Regulatory Guide 1.29

Seismic Design Classification

ML13350A385
Person / Time
Issue date:	02/28/1976
From:	NRC/OSD
To:
References
RG-1.029, Rev. 2
Download: ML13350A385 (3)

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.

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designed to with.

A

hL

B. DISC*

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 .

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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

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This guide describes an acceptable method of identi.

fying and classifying those features of light.water.cooled ter.q di~ to acopm d;*U]]

)Ilie spmt w

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ItIUIl connected 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 required.

USNRC REGULATORY GUIDES

commen*s, 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 sopl c*t, fagultaorey 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 *.suance 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 accom*

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 $

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1.29-2

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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

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