Regulatory Guide 1.29: Difference between revisions

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{{Adams
{{Adams
| number = ML070310052
| number = ML003739983
| issue date = 03/15/2007
| issue date = 09/30/1978
| title = Seismic Design Classification
| title = Seismic Design Classification
| author name =  
| author name =  
| author affiliation = NRC/RES/DFERR/DDERA/MSEB
| author affiliation = NRC/RES
| 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 3
| 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 COMMISSION
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
Revision 3 September 1978 REGULATORY GUIDE
OFFICE OF STANDARDS DEVELOPMENT
REGULATORY GUIDE 1.29 SEISMIC DESIGN CLASSIFICATION
A.


==A. INTRODUCTION==
INTRODUCTION  
General Design Criterion (GDC) 2, "Design Bases for Protection Against Natural Phenomena,"
General Design Criterion 2, "Design Bases for Protection Against Natural Phenomena," of Appen dix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," requires that nuclear power plant structures, systems, and components important to safety be designed to with stand the effects of earthquakes without loss of capa bility to perform their safety functions.
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


safety functions.
- Appendix B, "Quality Assurance Criteria for Nu clear Power Plants and Fuel Reprocessing Plants," to
10 CFR Part 50 establishes quality assurance re quirements for the design, construction, and opera tion of nuclear power plant structures, systems, and components that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the publi


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
====c. The pertinent re====
*/
quirements of Appendix B apply to all activities af fecting the safety-related functions of those struc tures, systems, and components.


of postulated accidents that could cause undue risk to the health and safety of the publi
Appendix A, "Seismic and Geologic Siting Criteria for Nuclear Power Plants," to 10 CFR Part
100, "Reactor Site Criteria," requires that all nu clear power plants be designed so that, if the Safe
*
Shutdown Earthquake (SSE) occurs, certain struc tures, systems, and components remain functional.


====c. The pertinent====
These plant features are those necessary to ensure (1)
the integrity of the reactor'coolant pressure boundary,
(2) the capability to shut down the reactor and main tain it in a safe shutdown condition, or (3) the capa bility to prevent or mitigate the consequences of ac cidents that could result in potential offsite exposures comparable to the guideline exposures of 10 CFR
Part 100.


requirements of Appendix B apply to all activities affecting the safety-related functions of those SSCs.
This guide describes a method acceptable to the NRC staff for identifying and classifying those fea- tures of light-water-cooled nuclear power plants that should be designed to withstand the effects of the SSE. The Advisory Committee on Reactor Safeguards has been consulted regarding this guide and has concurred in the regulatory position.


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,"
B.
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


Power Plants," to 10 CFR Part 100, "Reactor Site Criteria" (Ref. 2), continue to apply to operating license applicants
DISCUSSION
After reviewing a number of applications for con struction permits and operating licenses for boiling and pressurized water nuclear power plants, the NRC
staff has developed a seismic design classification system for identifying those plant features that should be designed to withstand the effects of the SSE.


or holders whose construction permit was issued before January 10, 1997.
Those structures, systems, and components that should be designed to remain functional if the SSE
occurs have been designated as Seismic Category I.


2 Dose values set forth in 10 CFR Part 100, "Reactor Site Criteria" (Ref. 2), continue to apply to operating license
C.


applicants or holders whose construction permits were issued before January 10, 1997. However, application
REGULATORY POSITION
1. The following structures, systems, and compo nents of a nuclear power plant, including their foun dations and supports, are designated as Seismic Cate gory I and should be designed to withstand the effects of the SSE and remain functional. The pertinent qual ity assurance requirements of Appendix B to 10 CFR
Part 50 should be applied to all activities affecting the safety-related functions of these structures, sys tems, and components.


of 10 CFR 50.67, "Accident source term," with the alternative source terms identified in the latest edition
a. The reactor coolant pressure boundary.


of Regulatory Guide 1.183,"Alternative Radiological Source Terms for Evaluating Design-Basis Accidents
b. The reactor core and reactor vessel internals.


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,"
c. Systems' or portions of systems that are re quired for (1) emergency core cooling, (2) postacci
to 10 CFR Part 50, requires that all nuclear power plants must be designed so that certain SSCs
* Lines indicate substantive changes from previous issue.


remain functional if the safe-shutdown earthquake ground motion (SSE) occurs.
tThe system boundary includes those portions of the system re quired 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.


1  These plant features are those necessary to ensure (1) the integrity of the reactor coolant pressure boundary, (2) the capability
USNRC REGULATORY GUIDES
Comments should be sent to the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, D.C. 20556, Attention: Docketing and Regulatory Guides are issued to describe and make available to the public Service Branch.


to shut down the reactor and maintain it in a safe shutdown condition, or (3) the capability to prevent
methods acceptable to the NRC staff of implementing specific parts of the Commission's regulations, to delineate techniques used by the staff in evalu- The guides are issued in the following ten broad divisions:
ating specific problems or postulated accidents, or to provide guidance to applicants. Regulatory Guides are not substitutes for regulations, and com-


or mitigate the consequences of accidents that could result in potential offsite exposures comparable to
===1. Power Reactors ===
6. Products pllance with them is not required. Methods and solutions different from those
2. Research and Test Reactors
7. Transportation set out in the guides will be acceptable if they provide a basis for the findings
3. Fuels and Materials Facilities
8. Occupational Health requisite to the issuance or continuance of a permit or license by the  
4. Environmental and Siting
9. Antitrust and Financial Review Commission.


the guideline exposures of 10 CFR 50.34(a)(1) or 10 CFR 100.11.
5. Materials and Plant Protection
10. General Requests for single copies of issued guides lwhich may be reproduced) or for Comments and suggestions for improvements in these guides are encouraged at placement on an automatic distribution list for single copies of future guides all times, and guides will be revised, as appropriate, to accommodate comments in specific divisions should be made in writing to the U.S. Nuclear Regulatory and to reflect new information or experience. This guide was revised as a resultr Commission, Washington, D.C. 20555, Attention:
Director, Division of of substantive comments received from the public and additional staff review.


2 This guide describes a method that the staff of the U.S. Nuclear Regulatory Commission (NRC)
Technical Information and Document Control.
considers acceptable for use in identifying and classifying those features of light-water-reactor (LWR)


nuclear power plants that must be designed to withstand the effects of the SSE.
dent containment heat removal, or (3) postaccident containment atmosphere cleanup (e.g,, hydrogen re moval system);
d, Systems' or portions of systems that are re quired for (1) reactor shutdown, (2) residual heat re moval, or (3) cooling the spent fuel storage pool, e. Those portions of the steam systems of boil ing water reactors extending from the outermost con tainment isolation valve up to but not including the turbine stop valve, and connected piping of 21/2 inches or larger nominal pipe size up to and including the first valve that is either normally closed or capa ble 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.


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
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 isola tion valves, and connected piping of 21/2 inches or larger nominal pipe size up to and including the first valve (including a safety or relief valve) that is either normally closed or capable of automatic closure dur ing all modes of normal reactor operation, g. Cooling water, component cooling, and auxiliary feedwater systems ' or portions of these sys tems, including the intake structures, that are re quired for (1) emergency core cooling, (2) postacci dent containment heat removal, (3) postaccident con tainment atmosphere cleanup, (4) residual heat re moval from the reactor, or (5) cooling the spent fuel storage pool.


under OMB control numbers 3150-0011 and 3150-0093, respectively. The NRC may neither conduct
h. Cooling water and seal water systemsI or portions of these systems that are required for func tioning of reactor coolant system components impor tant to safety, such as reactor coolant pumps.


nor sponsor, and a person is not required to respond to, an information collection request or requirement
i. SystemsI or portions of systems that are re quired to supply fuel for emergency equipment.


unless the requesting document displays a currently valid OMB control number.
j. All electric and mechanical devices and cir cuitry between the process and the input terminals of the actuator systems involved in generating signals that initiate protective action, k. SystemsI or portions of systems that are re quired for (1) monitoring of systems important to safety and (2) actuation of systems important to safety.


==B. DISCUSSION==
1. The spent fuel storage pool structure, includ ing the fuel racks.
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
m. The reactivity control systems, e.g., control rods, control rod drives and boron injection system.


of the SSE. In so doing, the staff designated as Seismic Category I those SSCs that must be designed
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 equipment, o, Primary and secondary reactor containment.


to remain functional if the SSE occurs.
p. Systems,' Dther than radioactive waste man agement systems, 2 not covered by items I.a through L.o above that contain or may contain radioactive ma terial and whose postulated failure would result in conservatively calculated potential offsite doses (us ing meteorology as recommended in Regulatory Guide 1.3, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors," and Regulatory Guide 1,4, "Assumptions Used for Evaluating the Potential Radiological 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.


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
q. The Class 1E 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 1 ,p above.


or capable of automatic closure when the safety function is required.Rev. 4 of RG 1.29, Page 3  
2. Those portions of structures, systems, or com ponents whose continued function is not required but whose failure could reduce the functioning of any plant feature included in items I .a through l.q above to an unacceptable safety level or could result in in capacitating injury to occupants of the control room should be designed and constructed so that the SSE
would not cause such failure,3  
3, Seismic Category I design requirements should extend to the first seismic restraint beyond the de fined boundaries, Those portions of structures, sys tems, or components that form interfaces between Seismic Category I and non-Seismic Category I fea tures should be designed to Seismic Category I
requirements.


==C. REGULATORY POSITION==
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 compo nents covered under Regulatory Positions 2 and 3 above,
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
2 Specific guidance on seismic requirements for radioactive waste management systems is under development.
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
3Wherever practical, structures and equipment whose failure could possibly cause such injuries should be relocated or sepa rated to the extent required to eliminate thiB possibility.


Seismic Category I may no longer have a safety-related function requiring Seismic Category I
1.29-2


classification, and certain passive SSCs in new LWR designs may be titled differently.
0, IMPLEMENTATION
The purpose of this section is to provide informa tion to applicants regarding the NRC staff's plans for using this regulatory guide, This guide reflects current NRC staff practice.


The pertinent quality assurance requirements of Appendix B to 10 CFR Part 50 shall apply
Therefore, except in those cases in which the appli- cant proposes an acceptable alternative method for complying with specified portions of the Commis sion's regulations, 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 re sult of suggestions from the public or additional staff review.


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
1,29-3}}
 
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==
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
 
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}}


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Latest revision as of 02:08, 17 January 2025

Seismic Design Classification
ML003739983
Person / Time
Issue date: 09/30/1978
From:
Office of Nuclear Regulatory Research
To:
References
RG-1.029, Rev 3
Download: ML003739983 (3)
v • d • e


U.S. NUCLEAR REGULATORY COMMISSION

Revision 3 September 1978 REGULATORY GUIDE

OFFICE OF STANDARDS DEVELOPMENT

REGULATORY GUIDE 1.29 SEISMIC DESIGN CLASSIFICATION

A.

INTRODUCTION

General Design Criterion 2, "Design Bases for Protection Against Natural Phenomena," of Appen dix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," requires that nuclear power plant structures, systems, and components important to safety be designed to with stand the effects of earthquakes without loss of capa bility to perform their safety functions.

- Appendix B, "Quality Assurance Criteria for Nu clear Power Plants and Fuel Reprocessing Plants," to

10 CFR Part 50 establishes quality assurance re quirements for the design, construction, and opera tion of nuclear power plant structures, systems, and components that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the publi

c. The pertinent re

  • /

quirements of Appendix B apply to all activities af fecting the safety-related functions of those struc tures, systems, and components.

Appendix A, "Seismic and Geologic Siting Criteria for Nuclear Power Plants," to 10 CFR Part

100, "Reactor Site Criteria," requires that all nu clear power plants be designed so that, if the Safe

Shutdown Earthquake (SSE) occurs, certain struc tures, systems, and components remain functional.

These plant features are those necessary to ensure (1)

the integrity of the reactor'coolant pressure boundary,

(2) the capability to shut down the reactor and main tain it in a safe shutdown condition, or (3) the capa bility to prevent or mitigate the consequences of ac cidents that could result in potential offsite exposures comparable to the guideline exposures of 10 CFR

Part 100.

This guide describes a method acceptable to the NRC staff for identifying and classifying those fea- tures of light-water-cooled nuclear power plants that should be designed to withstand the effects of the SSE. The Advisory Committee on Reactor Safeguards has been consulted regarding this guide and has concurred in the regulatory position.

B.

DISCUSSION

After reviewing a number of applications for con struction permits and operating licenses for boiling and pressurized water nuclear power plants, the NRC

staff has developed a seismic design classification system for identifying those plant features that should be designed to withstand the effects of the SSE.

Those structures, systems, and components that should be designed to remain functional if the SSE

occurs have been designated as Seismic Category I.

C.

REGULATORY POSITION

1. The following structures, systems, and compo nents of a nuclear power plant, including their foun dations and supports, are designated as Seismic Cate gory I and should be designed to withstand the effects of the SSE and remain functional. The pertinent qual ity assurance requirements of Appendix B to 10 CFR

Part 50 should be applied to all activities affecting the safety-related functions of these structures, sys tems, and components.

a. The reactor coolant pressure boundary.

b. The reactor core and reactor vessel internals.

c. Systems' or portions of systems that are re quired for (1) emergency core cooling, (2) postacci

  • Lines indicate substantive changes from previous issue.

tThe system boundary includes those portions of the system re quired 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.

USNRC REGULATORY GUIDES

Comments should be sent to the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, D.C. 20556, Attention: Docketing and Regulatory Guides are issued to describe and make available to the public Service Branch.

methods acceptable to the NRC staff of implementing specific parts of the Commission's regulations, to delineate techniques used by the staff in evalu- The guides are issued in the following ten broad divisions:

ating specific problems or postulated accidents, or to provide guidance to applicants. Regulatory Guides are not substitutes for regulations, and com-

1. Power Reactors

6. Products pllance with them is not required. Methods and solutions different from those

2. Research and Test Reactors

7. Transportation set out in the guides will be acceptable if they provide a basis for the findings

3. Fuels and Materials Facilities

8. Occupational Health requisite to the issuance or continuance of a permit or license by the

4. Environmental and Siting

9. Antitrust and Financial Review Commission.

5. Materials and Plant Protection

10. General Requests for single copies of issued guides lwhich may be reproduced) or for Comments and suggestions for improvements in these guides are encouraged at placement on an automatic distribution list for single copies of future guides all times, and guides will be revised, as appropriate, to accommodate comments in specific divisions should be made in writing to the U.S. Nuclear Regulatory and to reflect new information or experience. This guide was revised as a resultr Commission, Washington, D.C. 20555, Attention:

Director, Division of of substantive comments received from the public and additional staff review.

Technical Information and Document Control.

dent containment heat removal, or (3) postaccident containment atmosphere cleanup (e.g,, hydrogen re moval system);

d, Systems' or portions of systems that are re quired for (1) reactor shutdown, (2) residual heat re moval, or (3) cooling the spent fuel storage pool, e. Those portions of the steam systems of boil ing water reactors extending from the outermost con tainment isolation valve up to but not including the turbine stop valve, and connected piping of 21/2 inches or larger nominal pipe size up to and including the first valve that is either normally closed or capa ble 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 isola tion valves, and connected piping of 21/2 inches or larger nominal pipe size up to and including the first valve (including a safety or relief valve) that is either normally closed or capable of automatic closure dur ing all modes of normal reactor operation, g. Cooling water, component cooling, and auxiliary feedwater systems ' or portions of these sys tems, including the intake structures, that are re quired for (1) emergency core cooling, (2) postacci dent containment heat removal, (3) postaccident con tainment atmosphere cleanup, (4) residual heat re moval from the reactor, or (5) cooling the spent fuel storage pool.

h. Cooling water and seal water systemsI or portions of these systems that are required for func tioning of reactor coolant system components impor tant to safety, such as reactor coolant pumps.

i. SystemsI or portions of systems that are re quired to supply fuel for emergency equipment.

j. All electric and mechanical devices and cir cuitry between the process and the input terminals of the actuator systems involved in generating signals that initiate protective action, k. SystemsI or portions of systems that are re quired for (1) monitoring of systems important to safety and (2) actuation of systems important to safety.

1. The spent fuel storage pool structure, includ ing 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 equipment, o, Primary and secondary reactor containment.

p. Systems,' Dther than radioactive waste man agement systems, 2 not covered by items I.a through L.o above that contain or may contain radioactive ma terial and whose postulated failure would result in conservatively calculated potential offsite doses (us ing meteorology as recommended in Regulatory Guide 1.3, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors," and Regulatory Guide 1,4, "Assumptions Used for Evaluating the Potential Radiological 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 1E 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 1 ,p above.

2. Those portions of structures, systems, or com ponents whose continued function is not required but whose failure could reduce the functioning of any plant feature included in items I .a through l.q above to an unacceptable safety level or could result in in capacitating injury to occupants of the control room should be designed and constructed so that the SSE

would not cause such failure,3

3, Seismic Category I design requirements should extend to the first seismic restraint beyond the de fined boundaries, Those portions of structures, sys tems, or components that form interfaces between Seismic Category I and non-Seismic Category I fea tures 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 compo nents covered under Regulatory Positions 2 and 3 above,

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

3Wherever practical, structures and equipment whose failure could possibly cause such injuries should be relocated or sepa rated to the extent required to eliminate thiB possibility.

1.29-2

0, IMPLEMENTATION

The purpose of this section is to provide informa tion to applicants regarding the NRC staff's plans for using this regulatory guide, This guide reflects current NRC staff practice.

Therefore, except in those cases in which the appli- cant proposes an acceptable alternative method for complying with specified portions of the Commis sion's regulations, 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 re sult of suggestions from the public or additional staff review.

1,29-3


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