fuel storage poole.those portions of the steam systems of boiling-water reactors extending from the outermost containment isolation valve up to but not including the turbine stop valve, and connected piping of a nominal size of 6.35 cm (2.5 inches) or larger, up to and including the first valve that is either normally closed or capable of automatic closure during all modes of normal reactor operation (the turbine stop valve should be designed
 
to withstand the SSE and maintain its integrity)f.those portions of the steam and feedwater systems of pressurized-water reactorsextending from and including the secondary side of steam generators up to and including the outermost containment isolation valves, and connected piping of a nominal size
 
of 6.35 cm (2.5 inches) or larger, up to and including the first valve (including a safety
 
or relief valve) that is either normally closed or capable of automatic closure during all
 
modes of normal reactor operationg.cooling water, component cooling, and auxiliary feedwater systems
2 or portions thereof, including the intake structures, that are required for (1) emergency core cooling,
(2) post-accident containment heat removal, (3) post-accident containment atmosphere cleanup, (4) residual heat removal from the reactor, or (5) spent fuel storage pool coolingh.cooling water and seal water systems
2 or portions thereof that are required for functioning of reactor coolant system components important to safety, such as reactor coolant pumpsi.systems 2 or portions thereof that are required to supply fuel for emergency equipmentj.all electrical and mechanical devices and circuitry between the process and the input terminals of the actuator systems involved in generating signals that initiate protective
 
action
4 See the latest edition of Regulatory Guide 1.151, "Instrument Sensing Lines" (Ref. 4).
5 See the latest edition of Regulatory Guide 1.143, "Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants" (Ref. 5).
6 Wherever practical, structures and equipment of which failure could possibly cause such injuries should be relocated
 
or separated to the extent required to eliminate that possibility.Rev. 4 of RG 1.29, Page 4k.systems 2 or portions thereof that are required for (1) monitoring and (2) actuating systems 4 important to safetyl.the spent fuel storage pool structure, including the fuel racks m.the reactivity control systems (e.g., control rods, control rod drives, and boron injection system)n.the control room, including its associated equipment and all equipment needed to maintain the control room within safe habitability limits for personnel and safe environmental
 
limits for vital equipmento.primary and secondary reactor containment p.systems, 2 other than radioactive waste management systems, 5 not covered by items
1.a through 1.o above that contain or may contain radioactive material and of which
 
postulated failure would result in conservatively calculated potential offsite doses
 
[using meteorology as recommended in the latest editions of Regulatory Guide 1.3,
"Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Boiling-Water Reactors" (Ref. 6), Regulatory Guide 1.4,"Assumptions Used for Evaluating the Potential Radiological Consequences
 
of a Loss-of-Coolant Accident for Pressurized Water Reactors" (Ref. 7),
and Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating
 
Design-Basis Accidents at Nuclear Power Reactors" (Ref. 3)] that are more than
 
0.005 Sievert (0.5 rem) to the whole body or its equivalent to any part of the body
 
or total effective dose equivalent (TEDE), as applicableq.the Class 1E electrical systems, including the auxiliary systems for the onsite electric power supplies, that provide the emergency electric power needed for functioning
 
of plant features included in items 1.a through 1.p above2.Those portions of SSCs of which continued function is not required but of which failure could reduce the functioning of any plant feature included in items 1.a through
 
===1. q above===
 
to an unacceptable safety level or could result in incapacitating injury to occupants
 
of the control room should be designed and constructed so that the SSE would not cause
 
such failure.
 
63.At the interface between Seismic Category I and non-Seismic Category I SSCs, the Seismic Category I dynamic analysis requirements should be extended to either the first anchor point
 
in the non-seismic system or a sufficient distance into the non-Seismic Category I system
 
so that the Seismic Category I analysis remains valid.4.The pertinent quality assurance requirements of Appendix B to 10 CFR Part 50 should be applied to all activities affecting the safety-related functions of those portions of SSCs covered under
 
Regulatory Positions 2 and 3 above.5.Regulatory Guide 1.189, "Fire Protection for Operating Nuclear Power Plants" (Ref. 8), providesguidance used to establish the design requireme nts for portions of fire protection SSCs to meetthe requirements of GDC 2, as they relate to designing those SSCs to withstand the effects of the SSE.
 
Rev. 4 of RG 1.29, Page 5


==D. IMPLEMENTATION==
==D. IMPLEMENTATION==
The purpose of this section is to provide information to applicants and licensees regarding the NRC staff's plans for using this regulatory guide. No backfitting is intended or approved
The purpose of this section is to provide information to applicants regarding the NRC staff's plans for using this regulatory guide.proposes an acceptable alternative method for comply-ing with Tpecifled portions of the Commission's regula.tions, the method described herein is being and will continue to be used in the evaluation of submittals for operating license or construction permit applications until this guide is revised as a result of suggestions from the public or additional staff review.I This guide reflects current NRC staff practice.
 
in connection with its issuance.
 
Except in those cases in which an applicant or licensee proposes or has previously established an acceptable alternative method for complying with specified portions of the NRC's regulations, the NRC staff will use the methods described in this guide to evaluate (1) submittals in connection with
 
applications for construction permits, standard plant design certifications, operating licenses, early site
 
permits, and combined licenses, and (2) submittals from operating reactor licensees who voluntarily
 
propose to initiate system modifications if there is a clear nexus between the proposed modifications and
 
the subject for which guidance is provided herein.
 
REGULATORY ANALYSIS / BACKFIT ANALYSIS
The regulatory analysis and backfit analysis for this regulatory guide are availablein Draft Regulatory Guide DG-1156, "Seismic Design Classification" (Ref. 9).  The NRC issued DG-1156 in October 2006 to solicit public comment on the draft of this Revision 4 of Regulatory Guide 1.29.
 
7 All NRC regulations listed herein are available electronically through the Electronic Reading Room on the NRC's public Web site, at http://www.nrc.gov/reading-rm/doc-collections/cfr.  Copies are also available for inspection or copying for a fee from the NRC's Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDR's mailing
 
address is USNRC PDR, Washington, DC 20555; telephone (301) 415-4737 or (800) 397-4209; fax (301) 415-3548;
 
email PDR@nrc.gov
.8 All regulatory guides listed herein were published by the U.S. Nuclear Regulatory Commission or its predecessor, the U.S. Atomic Energy Commission.  Most are available electronically through the Electronic Reading Room
 
on the NRC's public Web site, at http://www.nrc.gov/reading-rm/doc-collections/reg-guides/.  Single copies of regulatory guides may also be obtained free of charge by writing the Reproduction and Distribution Services Section, ADM, USNRC, Washington, DC 20555-0001, by fax to (301) 415-2289, or by email to DISTRIBUTION@nrc.gov. Active guides may also be purchased from the National Technical Information Service (NTIS).  Details may be
 
obtained by contacting NTIS at 5285 Port Royal Road, Springfield, Virginia 22161, online at http://www.ntis.gov
,by telephone at (800) 553-NTIS (6847) or (703) 605-6000, or by fax to (703) 605-6900.  Copies are also available for
 
inspection or copying for a fee from the NRC's Public Document Room (PDR), which is located at 11555 Rockville
 
Pike, Rockville, Maryland; the PDR's mailing address is USNRC PDR, Washington, DC 20555-000
 
===1. The PDR===
 
can also be reached by telephone at (301) 415-4737 or (800) 397-4209, by fax at (301) 415-3548, and by email
 
to PDR@nrc.gov
.9 Draft Regulatory Guide DG-1156 is available electronically under Accession #ML062540294 in the NRC's Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html. Copies are also available for inspection or copying for a fee from the NRC's Public Document Room (PDR), which is
 
located at 11555 Rockville Pike, Rockville Maryland; the PDR's mailing address is USNRC PDR, Washington, DC
 
20555-0001.  The PDR can also be reached by telephone at (301) 415-4737 or (800) 397-4209 by fax
 
at (301) 415-3548, and by email to PDR@nrc.gov
.Rev. 4 of RG 1.29, Page 6 REFERENCES
1.U.S. Code of Federal Regulations , Title 10, Part 50, "Domestic Licensing of Production and Utilization Facilities," U.S. Nuclear Regulatory Commission, Washington, DC.
 
7 2.U.S. Code of Federal Regulations , Title 10, Part 100, , "Reactor Site Criteria,"
U.S. Nuclear Regulatory Commission, Washington, DC.
 
73.Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design-Basis Accidents at Nuclear Power Reactors," U.S. Nuclear Regulatory Commission, Washington, DC.
 
84.Regulatory Guide 1.51, "Instrument Sensing Lines," U.S. Nuclear Regulatory Commission, Washington, DC.
 
85.Regulatory Guide 1.143, "Design Guidance for Radioactive Waste Management Systems,Structures, and Components Installed in Light-Water-Cooled Nu clear Power Plants,"
U.S. Nuclear Regulatory Commission, Washington, DC.
 
86.Regulatory Guide 1.3, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Boiling-Water Reactors,"
 
U.S. Nuclear Regulatory Commission, Washington, DC.
 
87.Regulatory Guide 1.4, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors,"
 
U.S. Nuclear Regulatory Commission, Washington, DC.
 
88.Regulatory Guide 1.189, "Fire Protection for Operating Nuclear Power Plants,"
U.S. Nuclear Regulatory Commission, Washington, DC.
 
89.Draft Regulatory Guide DG-1156, "Seismic Design Classification,"
U.S. Nuclear Regulatory Commission, Washington, DC, October 2006.


9}}
There.fore, except in those 'cases In 'which the applicant 1.29.3}}


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Revision as of 00:05, 15 July 2019

Seismic Design Classification
ML13350A385
Person / Time
Issue date: 02/28/1976
From:
NRC/OSD
To:
References
RG-1.029, Rev. 2
Download: ML13350A385 (3)
v • d • e


U.S. NUCLEAR REGULATORY

COMMISSION

REGULATORY

GUIDE OFFICE OF STANDARDS

DEVELOPMENT

REGULATORY

GUIDE 129 SEISMIC DESIGN CLASSIFICATION

Revision 2 February 1976

A. INTRODUCTION

General Design Criterion 2, "Design Bases for Protec-tion Against Natural Phenomena," of Appendix A,"General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, "Licensing of Production and Utiliza-tion Facilities," requires that nuclear power plant structures, systems, and components important to safety be designed to withstand the effects of earthquakes without loss of capability to perform their safety functions.

nuclear power plants that should stand the effects of the SSE. J designed to with.A hL B. After reviewing struction permits pressurized water has developed a identifying p to withstan a splqol plications for con-o ngj 'enses for boiling and c r plants, the NRC staff* "gn classification system for ures that should be designed fec5 of the SSE. Those structumes, ents that should be designed to if the 4ZqIF n-t-vc ho rp, n vt.ei .S Appendix B, "Quality Assurance Criteria for Nuclear .-.. 1.Power Plants and Fuel Reprocessing Plants," to 10 CFR as~ic Lategory Part 50 establishes quality assurance requirements for C. REGULATORY

POSITION the design, construction, and operation of nuclear power plant structures, systems, and components that prevent e following structures, systems, and compo-or mitigate the consequences of postulated acc' n ts of a nuclear power plant, including their founda-that coubldc.aTe unuertisnto theqremntsof and of tions and supports, are designated as Seismic Category I apply to all activeit ing the e safeqtu.imd and should be designed to withstand the effects of the applyof those all rctivites, affeti ng the sfen SSE and remain functional.

The pertinent quality tions of those structures, systems, and conw~nents, assurance requirements of Appendix B to 10 CFR Part Appendix A, "Seismic and Geologic iSteria 50 should be applied to all activities affecting the for Nuclear Power Plants," to 10 CFR Part 100, safety-related functions of these structures, systems, and"Reactor Site Criteria," requ that all nuclear power components.

plants be designed so -the Safe Shutdown Earthquake (SSE) occurs, es, systems, and a. The reactor coolant pressure boundary.components import 0 remain functional.

These plant featur h essary to ensure (1) b. The reactor core and reactor vessel internals.

the integrity of th at oant pressure boundary, (2) the capab t the reactor and maintain c. Systems' or portions of systems that are it in a safe td'n ion, or (3) the capability to required for (1) emergency core cooling, (2) postacci-prevent or a. the consequences of accidents that dent containment heat removal, or (3) postaccident could result in tial offsite exposures comparable to the guideline exposures of 10 CFR Part 100. The- system boundary includes those portions of the system I~~~I4U.U

~ ~ ~ __r r.U "LAIIJUI -A ~.A~W I UI~UI This guide describes an acceptable method of identi.fying and classifying those features of light.water.cooled ter.q di~ to acopm )Ilie spmt w n~ onl aS ItIUIlconnected piping up to and including the first valve (including a safety or relief valve) that is either nornally closed or capable of automatic closure when the safety function is require

d. USNRC REGULATORY

GUIDES should be sent to the secreary of the Commission.

U.S aetlest Rgulato*r, Gweiel are fted to desincribe end make evaible to th. ib~l. Regulatory Commission.

Washongto,, 0 C 206. Attaenton Ooceotmg and methods acceptable

1o the NRC %lell of Implementing specific pont of the Sartce Sacton Commisson'

regulations.

to delineate techniques uled by the %I&" i ovoU The guides ar Issued , the following tor broad divsons at1" sglif¢c peOblerns or poouleated accidents.

or to proetsa jog.,dnce to soplfagultaorey Guides are not substitutes fat reiatraol.fs, and conpliance I Power Reactors 6 Prodvcte woth themr t not toqruied Methods and sOlutions ditferent from those tat Out on 2 Research and Tolt Reactors I 1tanspOrletDon the guidaes wil be acceptable J9 they provide a bel fot the finding$ realusilt to 2 Fuels and Materals Facilities a Occupatiorel HMeath the or conulruunce of a Permi or ocen

t. by the Commission

4 fnroonmenttl aend Sli.ng I AnttIuel Review Comment. and tuggesttunt ofa ,rmproomeflls .n that* guides are encouraged S Mterial& and Plant Protection

10 General at elf troeS and g;dmi wI t be , a.led at sporopa 0o g odlat caom manl a end to *ettIct new ,injotornron or oopefence Ioweve. Comment, on Copoge of pubklshed guides may be obltme/d by wirltn request indicating Ith Ihis guide. 0t receiead Winhr.t About two months &latet ISluafnce.

wilt be Par divisions desired to the U S Nuclear 0aegvletory Comigneitong.

Washmtlon, 0 C lcumI' usefutl in evaluating the need fat arn *calI revlsion 2065. Altaenton Director.

Office a9 Standl Oletevelopment containment atmosphere weanup (e.g., hydrogen re-moval system).d. Systems' or portions of systems that are requized for (1) reactor shutdown, (2) residual heat removal, or (3) cooling the spent fuel storage pool.e. Those portions of the steam systems of boiling water reactors extending from the outermost contain-ment isolation valve up to but not including the turbine stop valve, and connected piping of 2-1/2 inches or larger nominal pipe size up to and including nhe first valve that is either normally closed or capable of automatic closure during all modes of normal reactor operation.

The turbine stop valve should be designed to withstand the SSE and maintain its integrity.

f. Those portions of the steam and feedwater systems of pressurized water reactors extending from and Including the secondary side of steam generators up to and Including the outermost containment isolation vulve, and connected piping of 2-1/2 inches or larger nominal pipe size up to and including the first valve (including a safety or relief valve) that is either normally dosed or capable of automatic closure during all modes of normal reactor operation.

g. Cooling water, component cooling, and auxil-iaty feedwater systems' or portions of these systems, including the intake structures, that are required for (1)emzerncy core cooling, (2) postaccident containment heat removal, (3) postaccident containment atmosphere cleanup, (4) residual heat removal from the reactor, or (5) cooling the spent fuel storage pool.h. Cooling water and seal water systems' or portions of these systems that are required for function-ing of reactor coolant system components important to safety, such as reactor coolant pumps.I. Systems' or portions of systems that are re-quired to supply fuel for emergency equipment.

j. All electric and mechanical devices and circuitry between the process and the input terminals of the actuator systems involved in gpnerating signals that initiate protective acUon.k. Systems' or portions of systems that are required for (I) monitoring of systems important to safety and (2) actuation of systems important to safety.1. The spent fuel storage pool structure, including the fuel racks.m. The reactivity control systems, e.g., control rods, control rod drives, and boron injection system.'See footnote 1, p. 1.29-1.n. The control room, including its associated vital equipment, cooling systems for vital equipment, and life support systems, and any structures or equipment inside or outside of the control room whose failure could result in incapacitating Injury to the occupants of the control room.2 o. Primary and secondary reactor containment.

p. Systems,'

other than radioactive waste manage-ment systems, 3 not covered by itemns l.a through 1.o above that contain or may contain radioactive material and whose postulated failure would result in consrva-tively calculated potential offsite doses (using mete-orology as prescribed by Regulatory Guide 1.3, "As-sumptions Used for Evaluating the Potential Radio-logical Consequences of a Loss of Coolant Accident for Boiling Water Reactors," and Regulatory Guide 1.4,"Assumptions Used for Evaluating the Potential Radio-logical Consequences of a Loss of Coolant Accident for Pressurized Water Reactors")

that are more than 0.5 rem to the whole, body or its equivalent to any part of the body.q. The Class IE electric systems, including the auxiliary systems for the onsite electric power supplies, that provide the emergency electric power needed for functioning of plant features included in items l.a through Lp above.2. Those portions of structures, systems, or compo-nents whose continued function is not required but whose failure could reduce the functioning of any plnat feature included in items La through l.q above to an unacceptable safety level should be designed and con-structed so that the SSE would not cause such failure.3. Seismic Category I design requuements should extend to the first seismic restraint beyond the defined boundaries.

Those portions of structures, systems, or components that form interfaces between Seismic Cate-gory I and non-Seismic Category I features should be designed to Seismic Category I requirements.

4. The pertinent quality assurance requirements of Appendix B to 10 CFR Part 50 should be applied to all activities affecting the safety-related functions of those portions of structures, systems, and components covered under Regulatory Positions

2 and 3 above.*Lie indicate substantive changes from previous issue.'Wherever practical, structures and equipment whose failure could possibly cause such injuries should be relocated or separated to the extent required to eliminate this possibility.

'Specific guidance on seismic requirements for radioactive waste management systems is under development.

I $I 1.29-2

"I

D. IMPLEMENTATION

The purpose of this section is to provide information to applicants regarding the NRC staff's plans for using this regulatory guide.proposes an acceptable alternative method for comply-ing with Tpecifled portions of the Commission's regula.tions, the method described herein is being and will continue to be used in the evaluation of submittals for operating license or construction permit applications until this guide is revised as a result of suggestions from the public or additional staff review.I This guide reflects current NRC staff practice.

There.fore, except in those 'cases In 'which the applicant 1.29.3


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