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:U.S. NUCLEAR REGULATORY COMMISSION                                                                 March 2007 Revision 4 REGULATORY GUIDE
{{#Wiki_filter:U.S. NUCLEAR REGULATORY COMMISSION                                                                                                                         Revision 2 February 1976 REGULATORY GUIDE
                                    OFFICE OF NUCLEAR REGULATORY RESEARCH
OFFICE OF STANDARDS DEVELOPMENT
                                              REGULATORY GUIDE 1.29 (Draft was issued as DG-1156, dated October 2006)
                                                              REGULATORY GUIDE 129 SEISMIC DESIGN CLASSIFICATION
                                  SEISMIC DESIGN CLASSIFICATION


==A. INTRODUCTION==
==A. INTRODUCTION==
General Design Criterion (GDC) 2, Design Bases for Protection Against Natural Phenomena, 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.
nuclear power plants that should                              designed to with.


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 of postulated accidents that could cause undue risk to the health and safety of the public. The pertinent requirements of Appendix B apply to all activities affecting the safety-related functions of those SSCs.
stand the effects of the SSE.                        J
                                                                                                                                    A      hL
    General Design Criterion 2, "Design Bases for Protec- tion Against Natural Phenomena," of Appendix A,                                                                B. DISC*
"General Design Criteria for Nuclear Power Plants," to
10 CFR Part 50, "Licensing of Production and Utiliza-                                  After reviewing a splqol                            plications for con- struction permits                    o          ngj
                                                                                                                                    'enses          for boiling and tion Facilities," requires that nuclear power plant pressurized water c                                r plants, the NRC staff structures, systems, and components important to safety                                                                        "gn classification system for has developed a be designed to withstand the effects of earthquakes identifying                p                ures that should be designed without loss of capability to perform their safety to withstan                    fec5 of the SSE. Those structumes, functions.


The U.S. Nuclear Regulatory Commission (NRC) issues regulatory guides to describe and make available to the public methods that the NRC staff considers acceptable for use in implementing specific parts of the agencys regulations, techniques that the staff uses in evaluating specific problems or postulated accidents, and data that the staff need in reviewing applications for permits and licenses. Regulatory guides are not substitutes for regulations, and compliance with them is not required. Methods and solutions that differ from those set forth in regulatory guides will be deemed acceptable if they provide a basis for the findings required for the issuance or continuance of a permit or license by the Commission.
S                                   ents that should be designed to if the 4ZqIF n-t-vc ho rp,                       n vt.ei .
    Appendix B, "Quality Assurance Criteria for Nuclear                              .-..                                   1.


This guide was issued after consideration of comments received from the public. The NRC staff encourages and welcomes comments and suggestions in connection with improvements to published regulatory guides, as well as items for inclusion in regulatory guides that are currently being developed.
Power Plants and Fuel Reprocessing Plants," to 10 CFR                                        as~icLategory Part 50 establishes quality assurance requirements for                                                


The NRC staff will revise existing guides, as appropriate, to accommodate comments and to reflect new information or experience. Written comments may be submitted to the Rules and Directives Branch, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.
==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 apply to all activeit                      eing the safeqtu.imd                  and should be designed to withstand the effects of theI
applyofthose  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.


Regulatory guides are issued in 10 broad divisions: 1, Power Reactors; 2, Research and Test Reactors; 3, Fuels and Materials Facilities;
plants be designed so -                              the Safe Shutdown Earthquake (SSE) occurs,                                 es, systems, and                 a. The reactor coolant pressure boundary.
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.
components import                      0              remain functional.


Electronic copies of this guide and other recently issued guides are available through the NRCs public Web site under the Regulatory Guides document collection of the NRCs Electronic Reading Room at http://www.nrc.gov/reading-rm/doc-collections/ and through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. ML070310052.
These plant featurh                                  essary to ensure (1)                 b. The reactor core and reactor vessel internals.


In addition, Appendix S, Earthquake Engineering Criteria for Nuclear Power Plants, 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.1 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 maintain it in a safe shutdown condition, or (3) the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guideline exposures of 10 CFR 50.34(a)(1) or 10 CFR 100.11. 2 This guide describes a method that the staff of the U.S. Nuclear Regulatory Commission (NRC)
the integrity of th                at          oant pressure boundary,
considers acceptable for use in identifying and classifying those features of light-water-reactor (LWR)
(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 ter.q    di~  to
nuclear power plants that must be designed to withstand the effects of the SSE.
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                                                                                                                                  ~~.A~W__r spmt    w I
                                                                                                                                                    L*A.


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 under OMB control numbers 3150-0011 and 3150-0093, respectively. The NRC may neither conduct nor sponsor, and a person is not required to respond to, an information collection request or requirement unless the requesting document displays a currently valid OMB control number.
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                                                                                                                                                              n~
                                                                                                                                                                  "LAIIJUI
                                                                                                                                                              ItIUIl onl  aS
                                                                                                                                                                              -A
                                                                                    connected piping up to and including the first valve (including a This guide describes an acceptable method of identi.                            safety or relief valve) that is either nornally closed or capable fying and classifying those features of light.water.cooled                          of automatic closure when the safety function is required.


==B. DISCUSSION==
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 %lellof 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 themrt not toqruied Methods and sOlutions ditferent from those tat Out on    2 Research and Tolt Reactors                I 1tanspOrletDon the guidaes wil be acceptable J9they provide a bel fot the finding$ realusilt to 2 Fuels and Materals Facilities            a Occupatiorel HMeath the *.suance or conulruunce of a Permi or ocent. by the Commission                4    fnroonmenttl aendSli.ng                I AnttIuel Review Comment. and tuggesttuntt ofa,rmproomeflls .n that* guides are encouraged          S Mterial& and Plant Protection          10 General at elf troeS and g;dmi wI be , a.led at sporopa g0oaccom* odlat caom manla end to *ettIctnew ,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
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.
lcumI' usefutl in evaluating the need fat arn*calI revlsion                        2065. Altaenton Director. Office a9 Standl          Oletevelopment


1 Appendix S to 10 CFR Part 50 applies to applicants for a design certification or combined license pursuant to
containment atmosphere weanup (e.g., hydrogen re-                          n. The control room, including its associated vital moval system).                                                      equipment, cooling systems for vital equipment, and life support systems, and any structures or equipment inside d. Systems' or portions of systems that are                  or outside of the control room whose failure could result requized for (1) reactor shutdown, (2) residual heat                in incapacitating  Injury to the occupants of the control
        10 CFR Part 52, Early Site Permits; Standard Design Certifications; and Combined Licenses for Nuclear Power Plants, 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 or holders whose construction permit was issued before January 10, 1997.
                                                                            2 removal, or (3) cooling the spent fuel storage pool.                room.


2 Dose values set forth in 10 CFR Part 100, Reactor Site Criteria (Ref. 2), continue to apply to operating license applicants or holders whose construction permits were issued before January 10, 1997. However, application of 10 CFR 50.67, Accident source term, with the alternative source terms identified in the latest edition of Regulatory Guide 1.183,Alternative Radiological Source Terms for Evaluating Design-Basis Accidents at Nuclear Power Reactors (Ref. 3), is a voluntary option to meet the new positions in this regulatory guidance.
e. Those portions of the steam systems of boiling water reactors extending from the outermost contain-                       o. Primary and secondary reactor containment.


Rev. 4 of RG 1.29, Page 2
ment isolation valve up to but not including the turbine stop valve, and connected piping of 2-1/2 inches or                        p. Systems,' other than radioactive waste manage- larger nominal pipe size up to and including nhe first            ment systems,3 not covered by itemns l.a through 1.o              I $
valve that is either normally closed or capable of                above that contain or may contain radioactive material automatic closure during all modes of normal reactor              and whose postulated failure would result in consrva- operation. The turbine stop valve should be designed to            tively calculated potential offsite doses (using mete- withstand the SSE and maintain its integrity.                      orology as prescribed by Regulatory Guide 1.3, "As- sumptions Used for Evaluating the Potential Radio- f. Those portions of the steam and feedwater                logical Consequences of a Loss of Coolant Accident for systems of pressurized water reactors extending from              Boiling Water Reactors," and Regulatory Guide 1.4, and Including the secondary side of steam generators up            "Assumptions Used for Evaluating the Potential Radio- to and Including the outermost containment isolation              logical Consequences of a Loss of Coolant Accident for vulve, and connected piping of 2-1/2 inches or larger              Pressurized Water Reactors") that are more than 0.5 rem nominal pipe size up to and including the first valve              to the whole, body or its equivalent to any part of the (including a safety or relief valve) that is either normally      body.


==C. REGULATORY POSITION==
dosed or capable of automatic closure during all modes of normal reactor operation.
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:
q. The Class IE electric systems, including the g. Cooling water, component cooling, and auxil-              auxiliary systems for the onsite electric power supplies, iaty feedwater systems' or portions of these systems,             that provide the emergency electric power needed for including the intake structures, that are required for (1)         functioning of plant features included in items l.a emzerncy core cooling, (2) postaccident containment                 through Lp above.
  a.        the reactor coolant pressure boundary b.        the reactor core and reactor vessel internals c.        systems3 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.        systems2 or portions thereof that are required 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 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 reactors extending 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 operation g.       cooling water, component cooling, and auxiliary feedwater systems2 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 cooling h.        cooling water and seal water systems2 or portions thereof that are required for functioning of reactor coolant system components important to safety, such as reactor coolant pumps i.        systems2 or portions thereof that are required to supply fuel for emergency equipment j.        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
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
heat removal, (3) postaccident containment atmosphere cleanup, (4) residual heat removal from the reactor, or                2. Those portions of structures, systems, or compo-
(5) cooling the spent fuel storage pool.


k.        systems2 or portions thereof that are required for (1) monitoring and (2) actuating systems4 important to safety l.        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)
nents whose continued function is not required but whose failure could reduce the functioning of any plnat h. Cooling water and seal water systems' or                  feature included in items La through l.q above to an portions of these systems that are required for function-         unacceptable safety level should be designed and con- ing of reactor coolant system components important to               structed so that the SSE would not cause such failure.
  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,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 applicable q.        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 above===
safety, such as reactor coolant pumps.
2. 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.6
3. 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.
I. Systems' or portions of systems that are re-                  3. Seismic Category I design requuements should quired to supply fuel for emergency equipment.                      extend to the first seismic restraint beyond the defined boundaries. Those portions of structures, systems, or j. All electric and mechanical devices and circuitry          components that form interfaces between Seismic Cate- between the process and the input terminals of the                  gory I and non-Seismic Category I features should be actuator systems involved in gpnerating signals that                designed to Seismic Category I requirements.


5. Regulatory Guide 1.189, Fire Protection for Operating Nuclear Power Plants (Ref. 8), provides guidance used to establish the design requirements for portions of fire protection SSCs to meet the requirements of GDC 2, as they relate to designing those SSCs to withstand the effects of the SSE.
initiate protective acUon.


4 See the latest edition of Regulatory Guide 1.151, Instrument Sensing Lines (Ref. 4).
4. The pertinent quality assurance requirements of k. Systems' or portions of systems that are Appendix B to 10 CFR Part 50 should be applied to all required for (I) monitoring of systems important to activities affecting the safety-related functions of those safety and (2) actuation of systems important to safety.
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 4
portions of structures, systems, and components covered under Regulatory Positions 2 and 3 above.


==D. IMPLEMENTATION==
1. The spent fuel storage pool structure, including the fuel racks.
The purpose of this section is to provide information to applicants and licensees regarding the NRC staffs 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 NRCs 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.
*Lie indicate substantive changes from previous issue.


REGULATORY ANALYSIS / BACKFIT ANALYSIS
m. The reactivity control systems, e.g., control            'Wherever practical, structures and equipment whose failure rods, control rod drives, and boron injection system.                 could possibly cause such injuries should be relocated or separated to the extent required to eliminate this possibility.
        The regulatory analysis and backfit analysis for this regulatory guide are available in 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.


Rev. 4 of RG 1.29, Page 5
'Specific guidance on seismic requirements for radioactive waste
'See footnote 1, p. 1.29-1.                                          management systems is under developmen


REFERENCES
====t.     I====
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
                                                              1.29-2
2. U.S. Code of Federal Regulations, Title 10, Part 100, , Reactor Site Criteria, U.S. Nuclear Regulatory Commission, Washington, DC.7
3. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design-Basis Accidents at Nuclear Power Reactors, U.S. Nuclear Regulatory Commission, Washington, DC.8
4. Regulatory Guide 1.51, Instrument Sensing Lines, U.S. Nuclear Regulatory Commission, Washington, DC.8
5. Regulatory Guide 1.143, Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants, U.S. Nuclear Regulatory Commission, Washington, DC.8
6. 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.8
7. 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.8
8. Regulatory Guide 1.189, Fire Protection for Operating Nuclear Power Plants, U.S. Nuclear Regulatory Commission, Washington, DC.8
9. Draft Regulatory Guide DG-1156, Seismic Design Classification, U.S. Nuclear Regulatory Commission, Washington, DC, October 2006.9
7 All NRC regulations listed herein are available electronically through the Electronic Reading Room on the NRCs 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 NRCs Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDRs 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 NRCs 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.
"I                 


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 NRCs Public Document Room (PDR), which is located at 11555 Rockville Pike, Rockville, Maryland; the PDRs mailing address is USNRC PDR, Washington, DC 20555-000
==D. IMPLEMENTATION==
 
proposes an acceptable alternative method for comply- The purpose of this section is to provide information        ing with Tpecifled portions of the Commission's regula.
===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 NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html.
to applicants regarding the NRC staff's plans for using        tions, the method described herein is being and will this regulatory guide.                                          continue to be used in the evaluation of submittals for operating license or construction permit applications I    This guide reflects current NRC staff practice. There.


Copies are also available for inspection or copying for a fee from the NRCs Public Document Room (PDR), which is located at 11555 Rockville Pike, Rockville Maryland; the PDRs mailing address is USNRC PDR, Washington, DC
fore, except in those 'cases In 'which the applicant until this guide is revised as a result of suggestions from the public or additional staff review.
  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}}
1.29.3}}


{{RG-Nav}}
{{RG-Nav}}

Revision as of 00:14, 20 March 2020

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 Revision 2 February 1976 REGULATORY GUIDE

OFFICE OF STANDARDS DEVELOPMENT

REGULATORY GUIDE 129 SEISMIC DESIGN CLASSIFICATION

A. INTRODUCTION

nuclear power plants that should designed to with.

stand the effects of the SSE. J

A hL

General Design Criterion 2, "Design Bases for Protec- tion Against Natural Phenomena," of Appendix A, B. DISC*

"General Design Criteria for Nuclear Power Plants," to

10 CFR Part 50, "Licensing of Production and Utiliza- After reviewing a splqol plications for con- struction permits o ngj

'enses for boiling and tion Facilities," requires that nuclear power plant pressurized water c r plants, the NRC staff structures, systems, and components important to safety "gn classification system for has developed a be designed to withstand the effects of earthquakes identifying p ures that should be designed without loss of capability to perform their safety to withstan fec5 of the SSE. Those structumes, functions.

S ents that should be designed to if the 4ZqIF n-t-vc ho rp, n vt.ei .

Appendix B, "Quality Assurance Criteria for Nuclear .-.. 1.

Power Plants and Fuel Reprocessing Plants," to 10 CFR as~icLategory 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 apply to all activeit eing the safeqtu.imd and should be designed to withstand the effects of theI

applyofthose 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 featurh 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 ter.q di~ to

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connected piping up to and including the first valve (including a This guide describes an acceptable method of identi. safety or relief valve) that is either nornally closed or capable fying and classifying those features of light.water.cooled 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 %lellof 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 themrt not toqruied Methods and sOlutions ditferent from those tat Out on 2 Research and Tolt Reactors I 1tanspOrletDon the guidaes wil be acceptable J9they provide a bel fot the finding$ realusilt to 2 Fuels and Materals Facilities a Occupatiorel HMeath the *.suance or conulruunce of a Permi or ocent. by the Commission 4 fnroonmenttl aendSli.ng I AnttIuel Review Comment. and tuggesttuntt ofa,rmproomeflls .n that* guides are encouraged S Mterial& and Plant Protection 10 General at elf troeS and g;dmi wI be , a.led at sporopa g0oaccom* odlat caom manla end to *ettIctnew ,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- n. The control room, including its associated vital moval system). equipment, cooling systems for vital equipment, and life support systems, and any structures or equipment inside d. Systems' or portions of systems that are or outside of the control room whose failure could result requized for (1) reactor shutdown, (2) residual heat in incapacitating Injury to the occupants of the control

2 removal, or (3) cooling the spent fuel storage pool. room.

e. Those portions of the steam systems of boiling water reactors extending from the outermost contain- o. Primary and secondary reactor containment.

ment isolation valve up to but not including the turbine stop valve, and connected piping of 2-1/2 inches or p. Systems,' other than radioactive waste manage- larger nominal pipe size up to and including nhe first ment systems,3 not covered by itemns l.a through 1.o I $

valve that is either normally closed or capable of above that contain or may contain radioactive material automatic closure during all modes of normal reactor and whose postulated failure would result in consrva- operation. The turbine stop valve should be designed to tively calculated potential offsite doses (using mete- withstand the SSE and maintain its integrity. orology as prescribed by Regulatory Guide 1.3, "As- sumptions Used for Evaluating the Potential Radio- f. Those portions of the steam and feedwater logical Consequences of a Loss of Coolant Accident for systems of pressurized water reactors extending from Boiling Water Reactors," and Regulatory Guide 1.4, and Including the secondary side of steam generators up "Assumptions Used for Evaluating the Potential Radio- to and Including the outermost containment isolation logical Consequences of a Loss of Coolant Accident for vulve, and connected piping of 2-1/2 inches or larger Pressurized Water Reactors") that are more than 0.5 rem nominal pipe size up to and including the first valve to the whole, body or its equivalent to any part of the (including a safety or relief valve) that is either normally body.

dosed or capable of automatic closure during all modes of normal reactor operation.

q. The Class IE electric systems, including the g. Cooling water, component cooling, and auxil- auxiliary systems for the onsite electric power supplies, iaty feedwater systems' or portions of these systems, that provide the emergency electric power needed for including the intake structures, that are required for (1) functioning of plant features included in items l.a emzerncy core cooling, (2) postaccident containment through Lp above.

heat removal, (3) postaccident containment atmosphere cleanup, (4) residual heat removal from the reactor, or 2. Those portions of structures, systems, or compo-

(5) cooling the spent fuel storage pool.

nents whose continued function is not required but whose failure could reduce the functioning of any plnat h. Cooling water and seal water systems' or feature included in items La through l.q above to an portions of these systems that are required for function- unacceptable safety level should be designed and con- ing of reactor coolant system components important to structed so that the SSE would not cause such failure.

safety, such as reactor coolant pumps.

I. Systems' or portions of systems that are re- 3. Seismic Category I design requuements should quired to supply fuel for emergency equipment. extend to the first seismic restraint beyond the defined boundaries. Those portions of structures, systems, or j. All electric and mechanical devices and circuitry components that form interfaces between Seismic Cate- between the process and the input terminals of the gory I and non-Seismic Category I features should be actuator systems involved in gpnerating signals that designed to Seismic Category I requirements.

initiate protective acUon.

4. The pertinent quality assurance requirements of k. Systems' or portions of systems that are Appendix B to 10 CFR Part 50 should be applied to all required for (I) monitoring of systems important to activities affecting the safety-related functions of those safety and (2) actuation of systems important to safety.

portions of structures, systems, and components covered under Regulatory Positions 2 and 3 above.

1. The spent fuel storage pool structure, including the fuel racks.

  • Lie indicate substantive changes from previous issue.

m. The reactivity control systems, e.g., control 'Wherever practical, structures and equipment whose failure rods, control rod drives, and boron injection system. 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

'See footnote 1, p. 1.29-1. management systems is under developmen

t. I

1.29-2

"I

D. IMPLEMENTATION

proposes an acceptable alternative method for comply- The purpose of this section is to provide information ing with Tpecifled portions of the Commission's regula.

to applicants regarding the NRC staff's plans for using tions, the method described herein is being and will this regulatory guide. continue to be used in the evaluation of submittals for operating license or construction permit applications I This guide reflects current NRC staff practice. There.

fore, except in those 'cases In 'which the applicant until this guide is revised as a result of suggestions from the public or additional staff review.

1.29.3


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