ML20101G719

Rev 2 to Design Bases Spec 386HA931, Event Combinations & Acceptance Criteria

ML20101G719
Person / Time
Site:	05200001
Issue date:	07/29/1983
From:	Capozzo R, Garcia J, Love J GENERAL ELECTRIC CO.
To:
Shared Package
ML20101G709	List: ML20101G706 ML20101G719
References
386HA931, NUDOCS 9206260252
Download: ML20101G719 (72)

Text

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• NDCLE/.4 ENERST GENERAL ELECTfilC 3 86'tA831 s u o. 2 BU51NI5W OPE 4ATIONS

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TABLE OF CONTENTS

1. SCOPE 1.1 Prodset 1.ine Applicab.'Isty.

', 1.2 Decament Applicability 1.2.1 General Elactric Compacy 1.2.2 Ite Architect Engineer 1.3 Document Nonapplicability 1.4 Equipment and Cor;onent Aop11cability 3 .3 Building Structures Applicability

. 1.6 Eqcipmatt. Compo ne n t , and Structures Nonapplicability

2. APF1.ICABLE DOCUMEliTS g 2.1 Goneral Electric Documents 2.1.1 Supporting Docaments-2.1.2 Supplemental Documents 2.2 Codes and Statdsrds 2.2.1 Americas Society of Mschenical Engineers

.1. 2 .2 American f rom and Steel Institute 4AISI) 2.2.3 American 'astitste of Steel Construction (AISC) 2 .2 .4 American Concrete Institute (SCII Standards 2.3 Other Documents (Reference Only)

3. DETINITIONS AND ABBRiv!ATIONS 3 .1 Reactor Conditio*s 3.1.1 Normal 3 .1.2 Upset 3.1.3 Emerge..cy 3.. 4 Fau'.ted (Itaitins f aults)

J 1.3 Ineredible 3 .2 Load Cos-.itions f or concre te Containment Structures 3.2.1 Serrice Load conditiens 3.2.2 Fsetored Load Conditicas m e e.u inev. .i.n

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!NUCLIAR INERGY 386 2 31 s a o.

IUSINESS OPitATIONS GENER AL h ELECTRIC -

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4 TABLE OF CotiTDiTS (continned) 3.8 Stress Categories and Acceptaa:e Criteela I

3.8.1 S t re s s Ca t e g orie s - ASME EJ.PV Cod e S e c t i on III 3.8.2 Acceptance Criteria - ASME BAPV Code Section III 3 .t .3 Acceptance Criteria - AISC 3.8.4 Acceptance Critoria - ACI

, ' . ** Acceptance Criteria - AISI 3 .g .6 Asceptance Criteria - Aluminam Structures 3 .9 Ncmesclature 4 DESCRIPTION 4.1 General 4 .2 Appliestica 4.2.1 Design

, 4.2.2 Evaluatica

5. DESIGN RESPWSDILITY

6. EIQ~~ git 3 6 .1 Events and Event Combinationa

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6.1.1 Event Classifications 6.1.2 Event Coahinations 6 .2 Building Interactica 6.2.1 Seismic Interactions 6.2.2 Nonseismic Interactions 6.2.3 Interaction of Nonseismic Category I Structores with Seismic Category I Structnrse 6 .3 Load Phenomens 6.3.1 Safety Relief Valve Actuation Loads (SD) 6.3.2 Opera tional Transie:.t s Loads (OT) 6.3.3 Design Basia Accident Loads '(DBA) 4.3.4 Environmental and Site-Related Loads (ENV) 6.3,5 Selasic Loads 6 .3 ,G Testing Loans (TEST)

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1 4 NUCLEAR ENERGY 486 DA931 IUllNESS OPILATIONS G EN ER AL 4$[h ELECTRIC afv sH No. 3 2

TABLE OF CCNTDrfS (Continued) 3 .3 Events Used in Evaluating the Effect of Loading Ccebinations on Structures, Systems, and Components 3.3.1. Normal Operation (NU) 3.3.2 Operational Trar:1ents (OT) 3.3.3 Design Basis Accident (DDA) 3.3.4 Sei ssia Events 3.3.5 Enviror. mental and Site Event a (ENY) 3.3.6 Intrequent Operational 1raustent (IOT) 3.3.7 Non-IDCA Fault (N1.F) 3.3.8 Te st a 3.3.9 Reactor Building Y1 oration (RBV) 3 .4 System, Structure, and Com.sonent Claasification 3 . 4 .1 01assifications 3 .4 .2 Safety Fnaction 3.4.3 Safety Related System 3.4.4 Auxiliary System 3.4.5 Cooponent C1sesification 3.4.6 Component Sopport 3.5 Buildings and Structures 3 .5 .1 Reactor Batiding 3.5.2 Fuel Building 3.5.3 Anziliary Building 3 .5 .4 Radvaste Bnilding 3.5.5 Control Building 3.5.6 Diesel Generator Buildings 3 .5 .7 Secondary containment 3.6 Nuclear Island 3.6.1 Nuclear Island 3 .6 .2 Balance of Nucles? Island (BONI) 3.6.3 Nuclear Steam Supply Systis (NSSS) 3.6.4 Item 3 .7 Load Categories Used in Loading Combinations and Codes 3.7.1 Service Loads 3.7.2 Deatsn Lorda nro seta gnev., s e,es :

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4 1 A NUCLEAR ENERGY 386 RA931 sm.

l'J5lNE55 OPERATIONS G E N E R A L [h ELECTRIC REv 2 3

TABLE ?? CCNTDiTS (Continued) 6.4 Loads on Components 6.4.1 Cceponent Opersbilit) 6.4.2 Gen. oral Mechanical Component Loads 6 4.3 c:lectrical, Control, and Instrumentation Equipment Leads 6.5 Loads on Component Suppor.s 6.5.1 General Component Support Loads 6 .5 .2 Abnormal Loads in the Reactor Building 6.6 Loads on Structures 6.6.1 General Structural Loads 6.6.2 -Abnorzel Plant Loads on Structures 6.6.3 Loads on Containment Structuree 6 .7 Evaluating Load Combinations i f .7 .1 Initial Conditions 6.7.2 Combining Loads 6.3 Loading Loubinations and Acceptatco Criteria 6.8.1 Hechanical Components 6 . 8 .2 Electrical, Control, and Justrumentatiota Components 6.8.3 Cc=;cnect S ;;ert s G .8.4 Structures 6 .3.5 Raptured Pine Critorie so ss as-ev i ....

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' NUCLEAR ENERGY '""^ w ~o- '

EUSINE55 OPERATIONS G E N E R AL 'h ELE CTRIC nty 2 TABLE OF CMTS (Continned)

List of Tables 1 Basic Event Combinations 2.1 throssb 2.10 Loading Combinations and Acceptance Criteria 2.1(a) C!sse 1 Xechanicel Component e and Core Support htructurce 2.1(b) Cisse 2 and 3 Nechanical Components 2.2 Non-ASME Mechanical Componente & Elgetrical Centro! &

Instrumentation Equipment - Working Strass Method 2.3 Non-ASME Mechanical toeponents 4 Electrical Control &

Instrumentation Equipment - Strength Design Nothods 2.4 Deleted 2.5 Deleted 2.6 ASME Sectic'n III Compouent Supports 2.7 Concrete Containannt 2.8 Stems Containment 2.9 Concrete Structuress Other than Containment.

2.10 Steel Structures Othat th>in Containment 2.11 Load Combinettons and Acceptance Criteria Refueling Equipment 3 SRV Astnation Loads for Various Events 4.1 through 4.5 Loading Conditions vs Location uno en ainav.se, o

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t 4 NUCLIAR ENERGY 38 m S31 $m 7 RU51 NESS QPERATION5 GEN ER AL h ELECTRIC arv g

1. SCOPE 1.1 Prednet 1. in e App l i c a b i_l i ty, This decument specifies loading combinations and sintaus acceptance criteria f or the de sign of all evv and the evatustion of all- existing saf ety related essential equirment, cea pone nt s, and structures for BVR 4, 5, and 6 NSSS and Eark II sud III Conteicrent s and e4 socia ted Nuclear Island for the vsrious events occurring during the lif e of the plant.

1.2 Doc ume n t Applicability 1.2 .1 General Electric Comyanya The requirements of this document are manda tory f or eg ottaent c ompo r.e n t s , and structures suppl.ed by or to the General Electric Company Nuclear Energy Business Operations for the Nuclear Steam Supply System (NSSS) and the Balance of the Nuclear Island (BONI) as defined in Paragraph 1.1.

The document provides the loading combinations due to verloos events considerius loads upon stro6tures ir. areas of the Nuclear Island, and also groups of components located withis these areas and/or attached to these structures. It also provide a acceptance criteria to be still ed to determine the acceptability of theme loading combinations and the ability of this eq ui pme n t, components and stenctures to maintain their structural integrity ander the conlitions delined in this document.

.2 .2 Tse *rahi+--e E=qla-a . The *egisteements of thi s document are recommendations f or the Archttect Engineer escept where the. e is an interface with Generst Electric supplied equipment, c<eponents, and structures covered by this docur.ent. For these intcrface cases, the requirement s of this document are mandatory.

1.3 Document "oneneticability. Consideration of environmental conditions is not within the scope of this document. Evaluation of the stresses in the pipe containing s postulated line break and the pipe whip restraints are not withiu the scope of' this document.

1.4 Equipment and Component Applicability , The requirement s of this document apply to all cquipment and components which are classified as Soltaic C> tegory I and/or Safety Class 1, 2, or 3 in design docascatation.

1.5 Buildien Structures Applicabilits. The requirement e of this document apply -to saf ety related Sestmic Category I Building Structures, via the Reactcr Building, Fuel Building, Auxiliary Building, Control buildin,1, Die sel Generator Buildings (2), and the Radwaste Dutiding, af04074lptw t ele t )

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LUltNESS OPERATIONS 0EN ER AL h ELECTRIC nev 2 1.6 F4uiresne Component, and St ruc ture s Nosappligbility. The req ui rement.

of this document do not apply t o Non s af e ty 8.l a s s 1, 2, and J equ.pment e and ccaponents nos to Nonseismic Category I sts uctures entess f ailure of such eq ui pme n t , component or structure during en evaluation basis event would cause f ailure of or impair the required perf orvance of the saf ety role.ted equipment, componont and structure s. The requireseat s f or the Turbine Dailding and all systems, structures, and components wit'41n the Turbine Building are not within the scope of this document,

2. APPLICADLE DOCUMEh!3 2 .1 Ceneral Electric Documents. The f ollowing document s f orm a part of this specification to the eatent specified herein, 2.1.1 Supportina Docualats. Documents under the following identities are to be used in conjunction with this specification.

Reference D e s i n o s t o.r_

s. Containment Loads Report (CLR), Mark III Concsinspent M1/ A42-5 400

b. Artsagement, Reactor. Fuel, and Auxiliary A21/ A22-20 SO

c. Arrenge=ent, Ans 414 ary Building A21/ A22-2080

d. Arrangement, Fuel Building A21/ A22-2080

e. Arrangement, Control 3nilding A21/ A22-2080

f. Arrangement, Radwaste Building A21/ A22-2080

g. Ref erence Containment Definition A41/ A42-$ 170

h. MtC Regals tions laplementa tion ' A41/ A4 2-207 0 .

i. Regu14 r.ry Guide Impleaentation Positions A41/ A42-4070 J. Reactvr Cycles Ut 3-3 040

k. Product Safety Standards A41/ 42-4070 2.2 Codes and Stenda rds. The f ollowing code s and standards -(issue in effect at .he placceent dato of the purchase order or e s sta ted in this specifica tiou or its supporting document s) fera a part af this specification to the extent specified herein.

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NUCLEAR INER6Y 3 86 EW WU5tNil5 OPERAfloHS GENER AL $ 5LECTRIC arv sm 9 2

s.

2 .7 .1 Aserican Society of Mechenten! Foriaeers ASMF) Boiler and Pretsure Vessel Code

a. Sectica III, Diviolon 1 - Rules f or Construction of Nuclear Power Plaut Compon e o t s.

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b.Section I?I. Division 2 - Code for Concrete Reactor Vessels and Containments. J 2.2.2 _Aa_ofscan Iron _rnd ? teel Institute ( AISI)

e. Etainless Steel Cold-Fersed Structueel Design Manual 1974 Edition 2.2.3 American Institute of Beel Con struc t ion -f AISC)

a. Specificitiocs for the Design, Fabrication and Erection of Structural Steel for Soildings.

2.2.4 - Am e r i c a n on c r e t e Institute ( ACJ) Standards

s. 318-71, Buildtog Code Requirement s f or Reinf orced Concrets

b. 349-76, Code Requirements for Nuclear Safety Related Structuras 2.2.5 J o s_t.i t u t e_o_f_El e c t r i c abad_E l e c t ron i c, E_.n g,1,ne eg e _(I,EEQ

a. 279-1971. St.*dard Criteria for Protection Systems for Nuclear Power Generating Stae'ons

b. 308-1978, Standard Criteria ter Class 1E Power Systems fer Nuclear Power Generating Stations-

c. 379-1977, Standard-Application of the Single Failure Criterion to Nuclear Pvver Generating Station Class 1E Systems 2.2.6 hoerican National Standards Innitute ( ANSI)

a. N45.'t .11. Qt.ality Aa entance Req uirement s for the De sign of Naclent Power Pl a n t s , 1973

b. N176, Design Basis f or Protection of Nuclear Powe't Planta Agalast Effett s of Postulated Pipe Ruptue

~eo 4es inev. se,en I

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NUCLSAR ENERGY 386nm1 s a.o. 20 BU$1HE15 OPERATIONS GENER AL h ELECTRIC ngy 2-2 .2 .7 Aleminua As socia t ion ( AA)

a. Maninum Standards and Data,1979 tr . Maninsa Construction Hanaal (1) Section 1, Spectfication f or Muminue. Structures, 3rd Edittoa,1976 (2) Section 3, Engineering Data for AJ uninum Structures, 3rd Edition, January 1975 2.2.3 Code of Federet Renalstions (USQQ .

a. NUREG 04 &4, Ne thodology f or Combining Dynamic Respcasse

3. DEFINITIONS AND ABBREVIATIONS 3 .1 Plantfonditinns. The state of the reactor sad its associeted s t ruc t ure s, sy stems, and component s. M ao, see Paragraph 3.5.2.2.7. The event enconator probabilities f or these conditiosa are as ste.ted in the t

product saf e ty standards, ref e ence Paragraph 2.1.1.k.

3.1.1 Norust. Normal creditions are any conditione la the course of system startup, operation in t'ae design power range, scrual be,t standby with the main condenser available and system shutdown. other than opset, emergency, fantted, incredible, or testing conditions. Hot stu 4by wittost the mata condensst is an upse t candittoa.

3.1.2 Upset (iccidents of moderate erobabilitt of occurence). Amy deviattors f raes normal conditions anticipstod to occur ca ten enough that destga should include a capability to withstand the condi'. ions without ope r a tional imp a i rm e n t. The upset conditinos include those *ransients which result f rom any single operator error or control asifnaction, treasients caused by a f ault in a system component reqstring its l',olation f rom the systes, and transients due to loss of Icad or power. Upset conditions include any f. ;rmal incidents not resulting la a forced nutage, and also f orced outages for which the -

corrective action does not inc1rde any repeir of machemical damage to the -

primary system pressare bonadery.

3.1.3 F_sermency (inf rea sr&ce ndent s) . Those deviations frem norzel conditions which may rec, aire shutdown f or correction of the conditions or repair of damage in t!'e syste.s. The conditions have a low probability of occurrence but are lacladed to provide assurance that no gross loss of

, structural integrity will~ result as a concomitant effect of any damage developed in'the system.

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a j J .1.4 F a n t t e.d ( l i m i t i ng f ault sl Tho se combina tions of conditions a ssocia ted wit h e s t r s mely-Icw-pr ob a b s l i ty, p.st ula ted event s whose conseq uence a are such that the integrity and cretability of the nuclear energy system may be tepaired to the extent that considsrations of public health and safety are involved, i

3.1.5 fccredible. A deviation f rom normal conditions which has a such a 19,

, probability of occurrence that it need not be considered in the de sign.

l H ow e v e r , some exceptional evert s with such low probabilities are postula ed as de sign and ev alua tion ba se s by Regula t ory requireme nt s. These occurrence s shall be conssdered in the design.

3.2 1.oud Conditions foe Concrete Containment Structures i

3.2.1 Service Lead Co,nJitions. ConJitions encountered daring construct son a e.J 4.- the normal operation of the plant, socluding any anticipated transient on test conditions 4aring notaal startup and shutdown of the nuclear steam supp.,. ;af e ty rela ted, and mutille ry sy s t ems. Also included are those severe environmental croditions which may be naticipated during the life of the facility.

3.2.2 Factored 1.osd Conditions. Load combinations, including unitipliers 4

ba sed on conditions resulting f rom a postulated pipe bresh ir. the reactor primary coolant system or environmental conditions postulated as upper bound

, limits for the plant site. Also included are lead combinetsons, with other ruitipliers. or a postulated pipe break in the reactor primary coolant sy s t em plus severe or estreme environmental conditicas.

4

3 .3 E* ent s Used in Evaluating the Fffect of Loadina Cochinations on Structures. Systems. 6te Cear onen t s a

3.3.1 No rm a l Do e r s t i on ( 50) . Optration under any sondition permitted by the reactor techhscal specifications (as delineated in the plant-specific pS AR/FSARI and centrol systems, irrespective of the anticipated frequency of occurrence of that condition, which is plauned and deliberate but not is I

specif ic response to ope ra tor . error s, contrcl malfunctioss, component failure, or tr ansients das to loss of load or pewsr. hormal operation sneludes leads due to weight, temperature, prestress, pressure, fluid flow, and other loads due to soving parts within a cerponent or my s t es.

3 .3 .2 Operational Treentents ( OT) . Plant response s t o spontaneous single eq u n pue nt- t a tius e s or to ssuple oretator errors that can reasonably be carected during any norvel.or planned mode of plant oper:tions. For purposes of_ applying thi* specitscation, the range of operat' onal transiect s postulated and analyzed'in the plant's Preltainary or Final hafety Analyess Repart shall be constJereo in the evaluation. Operettonal trenssent event loads include suth events as safety rettet valve actuation, turosce st op valve closure, rapid v alve motion.

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• NUCLEAR ENERGY m"3 2 = 2:

BU51NEin CPERATION5 GENER AL @ ELECTRIC arv 2 3.3,3 Loss of Coolant Accident (1.0C3 A pestela ted acciden. even s f ree which the plant de sign ba se s ar e e stablished. Instuding both transient acd q ua si-s teady s ta te coedi tior.s re sulting f rce the LOCA, or other faulted conditions.

3.3.3.1 Destro Resis Accideet (DB A) . De transient depeessurization a ssocia ted with th= suduen, vouble-ended sever nce of a recircula tion line, or of a main steam line inside or out side the c antainment, or army class of pipe severence of equivalent f l ow cros s-se c tional area. For OSA and IBA, the s;ecific break 10 cations sust be postula ted a nd the leca tion-uniq ue con s eq ue n c e s such as je t ta pin g eme n t are constdered as design loads in the Code stress repos:. ne t e rn ir.a l e cd p i ;e b r e a k i s a t the veseet safe end-tept pe veld J oint. Vessel or saf e $nd dreaks are not considered. DB\

event loads include loads such as annitus pressuriz ation, went clearing, pool swell, condensa tion oscillation, and chugliBL.

3.3.3.2 In,,t e rw e d i a t e B r e e k Accident (ICAL This classification covers those breaks f or whics operation of the Emergency Core Cooling System (ECCL) w tal occur during the blowuva and which result in reactor depressurinetton, ne inventory ef fect during the blevdvin is accos ;ted f or with an ef f ective break area of a 0.1 ft break below Lee resetor s essel water level. nis break en is cho sen as being repre senta tive ct' che totermediste break area range. These bsens can involve either reactor 6: em or liquid blevdown.

3 .3 .3 .? N i! P*en * ~ u c e ty w a.) , Th, ,ites of primery ** stem bl a d e nt in this cates ry are tcose bitudowna enich till cot result in tapad reactor Jepressurization due to either loss of reactor fluid or attuan tic opera t.o1 cf t h e ECCS eq ui pm e nt . Follow nes the occui rence ci a breek of this siz e, 'ho reacter operators should initiate an orderly shutdown and depressuriza ttun of the plant. Specific break locaticas are not postulated. Inadvertent Automa tic Depres suriza tion hy stem ( ADS) actuation is included in the SBA category.

3.3.3.4 j'_o_S_t Accident Conditions. De qua si-steady state condit!ons which follow a LOCA o.' othe r t a ni t e d transient conditions.

3.3.4 Seismic Event:

3.3.4.1 S a .* e St's tdown Fo r t hqu a k e ( W) . nat *srtiquake which is ba sed upon no cvalaatton of the assinua cartr4uate potential constdering the regs(nal s ud

!ccal leulogy and seismology and s;ectf it, charauteristics ol. local subsurface caterial. It is that'eart!4 cake vtach t roduces the assimum vabretory ground motion for which certain structures, systems, and componeht s a re du signed t o remain f unc t ional, t

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NUCLEAR INERGY 386R^831 BU5173855 OrgtATIONS G E N E R A L 4)hh ELECTRIC REV g 5 ~v 23

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3.3.4.2 _ Ope ra t ina Da s l > Ea rt houcke (OPE) . The earthquake which, conssdering l

the regional and local geology and seismolcgy and specific characteristics of local subsurface eeterials, could reasonably he espected to affect the plant duries the operatiss life of the plant; it is that earthquake which produce s the <ibratory ground motion for wht:b those features of the nuclear pcwer plant necessary for continued operation without undas risk to the health and saf e ty of the public are de signed to rewata f unc tional. The 09E includes ths displacement limited (OBE y ) and the inertia laed (CBEg ) portions.

3 .3 .$ Envirotmental and Site Fvents ( ENV) 3.3.5.1 Desias Wind (W). The maalare de sign wind established for the plant site.

3.3.5.2 Desian Basis Tornsde (W') . The design basis tornado e stat *.ished f or the plant site, including the effects of missile impact.

3.3.5.3 Probable Maximum Flood ( PMF) . The de sign basis flood is that flood that nuclear power plant s should be designed to withstand without loss of capability for cold shardown and maintenance thereof. This load is considered as a static load on the loading crobination tables.

3.3.6 Infreene,t Operational Transient ( 107) . Any non-LOCA event which has a sufficiently low eccounter prob 6bility to be co.sidered as an emergency event, infesquent operr tional transient event 1 cads snclude fuel cask krop, fer example. .

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. 3 .7 Non-LOC A Fa ul t ' NTS) . Any non-LOCA event which has a sufficiently low encoa .er probability tv de considered as a faulted event.

, 3.3.8 Tests 2 Off-norest operating conditions imposed during preoperational-te sting or hy the opera t or. Test loads include hydrostatic and pneumatic pressors and weigFt effects.

3.3.9 Reac tor Buildint Vibrat ion ( RBV)2 Acceleration and displacement of the

• Reactor Duilding structures caused by seismic, accident (DBA), environmental events (wiud, tornado), and operational transient s (SRV,1SYC) .-.

3 .4 System. Structure, and Component Classification 3.4.1 CIansifications 3.4.1.1 Safety Classe1, Structures, systens, and component s are - c12 ssif iec at Safety Class 1, Saf ety Class 2. Safety Cisss 3, or Other in accordaace with the irportance of the safety functio.,s to be performed by such equipaent.

Eq u i pme n t is assigned a specific safety class. recognizing that s tecone t.t s within a system may be o t' differing safety importance. -A single system may thus have creponent s in more than one safety class. In addition, ene piece Of eq uipm e n t can have multiple safety functivas of differing safety classes.

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'wucuAR ENfBGY GEHERAL h ELECTRIC 3 "

• 22 5~ u IUSINESS OPCKATIONS arv 2 3.4.1.2 Seismic CletLjliceLica, Itcss structures, systeas, and components trportant to saf ety that are de signed to remain f unctional in the event of an e ar thquake are de signa ted 31 5 ~a 15 BUSINill OPERATIONS G E N E R A L J$$$ E LECTDic arv 2 3.4.5.1.4 Class kit Itese ccastractas in accords 6ee with the requirements of the ASME Bos ter and Pressere Vessel Code, Section 111, Subsection NE.

• 3 .4 . 5 .1. $ CJs W 1 Items constructed in accordance with the regstrement s of the ASKE Boiler asid Pre ; sere Ye ssel Code,Section I!!. Subsection NO.

3 .4 .5 .2 Con r_qe e n t s ( n o n- ASME),a I t ns from which a system is assembled, anch as, sensors, signal conditiones s. motors, pumps, valves, pipis g, best exchangers, and other devices.

3.4.f .3 Aettve Coevorent. Any component in which a change in sta te, measurement, or mechanical notion is needed to perf orm an autonssac saf ety function such as saf e shutdown of the reactor or mitigation of the consequences of a postulated pipe break in the Reactor Coolant Pressure B ound a ry ,

3.4 .$ .4 Pasalve Cearonent. A device chatscterited by an espected negligible change of state or cepilgible mechanical motion in response to en imposed dettgt basis load demand upon the system.

3.4.6 Ceesonent Serrort. Ta < e metal cirment s which transmit loads or provide a general support inaction between the nuclear power plant cespouent and the building strteture. The se element s are constracteJ in scsordance with 14e requirreents of ASHE Boiler and Pressure Yessel Code,Section III, bob s o r t 4 on he , or .'d >L, or Aadt.

3.3 guildiera sad Sttuctures 3.$ .3 Reactor Buildini , Jtructural comptes composes' of the drywell. Setwell, con t a i nme nt , and shield boilding. In the Mark 11 the resctor building define a the bosadary of the secons.ry contsi; ment.

3 .3 .1.1 Containment Pool Stracture. The structure surronnoing the pool of veter, located diaestly aboee the reactor, providing radiation shielding Juring power operation and refuelish, specifically as applied to ;he Mark III, 3 .3 .1.2 D_yve11. 1he structure surrounding the reactor and its recirculation loops which will channel steam resulting tram the LOCA through the suppression pool for condensatica. In the Mark II 1". la also a part of the conteir. ment

( ,. boundary, 1.$ .1.3 containment. The gas tight shell or other enclosure arouni e reactor (pre ssure ve ssel) to confit.o products tha t otherwise might be released to the atmosphere in the event of an accident, wroes.minew.seen i

NUCLEAR ENItGY 3 88 8A"1 BUSINES: CPERATIONS 0 E N ER A LM E LECTRICmay $~ 25 2

3 .$ .1.4 Socoression Pool. A structure surrounding the pool of water, ,loca ted lostdo the bass of the ccatainment, which providas the veter stel between the dryvell s.id the contait.3ent. Duritig safety relief valve discharge and postolated LOCAs, the pool serves as a best sink, and the structure serves as a pressure esppression shaa.ber. For the Mark 11 containment, the suppression pool is in the wetwe!! whsch with the drytell forse the containment.

3.3.1.5 Shield Buildina. n e reinforced concrete structure whicA incloses the drywell and containeont. Together with portions of the Antis ery Buildint and the Fuel Bn!! ding it forme a secondary contairment in the Mark 111 conf ig 4r a ti on.

l 3.5.2 Fuel Dutidina. Structural sc'aplet 6vataining the spent fuel storage pool.

3.$ .3 Anellia m 9aildina. Structural comptes containing major saf ety reisted sy s t ema s, 3.$ .4 Redweste Build!na. Structural cooptes containing redesete treatment systems.

3.5.$ Control Duttdina. Structural comples containing tas Control lloca.

3 . .* . 6 Diesel Generator Buildinas. Structural compleses containing the diesel g us t a t ur -i r...

3 .3 .7 Seconderr containment. The Shield Building Fuel Building, and those portions of the Ansiliary Duilding enclosing the ECCS. ne second barrier that ensloses the reactor.

3.6 Nuclear Island  ;

3.6.1 Nuclear Island. lie buildings, internal structuret, enclosed systens, and e;siement located within the Reactor Building, Fuel Building, Ausiliary Duildit ,, Centrol Building, Radcaste Building, and Diesel Generator Bu11 dings.

, , 3.6.2 Dalence of Nuclear Island (r$ L The conglomerate of systems, structures, and components ohtch, tegether with the Nuctsar Steam Supply System (NS$s) supplied by General Llectric Nuclear Energy Dosiness Crerstions, make up the Nuclear 1 stead.

3.6.3 Noelear Steam Supply Svs tem (NSSS) . The s) stems, structuras, and component s which genera te anJ nupply steam to the t alsace of plant.

=to seen inev. won l

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NUCLEAR ENf tGY sism1 am iv BUSINESS OFIRATIONT. GENER AL $ ELECTRIC arv 2 3 .7 , Load Catetories Used in Leadiet Codine t tons sed Codes 3 .7 .1 Service ! peds

s. Fechanical Load. Mechanic 61 loads are those loads leposed ud a system or component which cause load controlled stresses,

b. Sveteleed Lond. $nsteined loads are tbose loads which ocent for an estended length of time on a caponent,

c. Occasional t.osd. Occasional loads are those loads which occer daring normal sad abnormal opera tional pha se s of plant opera tion.

d. Toertict Lond. Inertial compocent of a dynaals load is the portion of the load which causes load controlled stresses.

e. p[splacement Limited Load. Displacement lialted component of a load is the portion which causes secondary stresees which are elf limit!,ng.

Example s of displacement limited Icads are eachor movenients (eg. t h e rriel and earthquake) of other items and thermal espanoton of she ites being considered.

f. Prestress Load. Prestress load is sa initial load imposed on a portior, of 6 cmponent or structure, usually to canse as laitini state of stress tu

, !Le a.c !.c. 4.t '.h nt .onditicas.

g. Total twnasio Leed s Total dynamic load is the aan of the inertist portion and displacenent limited portion of a load.

h. Drnesio Los*, Any time varying load application.

3.7.2 Deeins Londh. Design loadissa for Claee 1 couponeate and supports shall be as defired la NB-3112 and NF-3112 of the A5h2 Code. Design inadtags for Clas se s 2 and 3 cmponent s and Classe s 2. 3. NC. and CS support s are tho se pressures, temperatures, and asshanical loads selected as the basis for the itess. Destga loadings for Class MC vessels are defined in NS-3112 of the ASME Code. De siga loadings f or Claes CS structures are defined in hG-3112 of the ASME Code.

a. p,e s tan Pre s ente (Pph ne srecified internal and esternal design pressure shall not be less than the maximum dif ference la pressure between the inside and cutside of the component, or between any tuo chembers of a combination unit, which esists under the most severe loadings f or wLich the Level A service limits s4e applicable, ne design pressure shall include allowance s f or pre s sure surge s.

e t o .o r a (s e w, e s,se g m_____m_-____.____- _ _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

_ - - . - . - -- .. . . - - - - .. = - - - - - . . .

jl NUCl, EAR INERGY GENERAL' ELECTRIC 3 86 BA931 soo, 11:

BUSINE55 OPERATIONS -

arv 2 3.7.2 (ContinueJ)

b. De s ian _Teef e e ra tu re dD) . Except as otherwise defleed in NB-3112 for Class 1 samponents, the specified design tear tuse shall not be less than the espected aantaus mean metal tesiperature through the thicLness of i

the part considered f er which Level A limit s are specified. The 1 temperature so de signated may not cover the f ull range cf tsap.asieres f or which the component must be de signed.

c. De s lan We c h sgo s i f,o s d s ( b) . The specified design methanical loads shall l be selected so that ubes stat taed with the effests of desist. pressure,

they represent the most severe coincident loadings for which the Level A service limit s on primary stre ss are applicable. Note that it any be inadequate to consider only the design losf.ings sad the !.evel A service limits for initi nl design.

3.8 Stress Categorh_s... sed Acceptance Criterie 3.8.1 S t re s s Ca t e s o r i $ s - AWE 't& PV Cod e Se c t i on III 3 . 8 .1.1 freension Stresses bL Espansion stresses are those stresses resulting frca restraint of freic and displacenest.

3.8.1.2 Loe s! Prima rrymbrane St ress ( f.) . Cases arise in which a membrane stres: prods cd by pressure or Otter -tehatical lis414 and assoristti with s priesty or a discontinuity effect proacces escessive distortion in the.

transfer of load to other portione of the structure. Couserva t t aa req uire s that such 4 stress be cla ssified as a local primary membrane stre s: even thoosh it has some characteristics o.' a secondary strees.

3 . 8 .1.3 Peak Stress (#' Irb Peak stress is that increment of stress which I

is additive to the primary plus secondsry stresses by reason of local discontianities or locs! thermal stress, includius the effects, if suy, et stress concentrations.

3.8.1.4 Primary Bendian Street (Pit ). Bending strest is the variable ccuponent of normal stre ss, io, the strove which varies from the everage value with location across the thickness.

3 .E .1.5 Primary Membrane Stru s ( m) . Primary membrose stress is the ccuponent of corani stress .whica is unif ormly distributed and equal to the average vaine of stress across the thickness of the statica under consideration.

3.8.1.6 S e c ot.t a ry S t re s s ...(Q) . Self-equilibrating stress nece ssary to satisfy continuity of structure, aco seu msv. isisu 4

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NUCLEAA 8N886Y 3 861tA'81 5"o-8U5!N85$ OPERATIONS C EN ER AL h ELECTRIC mev 2 1'

3.8.1.7 Stres s intensitn ne dif f erence betese the sitebraiselly largest priusipal strass and tlc sigebraically ses!!est principal stress a t a polet.

with tensile stresses being ronsidered positive and scopressive stresses being considered esgative.

3 . 3 .1. 8 Load Contral'ed Stresses. Load controlled stresses are the strusses resniting from application of a loading, such as internal pressure, inertial loads, or the ef f ect s of gravity, vbose magnitude is not reduced as a Jes.st t of displacessat.

3 . 8 .1.9 ne rwel St r e s s , normal stress is a self-balameing s t re s s prodt.ced by a nonunif orm distribution of temperature or by differing thermal coefficients of espansion. Thermal stress is developad in an ites whenever it is-prevented from assuming the site sud shape that it normally shon1d unter a change in temperatore.

3.8.2 Acceptance _ Criteria - ASMr. BAPV Code Section !!!

. 3 .5 .2 .1 Desimo limits. The limit s f or design loadings.

3.8.2.2 Se rvic e Limit s. The De sign Specifies tlou shall designate service limit s f or nuclear pcuer plant components as defined bslow and as additions!!y defined in the appropriate Subsection of the Code.

3.8.2.2.1 te"e1 ' S "v8e* Li-ite. 1. . v ' .*

• t *

• 8 " 14 8** t-* %- " " *

• a '

• 11 sits which aust be satisfied f or all loadinf s identifieJ in the Design Specification to which the component or s ppon t may be subjected in the performsace of its specified sorrice f unction,.

3.8.2.2.2 Level D Se rv iu 8. !mit s. Level B 4ervice limit s are those set s of limits which must be se tastied f or all loadings ineatified in tha Design Specifications f or which these service limit s are designated.

3.8.2.2.3 level C Service 1.imits. Level C service limits are those sets of 11mit s which anst be satisfied 1o. all loadings identified in the Design Specifications for which these servico Itaits ase designated.

3.8.2.2.4 level D SJ rvice Liaim Level D service limits are those sete of limits which must to satisfied for s!! loedtags idtattfied in the Desian Spccifications for which the se service limit s are designated.

3.8.2.2.$ Testina 1,imits. Testing limits are those seta of lialts viict must be- satis!!sd durlag te sting of a component or system.

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' NUCt. TAR 8HER6Y 385 2 31 sa *. 20 SU51N855 OP8RATICM8 GEN ER AL @ ELECTRIC afv 2 3 . 8 .2 .2 .6 Alternative Servics Limits. Cca.ponent s or snpport s may be de signed using more restrictive service 1smits than specified in the design specification.

3.8.2.2.7 p .systent Nesenelature. For purpose of this document and for cooperips the nomenclature in this document with that which any appest in other General Electric documente it sha!! be naderstood that the following are eq uiv ale nt s llormel Limite = ASWT.Section III Service Level A l

Upse t Limits = ASME Section III Service Level B )

i hersency Limits = AsisSection III Service Level C  ;

Faulted Limit s = ASME Section III Service Level D This statement of equivalency does not apply to Table 2.d. ASME Service Level definitions applied to steel containment structures .if f or f rce the definitions applied to other ASHE Code component Service Leve14 3 . 8 .2 .3 A!;ovable Stress (s). Strese li-A t f or Class 2 and 3 component s and component suppos ts as specified in Ap;eadia I of the ASME Code, also stress limit f or cca ne te cont air.aent nr..sce ! cad. The allovable atte as sinit tplyirs factors shall be considered se rm uire, by tle appropriate acetica of the Aal f.

Code, 3 . 8 .2 .4 Dasien stress Intensity '3). e Design stress intensity velbes far Cl s: 1. Class CS, and Class kc components specifloo in Appendia I of the AME Coos.

3.8.2.5 Allowable Stress (U). Stress limit for concrete containment f actored load cambsna tion.

3.8.3 Acceptance Criterle - ATSJ 3.8.3.1 Alle,weble Elastic worklen Stress (S) . Allovohle stress limits as specified in Part 1 of t'. a AISC Specification for the Dess$n, f abrication and Erection of Structural Steel for Buildings. The appropriato nut:1pl31sg f actors are gives in the app!! cable tables et the end ci this document.

3.8.3.2 Alloveble Plastic Des tan Methode Lisit (Y) . Allowable limits e e specified tu Part 2 of the AISC Soecification f or the Design Fabrication sad Erection of Structurst Steel for Daildings. 11:e appropriate multiplying

, f actors are given in the applicable tables et the end of this doct.sent, i

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3.f.4 Acceptance Celterie - AC{

3 . 8 . 4 .1 A_11owable. Stresses per Aci-318-71 (t]. Allowable strees limits as specified in ACI-318-71 Duilding Code Eequirement s f or Reinforced Concrete.

3.3.4.2 Allovsble St re s s e e pe r ACI-34 9-76 (U) . Allowable stress limits as specified in AC1-34 9-76 Code Requirement s f or Nuclest Saf ety Related Concrete Structurse.

3.8.$ Accertsace Crtterte -

AISI 3 . 8 . 3 .1 Allowable Stresa 1,inits (S). Allowable stress limits as specified in Part 1 ef the AISI Specification for the Design of rold-Formed Steintese Steel Struct ural Member s. The orproptss te multiplying f actors are st een in the -

appsopriate tablee at the and of this document.

3.8.6 Accept euJe Criterie - Aluminum. St ructore s. Allowable stress 11msts se specified in Section 1 of the Aluminum Conattuction Manosl of the Aluminum Association. Al un icum structures located in DONI shon1d be designed according to the abeee specifications.

4 J .9 Nomenetature AA - Alumistum As socia tion ADS - Automatic Depressorisation System

' AISC - /.merican Institute of Steel Construction AISI - Aaerican Irca and Steet Institute AP - Annulus Pre s suriz a tic.n ASME Aenerican Society of Mechanical Engineers ATYS - Anticipated Transient without Scram B - Uplif t Forces on Buildinge BONI - Balancu of the Naclest Island CDF - Cumulative Distribution Function CUUG - Chugging Load CO - Condensation Oscillation CP - Cumpar tment Pres surisa tion 0 - Deadioed Including Construction Loads DF, - Dseparass Floor Reaction Load due to differential pressure between dryvn11 and vetwell for Mark 11 D

R - Dres Force on Item D

t

- Deadload for Test Condition D, - Deadloed of Water DBA - Design Basis Accident us o essa mev. ie,en I

- _ _ . - __ _ . - - - - - . ._ _ - _ . - . - - - - _- - - -. - .- . . - . - ~ _ . . .

' NUCLEAN IMIRGY 8"E^881 m o.  ::

IUSINill OPERAT60k5 GENER AL h ELECTRIC arv 2 3.9 (Continued) r.- . Eavfronssatal and Site Evente P - Pesk Stress or Aluminum Allowable Scress Limit Fg - Arternal Pressure f ru Flooding FIV - T10w Indeced Vibration F p3 - Pre stre ss Load MLS Floor Response Spectra F, - Friction Load R - Earth Pres sare

! - Impass Load -

l IBA - Int ersei.a a te Bre ak Accident IBL - latermediate Break Aecident Leading IEEE - Institute of Electrical and Electronics Essiseers 10T - lafrequent Operating Transient - Emergen:y Event L - Liv e Load IRA - Large Break Accident IEL - Large Break Accident Loading L3CA - Loss of Coolant Accident

. N - Nosmal Load fossisting of Pressore, Dead Tsight, and Finid Reactina Loads NEP - Nonetceedence Probability MS - Non-LOCA Faul t NO - Normal Operation (E E - Operating Basis Earttuttake (EE - (BE-Displacement Limited Portion D

ODEg - (SFr Inertia Load Portion OT - - Operational Transtants - Upset Event P - Pre s sure P, - Accident Pressure dos to DBA P

g - hinaq Bending Stress P - Design Pressure D

P - 1.4 cal Prl.sary Membr.no S tre s s P, - Primary Membrane Stress P, - Pre s sore During Normal Or.s t ating or Operational Transient Conditions 4

sec o ess a t arv t oise t e.- v em

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NUCLRAR ENERGY 3 "

• 31 $~- 23 IU51NE55 CPIRAfl0N5 GENER AL h ELECTRIC aav 2 3.9 (Continued)

P, - Pe ak Pr e s sur e P - Test Pressare t

PM;7 - Probshle Naziassa Flood PS - Pool Seell i l

Q - Secondary Strese '

1 - Reaction Load R, - Pipe Reactions (including R,) Generated by DBA 1 - Design Mechanical Loads 9

Rg - Fallback Loads R, - Reaction Loads Daring Norse! Operating or Shotdown conditions RBV - Reactor Estiding Vibrations due to SEY or DBA RPMS - Reactor Pump Notor Saisure RV1 - SEY Dischasse Loads I;Y2 - SEY Air Clearing Loads EV2,,, - hEV Discharge through t>ne Lins EV2 Low Set

- SRV Discharge through Low Set Lines XV2 3j g - SEY Discharge through All Lines RV2 Two SRV Dischstge through Two Mj acent Lines WC0 - Rapid Valve Closure or Openf ag Ry - Tetr Vall Eeaction Lead due to Nogetive Dryvell Cliferentim!

Fressure (Wark III)

S - Allevable Stresi S - Espansion Stress E

S, - Design Stresa Intenrit!

S p

- Pask Stress SBA - Sea 11 Breal Accident SB L - Small Creek Accident Leadtog SEIS - Seismic Event SL - $1oshing Load hRSS - Square Root of the t,u of N .es..

SRV - Safety Relief Ysive Event SIN - m-Displacement 1,1mited Portion D ,

SRV g . m -Inertia Load Portion S S T. - Sale Shutdown Farth vake SSif - $51;-Dis placeme nt Lii ited Por tior SbEl - ,-Inertia Loed Por.ico ar o e e n a i.ev. ..:. i h _ -

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T - Thermal Ef f ect t-Total 7, - Thornal Eff ects (incinding T,) During DBA Tp - Design Temperature T

E

- Therunt Espansion LoaJe N 7* - Thermal Ef f ects and Loads During Normal Operetlag or Shutdown

, Conditions Ty - Thermal Ef f ect s During Test Conditican TEXA - Tubular Exchanger Manuf acturers Association TEST - Te sting Conditions TSVC - Turbine Stop Valve Closure U

- Allowable Stress for Factored Loading Cembinations, or the Fatigue Usage Factee

)

VLC - Vest Line Clearing Load V - Desiga Vlad t' - Design Basis Tornado TCO - Ta t e r Ca rry ot e r Load Y - Tield Stress tj -Jet impingement Load Ya Ya

- Local Force due to Missile Impact (including pipe whip)

- Reactica Force due to DEA

4. DESCRIPTION 4.1 Ceteral. The requirements of this specifiestion apply to the design, applicattoa and evaluation of performance of inactics of structures, systems, and components which are Safety Class 1, 2, or 3, and also to assillectes for those structures, systems and components whose f ailure could casse f a!!ure of the s tie ty sy st wo, c ompone nt , or structure neder varions poststated Icedtag scubinations. The established acceptance criteria must be met to ensure structural integrity of these safety related structures, systems, and components. Additional criteria may have to be sat.sfied to resure

~

operabilits piping, of active compon sts and functiosa!!ty of passive components, es, ht e te r a tasy. t eret t A

8 3 8 ' "*' ' ' sa~o.

25 NUCLEAR INERGY IU$lHill CPERAfl10NS G E N E R A L i))) E LE CTRIC atv 2 4.2 feellection 4.2.1 Desita. The requirements of this docsaent shall be a:t at the systes design level, the component Je sign level, the subcomponent design level and structural design level. It shall specifically apply to the design. locatics, and mounting of safety related sechanical equipment such as vs,1ves, including operatore, piping. pumps. ho s t eschengers, air systems. beating, ventlasting and air conditioning systems, tank, pressure vessels, core support structures, filter desineralizers; electrical egolpment soch as local Jrnction bones, local seltchgea r elec t tles t pene tra tir,os, motor s, sotor-operator s f or valves, solenoid operators for valvas, conduit itstellations, rahle traysi and control and lastenanntation equipsect and congonents sneh as lecal senses local junction bones. Instrument t ubing, pe ne t ra tions, loca lly poonted inst rume nt s, locally moneted penet e and racks; sechanical devices such as baf fles, deflector platosi and Structures such as buildless, platforms, compones.t supports. rack s. cable trays, pipe hangers.

The design specifications and docoment s for all safety related structures.

aystess, and components shall reqeiro that an adequa te evaluation of ine ef fects of the loadi.g combinations are taken into consideration in the design. including location, installation, and avanting of these structores, systems, and components.

4.2.2 Fvaluation. The requirements f or the evaluation of esisting designs shall te the ssee a s in Paragraph 4.2.1. except that design specifice tions ani documents supporttua tsie or:g:Lal d6*ina ns.; uot be d 4. 4J c tcft :t *Fe evaluation. The eveltation shall be performed in accordance with all procedures applicable to the design ans shall be documented and the docnmentatios shall be retrievable.

'. DESIGN RESIONSIBILITT 5.1 Responsibility f or tne evelaation and documentation of the effect of the coe binations of loads on the varioso structores, sy stems, and componsat s covered by this specification depends on contractual relationships sod the division of scope of work between General Electric Cempany, the Architect Engineer, the Owner, and Subcontractors thersto. Because of the natore of the regoirements, substart;el interf ace will be required between the various organiastions and within the varions disciplices of these organtastions.

6 REQUIRENERTS Th e event combinations, 10:49, and seceptesce ersteris preseated in this spselfication are appitcable to all safe ty related D'#R 4. 5. and 6/Merk 11 and IIT sy st ems, structures, and compocent s within the scope of this docusect.

lie dess to loadings shall be e stabit sned considering *11 plant and system operatir.a conditicas antic ars ted or nostulated to occur dursag the trateeded service life of the component, sy stem. or st ruct urd.

w o esminew is en em

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) NUCLEAR INERGY 385 2 32 = 26 IU$lWill OFitATIONS GENER AL h ELECTRIC aav 2

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i 6.1 Events acd Event Corbinations 6.1.1 Event Classifiestione_. The f ellowing events sad sobevent e shall be evaluated to verify that the resviting itsds, vben cusbiaed as discossed below, meet the acceptance criteria spprorciste to the item and locatsos:

Events },n)3_ve n t s Normal Operation (NO)* Stortop, Poses Operation, Dot Stardby, Shutdoen, and Reiseling Opera.t tonal Transients (OT) .fety Relief Yalve Actuation (SRV)

Turbine Stop Valve Closure (TSYC)

Inf regsent Opera tion Transient- (!(TT) Cask Drop Loss of Coolant Accident (10(.A) Design Ba' sis 4ccident (DBA)

Intermediate break Accident (13A)

Sas11 Break Accident ( SB A)

Bigh Energy Lise Break (HF1B)

Envirnemental and Site (EIN Severs Environment (as defined in Estreau Environment Paragraph 2.1.1.b)

Stissic Operating Desis Eartlquake (OBE)

Soft Shutdown Ostt!.46.k. (OSE)

Te s t in t ( fi.ST) i

• nis definition of h0 is applicable to sost convo.se nt s. Contaicaent structures nod certain equipment use a dif ttrent definition which s s defined where applicable.

~eo ser. new ee eo

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IU$1NE55 OP1tATION$ G E N E R A L d$2h E LE CTRIC aiv 2 I

4 6.1.2 Event Cembinettens. Its enzbinations of event s for which t ondings h sha!! be evalua ted are .hcwn in Table 1.  ;

The ietsution of supertspos.ng the 00E Icads on the SBA event is t6 proeide I g

margin during the initial stages of the acetdent. Darang the period of controlled ds pre s suris a tion and cooldo=? folleving an accident, it is not l ne de saary to consider the seismic event. Thus, the peak thermal stress in the  ?

dry sell clus the OBE load is not a de sign loading combina tion, j i

In the evaluation cf the plant responne to se n tmic ev ent s (OBE and SSE), it j shall be assum:d tha t the seismic event will be assoc (sted with the -

i Operational Transient (OT) which produces the sost severe loads on ths i c ompone n t or structure being evaluated, This combination shall be seleited j consistent with the enco' inter probability criteria f or the 6ce64 tion defined >

in l' ara gs sph 3.1. The loaJs snell theu be defined by combining the normat asc opset operstang loads (including the thermal sifects for the OBE evaluation) with the dynimic resposse of the c asro ne n t or structure due to seismic load and the dyuuisc toad of the noss a tics.61 transiett, usually due to SEV or TSVC by the methods deflued in Paragrapn o.7.2. 111: practice is ac.eptante in lien of, and as pref erable to a tetas ted merbsnistic evalnation of tha seismic even6s.

When a seismic event is considered in carbination with a LOCA eveat, stresse: -

( '. e to plant trenstent responses to the seismic event shall not be ca .s.derec, assauss stresses shall te son 6sdered only ist 0.1,ws. 0.6e .. I coeponents, thus postulated seismic induced 1siJ ure due to spurious action of N hte x:mic Ca te gory ( component s shall not be e s se ssed. I 6 .2 Dulldier interacti_gn 6.1.1 Seismic interactions a .Se i sm ic eve n t s , since they originato in the soll, will escite all ba semets or f oundations and hence excite sy4tems, structuras, and components within a!! buildioss.

6.2.2 Ncasaismic Intera,ctigns t Buildinha founded on a conson basema t ar/ be af f ected by the dynamic motion of one of these buildings. ".te degree of interaction f or the given plant conditions shonid be determ.ned osing a combined baildirg/basemat nodel. If lateraction exists, the vibratory co. ion of the building is which the causst ave evset occurs stil excitw sy s t ee s ,

structures, end couponente located within en adjacent batiding.

Como potential crists f or interaction between adjacent buildings sapported on sep: rate f:unda tions. These potentist building interaction ef fects thould be considered. (Pased upon analytsc da ta f rom the DhP6/ Nark III Standard Platt analyses there is tusignilicant int e rnet ton be tseen the BONI buildings.

!!e n c e , ef fcct s froo LOCA si e aRV event s will be con sidered only f or tne s P,ea ct or Bu116t ag and the s/ st ems, structures, and cesponerts contained therein.)

c o e e r .. $rv i o .o

NUtittAt kN8kQY SU31NE$$ OPERATIONS G EN ER A 'L 4h$h E LECTRIC " ' " ^ ' $4~o. 23 nEv 2 6.2.3 Internetton of Nenselsnic cotesory (_Struttures with Seismic Categorv I Structures. The interf aces be t'eeen Seismic Category I and Nonssismic Category I structures and plant equipment sha!! be designed f or the dynamic loads and 4.8.splacreents ptaJaced by both.

In the event of the eclispse of any Notseissic Category I structure it shall not impair the integrity of Setssic Category f structures or components.

G .) I,oad Phenomena. This section describes the possible loads which may .act on systems, component e, ano s truct ures taring verlons event s. This document is not meant to be independeat af the Cortainasnt Loads Report Nark III or tne

-DFFI Mark II which outline tbs phasing of subevente end loads.

6.3.1 Safety Rollef Velve Actuation loads (91Vl4 fafety relief valve discharge 1 Anes are part of the main s tems sys tem. This system is subjected to two significant non-DBA and nonseismic events: safety relief valve actuatica and torsine stop valve closure. In addition to the internal fluid forces generated by these events the main steam system will be subjected to the llenctor Building Vibrations (RBV1 caused by these events. Due to the interconnection o1 the main steam lines and the safety relief valve discharge line, the dynamic resloose of one line conid induce a dynaalc response of the

" other line. The motica of one lice will ef fect the stresses in the other line.

Its sata ry re.4ef valve system coatain. soteral valves. Safet) .eli.f ..l s loads are caused by the actuation lof one or more of these valvea which tre connected to the main stesa system. Relief valve actuation cac be initiated by reactor eressure increa te to the valve setpsints or by an active system saca es ADS, or usanally. Depending upon main stets system operating conditions, and other plant . conditions several combinations of valve ausuaiiva. .s. possible. Each combination of valve proep actuations cause diff erent hydraulic loadt. (RV2) in the suppression pool and transient fluid fl ow f orce s (EV1) in the pipe . The ( RY2 ) loads catsc dynamie response of the Reactor Building Structure ( (RDVs, which ir turn, cause dynamic response of all systems connected to tne React or Building strsetures. The Containment Loads Report and the DFFE contain a sore desatted description of these loads.

DeLasse many events may restit in ame 3r more saf ety relief valve.acttations, and since the components and structure

• responding to the d/namic SRV loading ,

and the resulties Reacter Buildic t Vios stion (RDV) can esperJ ence a number of load cycles per valve actuation, she possibilaty of f atigue d.. mage nust be considersd. As analysin must be performed to show either that the cumulative usage factor U, resniting from SRV Icads is no? significant or thac when it in combined with any other cyclic condittor.s for which the componest or structure is designed, the established fatigue crites1e are datisfied. It this analysis, the number cf saf ety relsef valve actuatioca that shall he cnnsidered are as shavn sn Table 3.1. It is generally conservat;ve to consider that ter each valve actuation the f atigue damage vill be less than i

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$no 5.3.1 (Continued) seven cycle s of the response calcolated; therefore, using seven cycles is a conservative ba sis. If the f atigue usage datermin)d by considering all load cycles as peak cycles is excessive, a leas sonservative Iced distribution may be used. Sowh distributions sust be indhidually Jastified.

6.3.1.1 Relit f,Velve !.ift ( A:oustic_ vave) ( Rvi ) , vben the relief valve is opened, the troussent fluid flew ceuse s tise-depetdent f orce s to develop in the discharge pipe, ne relief valves discharging into the enclosed pipios systes create amentary unbalanced forces acting os the piping system during the first for milliseJoods f ollowing relief valve lict, ne pressure waves traveling through the tiping system f ollowing the rapid opening of the relief valve will cause the .elief vaive discharge piping to vibrate and cr. ate f orce s and a(ment s. his vibration creates loads oni the siain steam piping systen, relief valve discharge piping support systas, subserged discharge device in the suppressica pool, and reection loads on reactor >2ilding structures which interf ace with thir piping syetes. Firid forces induced by discharge flow within the relief piping, system are designated a s RV1. Under conditions of steady state flow, the f orces associated with a flew ecting on the system are virtually self-equalizing and do not create significant bending a ments in the piping s/ stem, 6.3.1.2 ,S_s f e t y Relief Valve Leeds Due to Air Clearint (RV2) . h e relisi vel $. di.G s p i'.a s h t'.e e r.s i c sc 4 p ';in; *,- t r-

• t i c h a

• r-t e r, *F* ****- 'a the suppre m on pool, no opening of the saf ety relief valve reitults in a rapid compre ssion of the air mass within the discharge pipe, which then drive s the water les out of the end of the submerged dischstge device and ej ect s a high pressure air bubble into the suppression pool, creating an oscillating pressure load on the pool valls and basemet. These pressure loads impart accelerations on the structures and differential movements on the piping; these effev s shall be considered in the design and analysis ot' the safety rettei valve oischarge pipid6 systesi. If these 1cado are sheen to be significant, no dynamic response of the structures cause s .'yuamic re sponse if components, supports, sy s tees, and structurde attached t o the uenc t or Building structures, no RY2 responses are typteally chars:terited by floor Response S.ectra.

Since the dyntaic phasing of the RV2 discharge f erees f rte muis t-line dischanges cannot be readily ascertained by detersinistf c methods, the use of the Ceneral Electric Company's SRVA Monte Carlo ccuputer program provide s a realistic methodology for determining these pool dynaast, loads for Mark AII containmente. Fethods f cr dettraining pool d)usseis leads f or Mark 11 are su:maarized in the DFl R.

ne local hydrodynamic t ect s f run the f ollowing load coaditions shall be considered for the portion of the discharge line and qt.enchet d2vice in th; si?pression pcol regson, ard any vther riechanhal senporeus s 1ccated in this r2n.in, osoeera'atv t oisil

NUCLEAR ENER6Y 3 88 8^831 5m 30 abilN885 OPERATl'ONS G EN ER AL $ ELECTRIC arv 2 6.3.1.2 (Continood)

The nowher and scobinations of valves that util opea during a reactor vessel pressuse transient are as follows:

a. RY20NE - Design pressure lead (positive or negative) on the anppeession pool botadery resulting f rom discharge of one safety relief valve into the soppression pool. First actuation and sabengnent estustion shall be con.idered,

b. RV2710 - Design pressure load (positive or negative) on the srppression pool boandary restiting f rom discharge of two adj acent safety relief valves into the esppression pool.

o. EV2 Lov SET - Destga pressure load (positive or negative) on the suppression pool bonadery resniting f rom discharge of all los set safety reitet valve e into the esppcession pool which essee the large et a symmetrio Aorizontal loeding.

d. Ec.!g g - Design pressure load (positive or negative) on the suppression pool bonadary restiting f rom discharre vf all safety relief valges.

e6 ADG - Le.i.. pressus6 10 4 (posi.*ve et eise.1ve) w- -. 6;,.v.s!;s f3o*

bonadery roastilag f rom Antamatic Depressoriar.tioa Es n (A28) discharge liese into the suppression pool.

The above EV2 loads can be divided late two 414ssifiesta. ** EV2 g (inettist portion and RY2D (displacwawat 1Latted portion).

6 .3 .1.3 Safety Dettet Velve Actua tion for various Event s. Sry actuation combinations f or various events are presented in Table 3.

6.3.2 coeretionel Transient s Loads (OT) 6 .3 .2 .1 Turbine Stoo Velve Closure (TSVC) . Prior to turbine stop valve closure, saturated steam fleva through main steam piping at naclear boiler rated pressuse and mass rate. Steam five to the turbine comes to a stop at the initsat the turbine stop valve closes. However, the flow of steam f ram the react 9r vessel continues in the main stess line until the f1 tid compressica wave produced by the stop valve closure reaches the vessel nostle.

Reported refisctions ot' this wave at the vessel end of the main steam line and

, at the turbine stop valve end generate time-<srying forces la the main steam piping. Systems, steponent s. and structure s in the Reartcr Ballding.

Auz!!iery Building-Stes,s it.nnel and Turbine -Building are af f ected.

o

%80 807 A lmtv. t Dr$13

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d wuctrAR twenor '""'* 5m 32 IUSINE'll CPERATIONS GENERAL D ELECTRIC nav 2 6.3.2.2 Rat id valve closure or Osnina (RvrW . Estrese y rapid velve closure or opening in a fluid systes can cree te large pressure veves which can liropagate through the piptog systen and teto connected couponents. nis rapid ialve motion conid be taused by operating characteristics of the volve (eg, stif f ne ss of d!aphrsta in pnecmatic operators), and fluid flow forces acting an the valve parts during all ardea of valve operations. Design changes should be reviewed f or FVC0 problems.

l .3 .2 .3 Flow Indac ed vib rej, ton (FIV) . Flow cf fluide past objects creates local pressure disturbance s which create f orce s on the object. Ro se f orce s can cause dynamic response if the f orcing f unction and dynamic characteristics of the object have sp;,ropriate relationship. Flow it.Juced vibrations have been noted in nuclest power plant systems whica contain vortes shedding conditions (eg, heat eachsspers, reactor internal structures), pump, (rec Jrocating and centrifugal), and thermodynsmic instability conditions,

s. Vorte Sheddina. Vortes shedding occurs at certain fluid velocities when the 11 aid flows past object s. h e most cemmon componeet et BWR plant s in whsch vortet shedding is possible is a heat eschaefer. The fl ow pa s t the tube bundles creates vortices which create forces and cause dynamic response s to the tube bunille s. The dynaalc response is controlled by proper spacing of the support plates for the tabe bundit. De vibration cannos be ellainated but it can usually be controlled. IT.MA standards provide guidance for proper design. It is toportant in these ca se s to cen*8dae all reta e tel =ed*

• or emnnnec t operation, llydrodynants mass effects shall be considered. Ar,other groep of components susceptible to 11ev ind' aced vibra tion are pres sure, flow, and temperature ser scrs which encroach upon the flow stseas. and the reactor internal structures such as incore guide tubes, jet pumps, shroud, etc, which are designed for anticips:ed FIY and confirmed through prototype testir.g. .

b. Pressure Fluctuations. Joth reciprocating and centrifugal pwsps create pressure fluctuations in the fluid system. In west system de signs, these fluctuations are insignificant. Ilov ev e r, the possibility esist s that these fluctuations, coupled with the proper system characteristics, cat cause vibrational response to the sy s t em. Pressure at ter.ua tion devtse s can signit teently reduce the ef f ect cf this phenomenon.

c. Therme:rsamic_ Instability. Under certain rystem design conditions and opers. ting undes, tiuid dyna.nic f orces can be generated which create large pressure vertations, nese have teen noted in certain f eedveter systema

.here a reistively cold fluid layer is in contact with a relativcly hot steam region; under certain operating mode s sigallicatt water-hammer-type phenomena have accurred causir.3 a breach of the pressure rsteining bonudary.

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' NUCLEAR ENERGY 386n*832 $" ~a RU$fN85$ CPIRATIONS G E N ER A L (h)b E LECTRIC arv 2 32 6.3.2.4 Safety Relief Yelve,Aetuations with Overstirnal Transients 2 SRy astustion loads associated with Operational 1rsostenta are shown in Table 3.

6.3.2.3 Infreanent Ocorettois! Transients Losde

, , n. Cask Drop (Fed). The Fuel Br'iding contains the spent fuel storage pool and serve e a s a tevporary deposttory f or new and spent f uel element s.

Puol elements are transferred to and from the reactor, during ref ueling outagee, through the fuel transfer tube. Af ter a reasonable storage tLae j to permit inel decay heat to subside, the spent fuel is loaded Auto casks j within the spent fool pool. These casks are then lif ted out of the pool and transported to a reprocessing plant or sont to a jung term parassent storage location.

The building structures immediately below the travel tone of the spent fuel cask shall be designed to sustain the ef fects of a cask drop if the j function of the saf ety related systems structures or components could be '

im p ai r e d. The cask drop event sha!! be assumed to occur simultaneously l with an OBE event.

6.3.3 pes ten Basis Accident Loads (DRA) . The design basis accident is a postalsted event associated with a high energy lion break (ULLB). Loads from this condition are used f or the desten basis for systeme, structores, ard c(aponeats.

• *, 6.3.3.1 Lalte _ Bre ak thCA ( LDL) . The large break LOCA is a postulated event

• associated with the ruptcre of a recirculatica line, main steam line, or any class of pipe severance of equivalert cross-sectional flow ares. These Acads are generally associated with the Reastor Dailding, and are discussed in more detail in t'as t'ontainment Loads Report,

s. Annuto s Pre s sortr at ion ( AP) . Af ter rupture of a pipe (recirculation or feeawater, etc) withis the reactor vessel shield wall annalus region, pr#4ssarinetton occars which imposes f orces (pressure, jet Lmpingenent, and jet reaction) on the reactor pressure vessel and the shield well. These are generally asymmetric loads whi$h vary in space and with time. The subsequent anceleraticas and movements of the reactor presente vessel and the shield wall are transmitted to the umbroken piping systoss at (to rossle connections and the shield vs11 piping support attacLaents. Itens s<1 thin the reactor vessel snield well en~>lus region will be s'abjected to the differential pressures and fluss that a re seterated by the postulated rup t ur e. Annulus preneurization loads can occur in other confined annulos regions to which high mass and-erergy flow rates are injected. Rupture of ths steam line at the norsle produces force and reaction effects similar to thost described here escept that ttare is no direct A?.

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I

i NUCLEAR ENEMY 3 86 8^" 2 $~ 33 j IU$1NE55 OPitATIONS GENER AL $ ELECTRIC arv 2  !

j 4 6.3.3.1 (Continued) j i

b. Comparteent Pres sucitation (CP) . ' Af ter the posts 1sted rupttre of a high energy line, high energy fluid is released into the surrounds (t

c om pa r tm e nt , h is release will result initially in loc 61 pressure surge 4

and ultimately in a unif orm pressuriza tion of the ocupartment. Both the symmetric and asymmetric (pressure and/or jet inpingement) loads should be considered in evaluating the systems, structure s, and ccaponent s within I the compartment in which the break was postulated to oc:nr er.d in adj acent I

c<mportments. Dif f erential pres sure lost.ings between campar tment s shnil j

i be sensidered, his loading is sieller to sanslas pressurization but it ocents 1 s a larger, less confined space.

I I b. Vent 1,ine Clearina 1.osde (Vlf). De increasing drywell pressure forces l the vetor level in the weit wall annulus regica dovreard, until the tcp mais vent in the dryvell vall is espossd to the e':yvell environment. Tne water initially standing in the vent system ascelerates into the pcol and the verts are cleared of water. During this vent clearing process, the water 1 raving the horisental vents forms jets in the suppression pool and causes water jet impingement loads on the structures within the suppression pool and on the contalinnent well opposite the vent s. Hi s de scription is specific to the Hark Illi a similar phenomen.,on takes place j in the Mark I; with resulting vertical loada on the pool floor.

l i

d. Pool Meall (Pfjg As the air is the dryveti is parted into the wetwe11  !

suppression pool through the asin vente, the = Ar la injected below the .

pool surf ace, givtas an upwa rd velocity to the water above the we.t esit. i nas upward action of the ve tur is called " pool swell". Pool swell .

, creates pressure Acadings or, the pool walls and basemat. de subsequent  !

accelesation and acvement Lat arts loads os piping and other items in the i suppression pool and items attached to structures inside the containment cr attached to the containment (EBY). Pool stell also aanse s a moveunt i , of the surf ace of the water in the poppression pool. Any ittas in the air l

region above the initial pool surf ace v111 be subjosted to the l

1 hydrodynamic eff ect s (impact anJ drag) cf the pool swell, op to assimum height of pool swell.

e. Coge,e s a t i on os c ill a t ion ( CO) . Ccadensation osci11stion is the short term transient dynamis lateral load at the main- vent s occur:1cs bo tossa the main vent clearing phencuenon and the longer ters main vent chugging j effects.

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e NUCLEAR ENERGY 388 * '1 34 EUllN855 OPERATIONS GENERAL @ ELECTRIC arv 2 5 " o-6.3.3 (Continued)

f. Wain Vest ChoAtina (CnnG). After the pool swell has subsided, the unstead) cadensation of steam produces pressure fluctuations on the walls and floor of the vetwell suppression pool. Af ter the steam flew has been reduced to a low level (or during some small break accidente), the stems chega in the opper row of the seats. The chugging, which consists of a i steam babble intermittently filling a vent and than collapsing, produ6us loadings on the vent. The loadings on the walls (weir wall, dryvell, contalement) and besenat rosmit f rom the endden collapse of steam bubble s l l

which form at the ends of the vents, and anbsequently lapart accelete. tion i loads on compon.ats withis the Reactor Building. This description is I specific to Wstk III; Mark II is similar.

3. Is t e r Ca r rvove r lo e d s ( #C0) . Water carryover loads are dde to entrainment j

of water in the esta steam 11ae surging f rom the reactor pressore vessel <

af ter the rupture of a asia etets line econte. The est er-s t eam sixture has a much higher density than steam and imposes loads om main steam {

piplas system at each change la direction sad at each chsage in crose-sectional area.

h. Reaction Lead (I 14 Eq stvalent static load generated by the reaction on the broken high energy pipe daring the postslated break, and lacluding an appropriate dynamic f actor to account for the dyansic natste of the load,

i. Jet Ispinattent ( i) . Jet lapingement equivalent static load goterated by the poststated brosk. and including and appropriate dynamic f actor to account for the dynamic asture of the load.

j. Missile Impact ( e) . Missile tapest egalvalent static load generated by or during the postulated break, like a whipplag pipe, and lactuding an appropriate dynamic f actor to acconat for the dynamic nature of the load.

6.3.3.2 Intermediate Break IOCA (IRA). Those breaks for which the operation of the ECr4 v111 occar during the blevdova and which resatt in reactor depressurization. EBA is charactertred by an effective break area of at least 0.1 f t break below the reactor vessv1 water level.

Load phcaomema associated with this type of break aret

a. VLC-Mais Vest Clearing (VLC)

b. Pool Swell (PS)

c. Cos 'ensation Occl11stion (CO)

d. Main Vent Chugging : CHUG)

I 1

( wroestatsav. seres:

NUCLEAR INERGY 3 85 't^$31 5 ~a 38 BU$1NEl5 critATIONS G E N E R A L 1([h E LE CTaivHIC 2 6 .,3 .3 . 2 (Cont!nned)

e. Jet Impingemenc (Y )

f. Annulus Pres surisa tion (.W) 3 Blowdown

h. Pipe whip restraint attacLaent reactions Load from this accident are generally bonaded by the DBA escept f or ef fect s in the Atmediate vicinity of the break.

6.3.3.3 _Fes!! Dre ak I.OCA ( EH A) , Small break blevdowns skich do not result in rapid reactor depressurisation due to either loss of reactor finid or automatic operation of the ECCS equirment produce SUL loadings.

Specific treek locations are not postulated, however, the SBA results in the folluving dynamic loading pa snomenons

a. Reactor Do11 ding Vibration

b. Tcp Main Vent Clearlet (Mark III only)

c. Chusging

d. Sk'Y actua tion In addition, the SRA caselly resnits in the largest accident temperature (T,)

inside containment.

6.3.3.4 High Enerav Line Dreak s t Compon e n t s , loca ted la regions where high energy llues are located shall consider, in addition to the environmental ef f ect s (P,, T,, EBV) caused by e DUA, the rotential for jet impingement, d '.p s whip impact, and pipe whip reaction loads. The orientation of the pipe ba sak and the trajectory of the jet stres. shall be considered. The postulatet break Iccations shall be consistent with the requirements of ANSI N-176.

Local pressure variations shall be considered in addition to overall coupar tment pre s sure trade.

Corponent s located ir. compartments which do not roetain high energy line s shall be evaluated f or acsident pressure (P,) and temperature (T,) effects.

!!oviver, j e t Ltpingcaent (Y ), pipe whip Lapact (Y ,) do not require cor. s i de ra t ion l

l a. Y - Jet Lapingssent I

l b. Y, - Pipe Whip Impact

=so seensavv.iereir j l

. . . . - _ _ _ - - ___ .- . .~. _ -_ _ _ . _ . - _ - . _ _ . . _ . . _ _ _. _

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, NUCLEAR ENIRSY 38'

• 32 8USINE55 OPERATIONS GEN ER AL h ELECTRIC afv 2 5= 35 6.3.3.4 (Continued)

c. T, - Pipe Thlp Reaction Load on supports and components attached to the broken line,

d. P, - Cc=partment Pressure Load (neymmetric and symmetric)

e. T, - Compar tment Temp >ra ture

f. E, - Reaction f orce by pipse or supports on compcaent jurisdictional bonadery, occurring during a postulated rupture in another train of the systesa being er:1sa ted or in another system.

6.3.3.5 ff e t DBA Load s. For f uel recovery subsegment to a design ba sis accident, the containment shall be conservatively designed for flooding to an appropriate level above the top of the active fuel in the reactor core.

6.3.3.6 Safety Relief Yelve Actuations with DBA Events. SEY sctuttion loads associated with DilA event s are shown in Table 3.

6.3.4 Environmen t al and Si te_ Rel a t ed Load s ( FRV) . Emvironmental loads described in this parstroph are associated with natural phenomena which can occur in the atmosphere and hence af f ect builditas er Structures esposed to the a tmosphere. Selssic loads are discoseed sepstately,

s. Cecisa Vird (w) , Loads generated by the desigs wind specif t d for the plant site.

b. po s tan as s i s Torn ado ( w') . Loads generated by the design taats tornado specified for the plant site including missiles.

c. Probable _Was ione Flt,od (PMF) . Hydrostatic load des to probable me:Imaa flood specified f or the plant site.

d. External _ Wissiles ( e) . Missiles generated by the design basis tornado and otbar probabit- sources as speelited f or the plart 61te.

6 .3 .5 Seismie Loads. Noclear power plants are designed for two basic sels sic events: Op6 rating Dasis Eart1 quake (WE) and the Safe Shotdown lar hquake

( SSE) . Upon s1sppage St f anit s le the carth's crest, dynamic waves are propagated through :he soll. Then these waves encounter a building foundation the f oundation undergoes motion dos to the dynaalc soil waves.

This motion of.the fasadation(s) causes dynamic time-'ct.sving response of botiding structures connected (directly or indirectly) to the lonadstlen(s).

Tae building dynamic e.otion in turn causes dynamic response of scoponent s.

I supports. systems, sad structures a ttached to the bLilding. Dynamic seismic response is typically characterized by Floor Response 6postra (FRS),

e.s o sev a ine v. s esso a

l wucun zusasy G E N E R A L y) E LE CTRIC 38

• 81 $ " o- 37 IU5fNt55 CPtaATloN5 me i 2 l

6.3.3 (Continusd) l l

Loads generated by the SSE include t l i

s. SSEg - SSE inertial portion of load. i l

b. SSEp - SSE aisplacement limited portica of load.

l Loads generated by the OBE includet

a. OBEg - OBE inertial port ton of load. t i

b. OBEp -

CDE displacement Itaited portion of load.

j 1

Because the component and strScture response to seisato loads results in a  !

number of load cycles the possib.41s ty of f atigue dsonge most be considered. {

The upse t event evaluation shall consider one OBE producing casolative uns te equivalent to 10 cycles of the pesh calculated resFonse.

6.3.5.1. Stoshina tilla Thenever a free water aorface exists in a ta nk or pool, dyntaic escita tion of tha t ites will cause the water surfsco.to cocillate. This oscillation phenomenon is called sloshlag. Sicshing is usually as sociated with sei tnic events; hewever, other dynmaic events (D8A, S RV) can also cause sloshing.

Cwassne.l. locat.J insi4e a powl re61wa L.luw the initial ..ist surf.c. <111 be subjected to hydrodynasio e.f f ect s associated with a seissio event.

Cenponent s located in a regios obere pool slosh will occer, for example inside the spent inet pool, compsner to partially filled with veter will be subjected to an additional loadies rosalting f rom Sloshing Ef fect s (SL).

$ .3 .6 Tet13a Loaf e (TEST) 6.3.6.1 Test conditions _, For short-duration of f normal conditions imposed by the operator f or code conformance test (eg, RPV: hydrostatic tosting) or system perf ormance test (eg, high-recircula tion-11ov to et f or flow-indaced vibration), the fo11& sing criteria shall apply,

s. Components to which jubarticle NB-3114 of the ASKE Code.Section III applice: Test cotidition criteria, as defined in Section I!!, Subsortion NB, shall be used f or component s and to se conditions to ubith Subarticle NB-3114 spp1t is.

D. All components to shich Subarticle ND-3114 does not app)y, other than i

. con t a i nme nt structures: Normel conditine crit + is may be used for acy I tests, r.co e st a t arv. ie >se t

_ . . _ _ - . - . ~ . . _ _ _ . .__ _ . _ _ _ _ _ . .m., _. . . . . . _ , _ ~ ,

_. .. _. --. _ - - _ . _=_- - - .-. . - - . - _ . _ .

I NUCLEAR INERGY 3 " "" 5" * '8 IU51 NESS OPERATION 5 GENER AL $ ELECTRIC aav 2 6.3.6.1 (Continued)

c. Ha s tana test condition partmetere for mechanical ccaponents and systems shall be considered in the design of e.tectrical control and instramentation devices when portions of these d:sicos are subjected to the test conditions but do not goality as part of the pressure re taining bocadary for the nochanical system.

6.4 Lords en Components. The snecific loads which act on a component in a given location, as a result of a spe:Afic event, are presented in Tsbles 4.1 throu gh 4.5.

6.4.1 Component Operability. The loading combinatloan and .cceptance criteria provided in this specification address only the structural integrity j of the itia to be evsinatedt these criteria do not ensure functional i capebility of the ites, which is beyond the scope of this specification. In addition to the structural integrity acceptance criteria presented in the ta' ole s in ths. section, act ive component s aball be de signed and evalua ted to assure operability and essential safety functionality nader the root severe of the event cambinettons presehted in these tables. Special stress or deformation criteria may be spplied to piping to assure that its fuaction of having adequate capaosty tv conduct finid is maintanned.

6.4.2 General Weetanica! Component 1,o . d a i Many mechanical systems are located in more than one of the DONI buildings. Each of these buildings es!.itit thcar ces ualque responsa scd environmental characteristi:s. For those portions of mechanical systems located within these batidings, the proper environmental response-characteristics of each bu!! Jing, for each portiJD of the systems, shall be considered, including the potential for building interaction as described in Paragraph 6.2 Thus, the portion of the main steam system located in the Turbine Building need not be scalyzed for the floor response spectra assocated with the Reactor Building. Howsver, the internal forcing functions caused by a turbine stop valve closing witl propagate throughout the entire main system, regardices of building location.

Ditterent structures, even when located within the name building (eg, dryvell vall and shield building) will exhibit different dynamis response characteristics and hence oscite similar component s which are attached to them, differently. Inr ther, the se structure s s ay exhibit out-of-pha se respotse stich can adversely load a component.

All of thec. f ectore shall b, cons,tacred in defininf the appropriate multi-support input dste for the asi.js*. of these systems or portions of sy s t em s.

Systems loco tsd wholly within one hullding, but supported frow another building, will be subjected to saltipla excitations and probable differentist building novement, nao ser a enew. ieseo m

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6.4.2 (Continued) ne loads which are gerers!!r applicable to all saf ety related Seismic Ca tegory I mechanical sospucent s include:

a. D - Teight of component including contents during normal operation and test ronditions.

b. P, - Wasinua operating pressure and dif ferential pressore which occars during norms! plant cperation.

c. T, - hermal ef fect s daring nornst pleet operation includings (1) T -En eraal esponsion due to operational loads and envirsumental temperature including displacement controlled motion of laterfacing systems, struc ture s, or ccgaponent s.

(2) T T- normal transi6nts includict radist constraint of pipe; priasrily applicable to Class 1 composents.

d. R, - Reac tion loads a t the jurisdictional boundary frcgi other systems, structures, or components; ILcluding pipe reaction forces on valves, pumps, and vessels; and component support reaction forces,

e. P - Design Pressure D

f. FIY - Flce Induced Vibration caused by finid flow past an iten located in the fluid stream,

s. F pg - Prestress loads on bolt s, flanges, and gaske ts.

F. O!!E - Opera ting Ila s t s Earthquake

1. SSE - Saf e Shutdown Earthquake 6.4.2.1 Abneraal Plant 1. cads on Meeheniect Components

a. T, - Th e rmal ef f e c t s dise t o n o " or abnormal (D3A) plant conditions

b. P, - Pr o s este e f f e c t s 6, R, - Reaction leads at jurisdictional bous daries I

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BUSINESS OPERAf t0N5 G E N E R A L 4($) ELECTRIC arv 2 6,.4.2.2 Abnormal Loads in the Weset_cr Cuildina. In addition to the loads defined in Paragraphs 6.4.2 and 6.4.2.1 above, the f ullering loads shall be considered for all component s lose ted sitLin the kesctor Building:

a. LBV - Loads dos to Reactor Building vibrations cautee by so SRY ct 034 event,

b. CDC0 - Loads f rom couponent response er direct fluid feece6, on a c ompone n t located in the suppression pool, caused by a bagging phenomenon.

c. 00 - Loads f rom ccaponent sesponse or direct fluid forces, on a component located in the suppression pool, caused by condensation oscillation phenanecon.

d. YLC - Loads f rom component response or direct fluid f orce s, on a component

!ccated in the suppression pool, caused by main vent isne clearing ph e n om e no n,

e. Rv2 - Loads f rom component response or direct fluid force e, on a cersponent located in the suppression pool, caused by safe ty relief valve air clearing phenomenon,

f. PS - Loads f rom component response or direct fluid forces, on a component located in the .uppression pool region affected by the pool swell, caused by the pool swell phenomenon.

g. AP - Loads f rom cosponent response or direct steszflow force s, on component s located in the re.ctor ve asel shield well annuins regina, caused by annslus pressoriastaen.

h. SL - Loads f run component response or direct fluid f orces, on component s loca ted in the sloshing goes of a pool or component, corsed by the s1cshing phenouenon f rom any dynamic event.

6.4.2.3 Other Loadings. There are other loadiny.e which may be of a locs! or restricted unture bot whist may have a decided cife6: on some compone.t or s t ruc t ur e . The responsibility fut identification and apple. cation of such loadings lies with the design enginear.

6.4.3 elec t rie el Con t rol . . and in s t rume a t a t ion _ Esc ingent Leads. Electrical, control, and instrumen:stion eqcirment .cd cestocent s ato their surport s mounted on or attsched to the structures and mechanscal compocents will be subjected ta dyntaic ! cads. The s e l oad s re s ul t f rom the tebponse of situctures and mechanical coepenenta and to directly applied loads resultin:

f rom postula ted event s. The magnitude of the load la dependent upon the location cf the item and the causa tive event, ht o 5 6 f a t elv, t t/sil

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Instrument s which sense process system paremeters by direct connestion to ficid linst and are not considered as part of th,* sechasic al sy s t ea pre s sere bo tada ry ( i n t e rm s o f t h e ASXF j ur i s di c t i c al boutdary) .*s e conside red a s electrical equipment.

Items which are located to areas which contain hagh energy piping shali be designed for the following ef fects assveisted with a GUA Jet impingeesst and pipe whip impact. In addition, in other areas adjarent to areas containiss high energy linen, the *ffects of t,. P, will prebably requise censideratien.

This doensent addresses only the structural integrity ef fects assosisted with these loading conditions.

The loadn which are generally applicable to all safety raleted Se! ste Categ ory I elec t rical, in s t rument a tion, and con t rol equipmen t incicce

a. D - Teight of ites

b. P, - Maatsum internal ope.ating pressere for items connecteJ to finid lines.

c. R, - Reaction loads on interf ace boundary f rom structures or componsats.

d. ORE - Operating Sasis Earthqnske

e. SSE - Saf e Shutdown Entthquake

• f. FIV - Loada das to finid flow conditions (tlow induced vibration) 6 .4 . 3 .1 distItal Plant Leeds on Electriesb Control, and inst"smenta* ion Fontenent

, s. P, - Wasiana external presente due to s.ny nors el or abnormal (DB A) plant condition.

b. 7, - Naziscs c aterne t temperatura den to any normal or abnorcel (DR A) plant condition, i

c. R, - Mastana reection force on interf ace bounda ry due to neraal or a

s'anormal (D3 A) plant condition.

l

d. Y - Je t irpingement load.

j

e. Y, - Pips whip impact load.

f. RRV - Loads due to Fesctor Batiding vibestions cassed ry an SRV or DBA event.

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' ' 6 .5 Loads on Component Supports. The specific loaJe which est on a ecsapcnent support in a given location as a result of a specific (~eu are pres?Med in h Table s 4.1 through 4.5.

11 loads (a component support anchcrages shell be cor.ssdered as primary inad% Loads frca che restrained free end notion of pipirs shall he

,., y onstvered a s priury loads in cot.ipontat and pipind suppor ts.

6.5.1 Ge n e r e t Corm _om e t S u p p o r.,, tlc a d s .

'n.* f oll owin g l oad s .; o T be considered in the design of all component supports which transfer loads f rons g setety related Swismic Category I items to Lu11 ding structures:

. . a. D - Weigbc of support and supported i t.e m ,

f b. L - Live loads includina construction losas.

4 T, Loste due to haut tr ansf er f r om ccmponen't during normal and abnormal operating conditions,

d. Tg - Thermal expansion loads f ror.a tapported companent including frictional M. effects. "

s. Our ge, Mional Basis Earthquake leads f rca supported ite, (including ine s t a l c. i displacemer.t portiot.s) and dynaase rs aronse of caponent suppos;.

. f. SSE Saf e Shutdnt Ear thquake loads f rom snpported item (including inertini :...i displacement ,ortions) and dynesic response of ccuponent sy gort,

s. T, - Lead on support due to envirorsental thermal conditions.

h. f j- Load on support due to je t impinseent effects.

1 Y, - Load on support lue to pipe whip impact.

' j. Y, - Load on support he to 'ptoe whip reaction,

k. F, - Friction loads due to pipe mov: ment.

5 .5 .2 iba o rm a l Lc a i s su ,tye Reacto? Otilding. In addition to the leads Jefined in Paragraph 6.8.1 above. the following loads shall be considered fee al; component supports ocated within the Reactor Building:

a. RDV - Loads dce to Reactor Building vibrations caused by = = '" or DD A evens.

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b. CIUG - Loads f rom component response or direct fluid forces, on a support located in the suppression pool, caused by chugging phencuenon.

, c. CO - Loads f reu compone nt response or direct fluid f orce s, on a support located in the suppre ssion pool, ca'ased by cosdensation oscilla tion phenomenon.

c. VLC - Loads f rom criaponent response or direct fluid f orces, on a support located in the suppre s sion pool, caused by mata vent cle aring phonctaanon.

e. PV2 - leads f ece component respense er direct fluid forces. on a support located in tha st.ppression pool, caused by safety relief valve air

, clearing lo.sds,

f. PS - Leads f ree component responso or direct fluid force s, on a support

' located in the soppression pool region aff ected by the pool swell, caused by the pool swell phenosanon.

g. AP - Iceds f rce component responsv or direct steamflow forces, on supports s Incated in the Reactor Vessel shield wall annalus realoe, caused by a:nalus pressurination,

h. GL - leads f rce componer.t tosponse or dit rei finid f orce e, on auppert s located in the sloshina, mone of a pool or component, caus3d by the sloshing pl.enomenon f rce any Jynamic event.

6.5.3 Other Loadings. See Para graph 6.4.2.3.

4.6 'oads ou Structures. The specific loads which act on a given stru6ture as a result of a avecific event ure presented in Tables 4.1 t'tronh 4.5.

6.6.1 General Structural Loads. The loads which are generally app:1 cable to all BWI building and structures inctnde:

a. D - Dead load of structure plus any other perunnent 1:ade and incluoss vertirsi and lateral pressure or liquids.

's . L - Conv6ntional floor or roof live loads, mor^1e equipses.t loads. Ioads caused by statie or esi*Mc ear 5 e pressure s, sv*. other variable loads such as construction loads and hoist loads. The most severe value(s) shs11 be considered.

c. E, - P.;e reactions during normal operating or shardown cond.tions, ba sed

_ __ on the most :ritical transient or steady state condit!. : .

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IUllNill OPERATIONS G E N ER A L 4(h) ELECTRIC 2 Sa ~a l Rev 6.6.1 (Continued)

d. T, - Thermal ef f ect s med Acads during normal operating or shutdoen conditions, based on the most critical transiest or steady sta te

_ , condition.

e. Construction Louds - Construcaios loads are ;c As which are applied to the structures from start to completion of construction. The definition for D, L, and I, shall be based on actual construction methods and/or conditions.

6.6.2 Abeormal Pteat Loads on Structures. The following are abnormal plant loads which may need to t* considered f or structorse:

s. Fg- Internal pressores resniting f rom flooding of compartments; Lydrostatic head.

b. P, - Accident pressure due to DBA.

c. R, - Pipe rescLions (including R,) nader thermal conditions generated.by the postulated line break, d ., T, - Thermal ef f ects (includits T,) which may occur during a de sign accident (high energy 11ae break-EELB).

e. Y) - I,ocal pre s sure on structure due to j e t Lapingement caused by the postulated line break including impact ef fects and drag loads,

f. Y, - Local for6e on structure due to missile (laciuding whipping pipe or other missile).

3. Y, - 2eactien force on structure due to postulated line_ break.

6.6.2.1 Ab no rma l Lo ad s in t h s' Ro s e s o r Ru t ld in,34 In addition to the loads

.s defined in Paragraph 6.6.1 and 6.6.2 :above, the following loads shall be con.idered for all component supports locate i within the Reactor Building:

s. RBV - Loada due to P.eactor anilding vibratiors caused by am SEY or D3A event,

b. CDUG Lcada from coeponent response on direct fluid forces. on compone n'.s located in the suppression pool, caused by chugging phenomenon.

c. CO - Loads f rne ccuporent response or direct fluid forces, en a component located in the suppressian pool, cauced by coadensation oscillation ph e n om e non, l

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NUCLEAR ENERGY 3"""' s % o' 45 BUSINESS OPERATIONS GENER AL h ELECTRlC m 2 6.6.2.1 (Continued)

d. VLC - Loads f rom component response or direct finid f orce s, on component s located in the suppression pool, caused by main vent line clearing

• • phenomenon.

e. EV2 - Loads f rew temponent response or direct fluid f orce s, on component e located in the suppression pool, caused by safety relief valve air c1 caring loads,

f. PS - Loads f rom component response or direct finid f orce s, on component s located in the eippression pool reglen affected by the pool swell, caused by the pool swell .chenomenon.

g. AP - Loads f rom component response or direct steaaflow forces, on components located in the Reactor Vessel shield wall anenlas region, caused by annulos pressuritation,

h. SL - Loads f ree component response or direct fluid f orces, on component s located in the sloshing zone of a pool or ccuponsat, caused by the sloshing phenomenon f rom any dynamic event.

6.6.3 Loads on Contalement S t re e tte r e s . Co n t a ina.e nt structure s may sustain the follwing loads:

a. P, - Fressare dif f erence os ti.een the interior saa essertor vi the containment, considering both interior pressure .:nsngu because of heatiot or cooling cad etterior atmospheric pres. re variations.

b. SRV - Safety Relief Valve Loads.

c. I - Impset load due to crane operation.

6.6.3.1 Preoperational Testion 1.osde

a. P t

1' essure e8 los e typ ed nr ng the structural integrity te st,

b. T g - nsensi effects daring the structural integrity test.

6.6.3.2 Severe Environmentel Loads. H e lead due to the Op.srating Basis Eart*.aquahe (OBE) including Fuel Transf er f ool and Suppression Powl Sloshics effects.

6.6.3.3 Futreet Environmental Loads. The load due to the 3sf e Shutdown Earth t oske (SSE) incisding Iuel Transf er Pool and Suppressica Pool Stoshing effects.

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I BUSINESS OPERATION 5 G E N E R A L 4($h E LECTRIC nev 2 6.6.3.4 Abnormal Plant Loads

. 4 P, - Local force er prest sre on containment struct ure cr pe r.e tration catssd by rupture of any one espe, s

(1) Local compartnental pressures (* the contsimment vall from ruptured pipe s in compar tment s out side th, wall, within the Ausiliary Building.

(2) Local compartaantal pressures inside the containment Luclude those from the meia steam pipe tunnel.

b. Design Basis Accident (DBA) - The Mark III containment shall be designed to sustain the f ollowing loadings as a result of a Loss-of-Coolant Accident (LOCA).

(1) P, - Containment design pressure load duries a loss of coolant accident design baa8e accident.,

(a) PCDC - Containment and drywell design pressure, during a design base accident (DBA).

(b) Design Basis Accident (DBA) suppression poo1* loads - Immediately fatto=ing a destrn basta accident. the containment wall, f ounda tion est, drywell well, weir vall, and structures loca ted above the suppression pool will esperience various pressare and s dynamic loadings as a result of the vent DBA clearing process

into the suppression pool.

( i) PS - Pool sveil bubble pressure on containment anJ drywell wall (axisymmetrical) and monazisymmetrical on containment.

( 11) P 7D

- Containment wetvell dif ferential pressure exisymmetrical due to flow restriction of the hydraulic control unit (BCU) floor to pool swell concurrent with a differentist dryvell pressure.

(iii) I, D - Impact and drag loste on structures located above R

the suppression pool to upward pool swell.

( iv) ' Ep - Fallback loads of casc.ading water onto structures stove the suppression sool immedi .tely following' pool swell.

(v) CO - An oscillatory dynamic loading (condensation oscillation), on the suppression pool bcondary due to steam condensa tion at the vent ezits during the ;<riod of

=en eera taav. teren high steam as so flow through the s eats f ollowing a LOLA.

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( vi) CHUG - An oscillatory dynamic loadlag (chugging) in the top vent and weir sanslus or on the suppression pool boundary due to steam condensation inside the top vent or at the top vent exit during period of low steam eass flow in the top vent only following a LOCA, (vii) R, - Invard load through the venta applied to the weir wall, due to negative drvvoli pressure differential.

(2) T, - Added thermal effects (over and above operating thermal effects) which may eccur during a design accident and which correspond to the unfactored design pressnre, P,. Accident thensal gradients for the containment done, wall, and foundation est shall be considered.

c. The Design Basis Accident (DBA) in the Mark II containment producee loading ef fecta which are similar to thoss described f or the Mark III.

kef erenco should be made to the DFFR (Paragraph 2.1.2.b) for details. '

6 .7 Evalnatter Load Con *oinatione 6.7.1 ILittel conditions -

6.7.1.1 fil Loadina combinations. The initial condition for each loadirs '

combination shall consider all of the even?e and conditions defined Jn <

Ssation 6.J. l 6.7.1.2 Static Events. For plant start-up mode, the initial conditions shall be ambient temperature and pressure associated with the plant cold shosdown c o ndi tion.

For operating transients, the initial conditions shall be the most limiting conditions rhich exist during steady state power operstlou.

6.7.1.3 Event conbinations Not Including _, Seismic Evest a. The initial conditions f or all nonsot anio event and loading combinettons shall ba steady state plant operation at any permissible power condition, including pc.er  !

levela up to that corresponding to rated stesa flow and including zero power i hot standby with the main condenser in service.

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6.7.1.4 Event Combinations factuding Seismic Event. The initial conditions for all event and loading combinations, including selssio events, shall be the stesef state operation at rated' recirculatiin flow unless the combined

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probabili ty of the seissic event and the plant conditlos considered require a o different initiation basis and acceptance criteria.

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EU5!NE55 OPERATl40NS GENER AL h ELECTRIC arv 2 6.7.2 Combinint I,osds. The event combinations shows in this document are meant only to provide a description of which events are to be considered in i rombination with other event s and are not meant to indicate that the Iceds or stresse s f rom these events are to be ceabined algebraically. Th e follering

, paragrapha are intended to serve as a guide in combining loads and stresses I

from at least two dyntsic events on loadists phenomena f rom the same event.

i 6.7.2.1 Wethods of Nethematteal Combinatijne for Static Events, Static everts are considered as deterministic eve nt s shall be added algebraically.

Veight, thermal expansion, and static anchor movessat loading conditions are l

determinletic events.

6.7.2.2 Approach to Mathematical Combinatiot of Re secuse Time-His torie s.

5 This paragraph pre sents guideline s f or combining response t Lor-histories due to:

l

a. roncurrent loads due to a single event.

b. Concurrent loado due to a combination of events.

t l There are methods for combining responses that yief d results which are or most i

likely to be *c~arage" (ie, result s approximating the expected peak value),

l and which yield

• conservative" results (ie, calculsted peak value e which have a low probability of being exceeded). In general, it la preferable to use a best e st Laate method since ' conse rvatism introduced in the mathematical conuination prosess is artificial. .'. t is, it is ;raferable to intr du:= the l desired measure of conservatism in -the evaluation of loads and stresses, and j in the allevabis Itaits, where the conservatina is tiod directly to the phenomena and is thus phy.ically comprehensible.

In partionier, in dealing with low probability combinations of events, consideration of the total probability (the product of the probability of simultaneous event occurrence and the probability-of exceeding the calculated

, combined value) any justify the use of a best-estiaase method even when the application of such a asthod is subject to uncertainty. In.several tr.por t a n t load combinations of Tables 4.1 through 4.5. either the probability of time-history overlap is very low (eg, LOCA + SSE) or the probability of one of l the events alone is very low (eg, SSE, including relief valve loads due to turbine trip). In such instraces, too probability of the event or event combination as well as the probability of exceeding the calculated combined

! , value should be used in considering the appropriateness of the approach, i'

Then loads can be shown to occur with insignificant overlap in time (eg, pool swell and post-LOCA contattament flooding resulting from a ihCA), .1 is not necessary to combine them at all. )

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$m 6.7.2.3 Wethode for Natheast teal Combination of Response Time-Bistories, (Ref er to Paragraph 2.2.8.a)

a. Ab solute Sua of Pe ak Amp 1 s tudes ( ABS) . Combining concorrent responses by their peak amplitudes is the most s traightf orward combina tior m e thod. If peak amplitudes are added irrespective of their time sequence, a conservative result will be obtained. The method approaches a probable result when the loads are fully espected to overlap la time for mechanistle reasons or by assnaption, and ween the overlapping response time-histeries have substantially diff erent dominant frequencies. It respeares are added as functions of time when the phase relationship is known, the method yields a probable result. The messare cf conservatism in ABS is extremely variable, being strongly dependent on the frequency content and the phase relationships of the loads belas combined. ABS should be used in design evaluations only wk.o (1) The waveshapes and timing are auch that ABS gives a probable result (2) A conservative combined value is Jesired (by contrast to a probable tombined valna of conservatively determined loads), es, for active components importsat to saf e ty a s it. dica ted la Paragraph 6.7.2.4.

b. Sanare Boot of thelas of Soneres (SRSS) . Comblains concurrent response time-historie s as the square root of the sua of the ogsares of- their peak amplitudes is f requently a good approzination and is widely used. Use of Sr.'S tas teen justif t:d by the use ef r:b-bility threry. P.SS isy be used for the following cases:

(1) Combination of SSE and LOCA dynamic responses for all ASME Class 1.

2, or 3 systems, components, or supports.

(2) Combining responses of dynamic loads other than LOCA and SSE provided a nonescoedance probability (NEP) of g4 percent or higher is achieved for the contined SRSS response. Aa acceptable method of achieving that -goal is if Conditions A and B are both satisfied.

(a) Conditiva A The dynamio response time inactica is varying.

Duration of the strong action of the function is short.

Fanction constate of a f ew distinct high peaks which are randoa with respect to time.

Response is calculated on linear elastic basis.

Time-phase rotationship among functions to os met:sd la randon.

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(b) Condition R. For loads which meet Condition A, the SRSS method  !

may be used provided a nonesseedence probability (NF/) of 84 percent or higher is achievsd for the combined response. An acceptable method of attaining that goal is meeting all the f ollowing requirements:

Define loads at approatmately the 84 percentile or 1.15 times the medias, whichever is greater.

ne SRSS value of the response combiastion has a NEP

). $0 percent selected frosi a Cumalative Distrabution Function (CDF) curve constructed on the assumption that ladividual response amplitudes are kn.en and only raudcun time phasing defined by its probability density function exists, no CDF curve any be developed using the procedures of Appendiz N of Section III of the ASME Code, or alternatively, methods developed by Brookhaven Nations! Laboratory, using absolute (unsigned) vaines of response amplitado may be used. The method selected shall be justified in the sabaission for the application being analyzed.

1.2 timer the SRSS vaine of the response combination has an NEP 15 perc6nt from the CDF curve constructed as in "b" above, sas :::1rg -nly rande: ti=r rhasit:.

6.7.2.4 Methods foe Mathematical Load Combination for Active Components (Design Criter u f or Ac t tva Component s) . Active component operabilsty to required limit s avat be demonstra ted by testing or analytis at absolute sum levels of combinod peak response, n is demonstration must include ,,ecific consideration of the dynamic nature of the loadings and response, component f ailure modes and margins, as well as evaluation of the extent to which transient loss in operability af fects plant safety.

6.8 Loading Combinations and Acceptance Criteria. Loading combinations and their corresponding acceptance critorie which are applicable to components, component supports and structures presented in Tables 2.1 through 2.10.

6.8.1 .N_e,c h a n ic a l Compo ne c t s 6.8.1.1 ASME Mechant al,Jomponents. Loadits combinations and acceptance criteria f or ASME mechanical components are pre sented in table s 2.1(a) and 2.1(b).

4 to gera inev e sset)

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NUCLEAR liHEllSY 3888A2 $m 32 SUSINE55 CrERAflO!!$ G E N E R A L h E Lf.CTRIC scv 2 6.8.1.2 h.g- AWE We e b e n i c a l Cemeoc ch Loading combinaticas and acceptance criteria foi non-ASME mechanical corponent s are presented in Tables 2.2 and 2.3.

Acceptance criteris for this type of equipment shall be based upon either

e. Analys.is tu acccrdance with the requirements of IEEE, ASME Section III.

ASME *iection VIII, AISC Specifica tion. AA Specifications, or other appilcable Standards.

b. Test per JEEE-344 (1975) or equivalent.

6.8.2 Electrieel. Control. end instramentation Coseonents. Leading combinations, and ecceptance critoria f or electrical, control, and inst rnrect a tion equipment and components are presented in Tables 2.2 and 2.3.

In general, the acceptance criteria for this type of equipment shall be tased upon either

a. Analysis in accordance with ASXL Section III, the AISC Specification, or other applicable Stendards.

b. h Testing shall be in accordance with IEEE-344 (1975). Nonmetallic parts will require special consideratiou if their structural integrity must be assured.-

5.8.3 Comeon e n t,..he r t s 6.8.3.1 Oneconent Sn norts for ASVE Section III Comeonents. Leading combinations ana accept ance criteria for component supports are presented in Table 2.6.

6.8.3.2 Cemeonent Su u.2Jts far Non-ASVE Comucs:9's. Loading ccabitaatious and j scceptance criteris f or component supports, valch transsit load: free non-ASME l components (mechanical, electracal, control, and instruaentation) to brilding l structures or other componente, are presented in Table 2.10.

6.8.4 Structures 6.8.4 I Contalement. Loading combinaticas and acceptance criteria for containment are pre sented in Tabios 2.7 and 2.8.

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30$1NE$5 CPERAlloN5 G EN E R A1. $ E LE CTRICarv 2 6 .8 .4 .2 Structures Other than Containment. Ioadius ccobinations and acceptance criteria f or Seismic Category I buildings and structures are presented in Table 2.9 f or concrete structures and Table 2.10 for st el structures. Losi combinations and acceptance criteria for refueling sad servicing equipment are presented in Table 2.11. l l

6.8.5 Ruptured Pipe Criterie. The criteria for evaluating the ef fect s of a postulated rupture of a high energy line shall b= in accordance with ANSI N176 l e

or other stallar standards. The detailed requirements for this condition are I beyond the scope of this document, j i

TABLE 1 BASIC EVDiT COMBINATIONS AND CATEGORIZ ATIONS NO SRV SEISMIC ACCIDENT N

EVDsT( S) OBE SSE G L(114) G L(4) CATEGORY OT I I Upset <

CT & OBE I I I Duergency OT + SSE I I I Faulted IBA or SCA I I I Emergency IB A or SDA I I I I Faulted

+ OBE IB A o r hdA

+ SSE I I I d Fa ul tsd M A + SSE I I I p2) Faulted NO I Norm al NO + 00E I I Upset (1) Use SBA or IBA shichever is governing.

(2) Loading due to IRA / SBA/IBA is de s c ruined from rated steady . state conditions, (3) NO - Ncrual los(. cr nalsts of pressure, dead weight, thermal and fluid reaction

• loads.

(4) SEA, IBA, and LBA shall include all event induced Icads whichever art applicable. Examples of these possibly applicable loads incisde annulus '

pressurization Icad, pool swell load, condentation oscillation load, and chugging load.

(5) See Table 3 for the possible sets of simul tane ous SRV *a'a+ ; 3 ~

-a ar r. r r e n t -!O s the eve nt under considera tion.

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4 TABLE 2.l(s) CLASS I MECHANICAL COMPONENTS AND CORE SUPPORT STRUCTURES EVENTS 14ADING COMINATIQ4 I '

ACCEPTANCE CRITERIA Design 0"I I" PD+TD*ID NO N+T Level A OT N + T + TSVC (3) Level B N + T + SEYg + SRVD I'"'I O NO + WE N+T+WEg + M F) Level B OT + 28 N + TSVC + W Eg + WED+T Level B N + SRVg + SRVD I* D* ' " '

SBA N + SEVg + SRL Level C (5) 10T N + SEV g Level C (5)

IB A o r SB A + W E N + SRVg+Ug + B1, (or E.) s.netD W OT + SSE N + TSVC + SSE 1.evel D (5)

I IBA or SB A + SSE N + SRVg + SSEg + IBL (or SBL) Level D (5)

NLP , N + SEVg + TSVC Level D {5)

N + SEV7 + SSEg ' Level D (5) 12 A + SSE N + SRVg + SSEg + IRL Level D (5)

Teat P,. + IL g. NB 6000 (1) Faire SRV, GIE, SSE, SBL, IBL, and IRL are used, this assas the component or structure response to the Pescsor Building Vibra tion (RBV) dae to these events.

l (2) See Table 3 for the specat c SRV sctuation cases which combine la each <

event.

l (3) TSVC loads apply only to the usin steam piping sad enponents to or mounted thereon.

(5) For active essential valves, the pressaro for which the valve must cpet or close shall not exceed the valvo desism pressnee rating.

(6) Service lieles per ASME Se. tion III Subsection NB or NG.

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TAh!E 2.1(b) CLASS 2 AND 3 MECANICAL COMPONSTS I

m'13 LOADING COMBINATION l' ' ACCEFTANCE CRIIT'.IA 3' I

Design Design P3+TD+RD NO N+T Level A OT N + T + TSVP (3) Lnel B l N + T + SRVg + SRVD 1,nel B

, NO + OBE h + T + OBEY + OBEp Level B

OT+WE N + T3VC + CBE Level B 7

N + SEYg + TEy Level B SBA N+ ,'y + SBl. Lnel C h)

IOT N + SEV g L uel C (4)

IBA or SB A + WE N + SEVg + CBEg + IBL (or SBL) Level D (4)

OT + SSE N + TSVC + SSE g Level D (4)

N + SEVg + SSEg IB A or SBA + SSE , N + SRVg + SSEg + IBL (or SBL) Level D (4)

NLF N + SEYg + T3VC Level D (4) l IRA N + $1Vg + SSEg+!.E Lnel D '4)

Toat P+D NC 6000/

. T

. NU 6000 (1) There SEV, OBE, SSE, SBL, I3L, sad LBL are used, thle acans the caponent or structure response to tae Reactor Building Vibration (RBV) due to these events.

(2) See Table 1 for the specific SEV actuation cases which combine in each event.

(3) TSVC loads apply only to the main stern piping and components to or s.ounted thereon. ,

g4) For activa essential pumps and valves tbe Design and Level B criteria 1i5s11 be satisfied except that she ba sis; allos ble stress shall be 1.2 times the design allowable stress.

l (5) Service Limits per ASME Code, Sectiou III, Subsections NC and ND.

i mfd SOFA (pty.00tet*

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IU5tNE55 QPERATION5 GENER AL h ELECTRIC arv 2 55 TABLF. 2.2 14ADING COMBINATIONS AND ACCEPTANCE CRITERIA i

NON- ASME MECTIANICl.L COMPCNENTS AND ELECTRICAI- CONTROL AND INSTRUMENTATION EDDIPMD4T WORIlhG STRESS men!OD EV1W W I.4AD ACCEPTANCE CRITERIA NO D+L S D+L+mE S OT D + L + SRV S NO D + L + T, + R, S D+L+T +R + WE S o o

. OT D + L + T, + R, + SRV S OT + EE + SRV D+L+T o +R o + T E + SRV 1.2

~

OT + Sd2 + SRV D+L+T 0 +R 0 + SSE + SRV 1.!S or 1.2Y but

' O.7 ITri OT + TB A + W E + D+L+T a

+R e + CB E + P + 1.5S or 1.2Y but SRV (6) a SRV + Y) + Y, + Y, ( 0.7 ITIS OT + SSE + G A D + L + T, R, + P, + SSL + T3+ 1.5S or 1.2! but Y, + Y, + bxV < 0.7 UT5 (1) Yhere SRV OBE. SSE. SBL IBL and GL are used. *his means the component or stucture response to the Reactor Ballding Vibration (RBV) due to tnese events.

(2) See Table 3 for the specific SRV actuation cases which combine in each event.

• (3) nermal combina tima a ssume that thermal stresees Jr.e to T* and R* are secondary and sel f-limiting.

(4) S is allowable stress from AISC Part 1.

(5) Locs1 buckling shall be considered in region s of compressive itsess.

(6) Uso higher of SBL er IEL.

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7 NUCLEAR fMERGY 3 86 aA931 mo IUSINE55 CPERAT!ONS GENER AL h ELECTRIC may 2

56 TABLE 2.3 LOADING COMBINATIWS AND ACCEPTANCE CRITERI A NON-ASME KI'Q ANICAL COXPONFNTS AND ELECTRICAL CONTRM. AND INSTRUMENTATION EQUit' KENT STRENGIB DESIGN METBODS 3 I4II3 EVDiTS LOAD ACCEPTANCE CR!II:RIA NC 1.7 D + 1.7 L Y OT 1.7 (D + L + SEV) Y NO 1.3 D + 1.3 L + 1.3 T, + 1.3 R, Y OT 1.3 D + 1.3 L + 1.3 T6 6 1.3 R0 + Y 1.3 SRV 07 + ODE + SRV 1.3 D + 1.3 L + 1.3 T, + 1.3 R, Y OT + CBE + SEV 1.7 D + 1.7 L + 1.7 OBE + 1.7 SRV Y OT + S S E + S RV D + L + T, + R, + S S E + S EY 0.9Y OT + IB A + OG E + D+L+T a +R a + 1.2 5 P a + ':j + 0.9Y SEY Y, + Y, + 1.23 GBE + 1.2 5 SRV OT + SSE + IR A + D + L + T, + R, + P, + Y, + Tj+ 0.9Y SRV Y,* SSE + SRV L. _

(1) There SRV, ODE. SSE. SBL, IDL, and 12L are used, this aesna the component or structure response to the Reactor Building Vibration (RBV) due to these events.

(2) See Table 3 for the specific SEV actuation cases which combine in each event.

(3) Thercel loads can be neglected when it can be shcwn that they are secondary and self-Itatting in nat ure and where material is ductile.

(4) Y ir yield stress free A15C ? art 2.

($1 a*ctential for buckling shall be considered in regions of ccupressive s'ress.

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G EN ER AL h ELECTRIC IU$1 NESS OPERAT10h5 arv 2 TABLE 2.4 Deleted TARIE 2.5 Deleted T.*R LE 2.6 14ADING CCieINATIONS AND ACCEPTANCE CRITERIA ASME SECTION III COMlU49fT SUPPORTS EVENTS LOADING COMl! NATION (I ' ' ACCEPTANCE CRI1T.RIA( I

~ ' ' '

De si gn PD+D+RD* D Destga NO P, + D + T + R Level A OT P, + D + T + TSVC + R ( 3 )

Level B P, + 3 + T + $dVg + SWD+k Level B NO + WE P, + D + T + W Eg + WE3 Level B OT+WE P,

• 0 + *iSVC + W Eg + WED+T+R Level B P, + D + Stig + SEVD+WEg+ Leest B OBED+T+R

- 2A Pp + D + S RV7 + SEVD + GP '., + T + 2 Level C 10T Pp + D + SRV7 + SRVn+T+R Level C IB A or SB A + WE Pp + D + S EVg + (B Eg + 1BL (or Level D S3L) +2 OT + SSP, P, + D + 13VC + SSEg+R Level D IB A or W A + SSE P, + D + S RVg + SSEg+R Level D P, + D + SRVy + SSEg + IBL (or tecel D SBL) + R NLP P, + D + S RYg + T5TC + R Level D 1.B A + SSE P, + D + SRV7 + SSEg + LB L + R Level D Test P, + D t +R t See Notes os following page.

co se u inev..., o C.523 22 I _ E_TAJJ rr _Y

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Note s for Table 2.6 (1) %ere SEV, OBE, SSE SBL, IDL, sad 1.BL are used, this neans the component or structure response to the Reactor Building Vibratica (fJ1V) due to these ev.ats, I

(2) See Tablo 3 for the specific SRV actuation cases which combine in each event.

(3i TSVC : cede app' only to the main steam piping and components to or mounted

, thereca.

(4) Deleted

($) Service Limit s per ash'E Section III, Subsection NF, P

9 me 4

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?

TABLE 2.7 LOADING COMBINATI0t's AND ACGPTANG GITalA CCNGETE CONTAINMDiTS EVDrI3 LOADING COMBINATICH ACGPTAWG CRITERIA (2)

NO D + L + T, + P, + R, S OT D + L + T, + SRV + R, + P, S OT + OBE D+L+* + OBE + SRV + R, + P, S NO + CBE D + 1.3 L + T, + R, + P, + 1.5 CBE D NO + ENY D + 3 .3 L 4 T, + R, + P, + 1.5 Y U D + L + T, + P, + T ' + R, U OT + SSE D + L + T, + S S E + S RV + R, 4 P, U OT + DBA (1) + L + L + T, + P, t x, + SSE + Y, + U SSE T j + S RV( 3 ) + Y, OT + DBA (1) D + L + T, + 1.5 P, + R, + 1.2 5 SRV U OT + DBA (1) + D + L + T, + 1.25 P, + R, + 1.2 5 U OBE OBE + 1, & Yj + T, + SHY

• OT + DBA (1) +

n & L + T, + 1.25 P, + R, + 1.25 Y U ENV

" ,+ Y, + T, + SRV NO + CCE D+L+ + CBE + T 0 Post Accident '- S NO + ENV Post Accident D+L+FL+W+T 0 S a

Test D, + L + Pt+T t 3 (1) DB A = Mo s t severe of LRA IBA, or SBA.

3 (2) U = Sectica strength based on the strength Design Methods given in ASME Section III Division 2 Article CC-3420.

S

= Required section strength based on the design methods given in ASME Section III Division 2 Subarticle CC-3430.

(3) See Table 3.

meo e na t.ev. i.ien t' '_ X C ? T

• V W , z u 2 . x = m + :_ a C -

_ . mr =c,;:: e wa ___ _ ;- . _ _ _ . -_

4 MUCLEAR ENERGY 3868A931 s a o. 60 BU51NkSS OPERATIONS GENER AL @ ELECTRIC arv 2 j

TABLE 2.8 LOADING COMBINATIONS AND ACCEPTANCE CRTTI2IA STEI2, C0hTAINELNIS II EVENTS LOADING COMBINATICN ACCEPTANCE CRITERIA I2 Design D + L + T, + P, + R Design NO D + L + T, + R Level 4 (Tr D + L + T, + R, + S RV Level A OBA (3) D + L + T, + L, + P, Level A NO + OHE D + L + T, + R, + CEEg + OBEp I,nel B 1:0 + SSE D + L + T, + SSEg  ?,nel C OT + OBE D + L + T, + R, + SRV + OBE Level B OT + SSE D + L + T, + R, + SRY7 + SSEg Level C DBA (3) + CBE D + L + T, + R, + P, + DB A + OBE + Level B SRV (6)

DBA (3) + SSE D + L + T, + R, + P, + S SEg + DIA Level C DB A ( 3) + SSE D + L + T, + R, + P, + SS Eg + SEYg Lon1 D (7)

+ n9A + Y +Y +Y ,

LOCA Post Accident D+L+FL + CSE (4)

Flooding Test Dg*L+P g +T (5) t

( __

(1) MC Caponents such as the drywell head, down comers, equipment hatches, etc, and a liner subjected to mechanical loads condition.

, (2) Service Limits as defined in ASME Section III, Subsection NE.

1 (3) DBA = Most :levere of LRA, IEA, or SBA.

(4) For Post Ac tident flooding Pa 1.5 S,; Pg or P g+PB 1*88 mor 1.5 S 7

, ($) Requitements per ASKE Section III = Article FS-5000.

l l (6) See Table 3.

(7) For local effects only.

so sera inn. ionn

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.- - _- , ---. , . . , , . , . . ~ ~ . .

o

- MUCLEAR ENERGY 388 m 32 5m 61 BU$1NE$$ OPERATIONS GENER AL @ ELECTRIC may 2

i TABLE 2.9 LOADING CON.1INATIONS AND ACCEPTANCE CRITI21A CONCRETE STRUCTURES OTER TB/.N CONTAINMENT EVENTS LOADING CONBINATION ACGPTANCE CRITERIA (II Normal Operation 1.4 D + 1.7 L U 1.4 D + 1.7 L + 1.7 SRV U 1.4 D + 1.7 L + 1.9 ODE U 1.4 D + 1.7 L + 1.7 F U i

1.4 D + 1.7 L + 1.9 OBE + 1.7 SRY U 0.75 (1.4 D + 1.7 L + 1.7 T8+ U 1

1.7 R,)

0.75 (1.4 D + 1.7 L + 1.7 T0+ U

, 1.7 R, + 1.7 SRV) 0.7 5 (1.4 D + 1.7 L + 1.7 T, + U 1.7 R, + 1.9 OBE + 1.7 SRV) 0.75 (1.4 D + 1.7 L + 1.7 T0 + U 1.7 R, + 1.7 T)

Abnorm al/Ex t e tuo D+L+T +0 2 +0 SBV U Fevironmental Condittoas (1) D + L + T, + R, + SSE + SRV U D + L + T, + R,, + 3 RV + W ' U i

D + L + T, + F., + 1.5 P, + 1.2 5 SRV U j D + L + T, + d, + 1.2 5 P, + U 1.2$ CBE S2V + Y, + Yj + Y, +

D + L + T, + R, + P, + (SE + Y, 4 U Y) + Y,+ SRV l i

i l (1) U =_ Section strength based on strength Desism Methods as defined in ACI-318-71.

I 1

Nto 907 A (NEv.16/016

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, f ,

4

.-, -, ,.- . . -,m,-.. - , , ,, , , _ . , _ -

4 NUCLEAR ENERGY IU$lNE$$ OFIRATIONS G EN ER AL h 8I.ECTRIC Rtv386 m2 1

- sm 62 TABLE 2.10 LOADING COXBINATIONS AND ACGFTANCE GI1DIA S1T.EL STRUCTURES OIIIER THAN CONTAINMENT EVENTS LOADING COMBINATI(N ACCEPTANCE CRITI2IA(3I Normal Operation a. Working Stress Design Methods (1)

NO D+L S NO + ENV (SEIS) D + L + W (or OBE) S OT D + L + SRV S If thermal stresses dse to T, and R, are present and are secondary and self-limiting la nature, the f ollowing combina-tions shall also be satisfied.

I NO D + L + T, + R, 1.3 S OT D + L + T, + R, + SRY 1.3 S OT + SEIS D& L + T, + R, + OB E + S RV 1.5 S NO + DTV D + L + T, + R, + W 1.3 S

b. Strength Design Methods (2)

NO 1.7 D + 1.7 L a NO 1.3 D + 1.3 L + 1.3 T, + 1.3 R, Y OT 1.7 D + 1.7 L + 1.7 SRV Y OT 1.3 D + 1.3 L + 1.3 T, + 1.3 R, Y

+ 1.3 SRV NO + ENV 1.7 D + 1.7 L + 1.7 Y Y NO + ENV 1.3 D + 1.3 L + 1.3 T, + 1.3 R, Y

+ 1.7 Y NO + Div 1.7 D + 1.7 L + 1.7 OBE + 1.7 SRV Y OT + GE 1.3 D + 1.3 L + 1.3 T, + 1.3 R, Y

+ 1.3 OBE . + 1.3 SRV (1) Use Part 1 of AISC (2) Use Part 4 of AISC

( , S = Required Section strength based on Workims Stres: b-st n h ttcds (4) Y = Section strength based on Strength Design Methods ato sera usv. imo

'I_ [

. _ ~ ' E}lE _ _7_+ 2 l J _1_~

~~.__' ,L' T __ l ' Z J k C .~I b L '_ Y _ ,[ -

- . . . ........m .. m._m-- - - - - - - - - - - " ' " ~ ' - - - - ~ " " ' " ' - " ' " ' - ' - ^ ' ' " ' - " ' '

o NUCLEAR INERGY 386EA931 suo, 53 BU51N155 OPERATIONS GENER AL @ ELECTRIC nty 2

i TABLE 2.10 (Continued)

EVGTS LOADING COMBINATIm ACCEPTANCE CRITERIA I3I Abnornsl/Estreme Enytrommental Conditions (5)

a. Working Stress Design Methods (1)

OT + SSE D+L+T

• E, + SSE + SRV 1.6 S No + mV D + L + T, + R, + I' 2.6 S OT + DBA D + L + T, + R, + P, + S EY 1.6 S OT + D8A + CBE D + L + T, + R, + OBE + P, + 1.4 S SRV + (Yj + Y,+ Y,)

4 OT + DB A a nd S SE D + L + T, + R, + S SE + P, + 1.7 S SRV + (Y) + Y,+ Y,)

{ b. Strength Design Methods (2) l OT + SSE D + L + T, + R, + SSE + SRY 0.9 Y

, NO + ENV D + L + T, + R, + V ' O.9 Y OT + DBA D + L + T, + R, + 1.5 P, + 0.9 Y 1.25 SRV OT + DBA + CBE D + L + T, + R, + 1.25 CBE + 0.9 Y 1.2 5 P, + SRV + (Y) + Y, + Y,)

OT + DEA + SSE D + L + T, + R, + P, + ( Yj ' + Y, +

Y,) + SSE + SRV (1) Use Part 1 of AISC

. (2) Use Part 2 of AISC I (3) S = Required Section strength based on Vorking Stress Design Methods i

l (4) Y = Section strength based on Strength Design Methods i

(5) Thermal loads can be neglected when it can be shown that they are sscondary and self-limiting in naure and where the materisi is ductile.

I I

reso set A Inev. toise t AO ' 1EA J . =' . A

' ' L} ' [ _ 2_L K '[ '_Q

• '},, "

3 m ==

}1 '

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IU$1 NESS OPERATIONS GENER AL h ELECTRIC Rev 2 s a o.

TABLE 2.11 14AD CO21 NATIONS AND ACCEPTANCE CRITT.RIA REFUTLING EQUIINDfT 1 (2)

EVENTS LOADING CO2 INATION ACCEPTANCE CRITERIA NO D+L+T, Level A NO + UT D + L + T, + SRV7 + SEVD Level B NO + WE D + L + T, + W Eg + WED LI N NO + G E + OT D + L + T, + SEV7 + SEVD+ Eg+GQ Lent & t;,

NO + SBL D + L + T, + SBL + P, Level C NO + 2L + WE D + L + T, + P, + SBL + W Eg Level D NO + IBL D + L + T, + P, + DL Level D NO + LBL D + L + T, + P, + LDL Level D NO + SSE D + L + T, + SSEg Level D NO + SSE + OT D + L + T, + SKVg + SSEg Level D NO + SSE + SBL D + L + T, + SB L + S S Eg Level D NO + NLF D + L + T, + SRVg + SSEg Lent D NO + SSE + IBL D+L+T e +P a + IBL + SSE I Level D NO + SSE + IRL D + L + T, + P, + IF + S S Eg Level D (1) There SRV, WE. SSE, SBL, IBL, and 12L are used this seams the component or structure response to the Reactor Building Vibratica (EBV) due to these events.

(2) See Table 3 for th6 specific SRV actuation cases which combine in each 3 event.

(3) No f atigue analysis required.

(4) See Sheet 65 for Note 4 NEO GS F A (ferv. B e!gg g l

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nuci.e4a swuer GENER AL @ ELECTRIC 2 OU$1 NESS OPERATIONS nev Notes for Table 2.11 (Continued)

(4) Acceptsace Cri teris Code Stress 1.evel A Level B Level C Level D AISC, and Arial 0.60:7 0.80sY 0.803T 1.2:Y but (0.7 A3ME Sub-section NF Bending 0.66:Y 0.88:Y 0.88xf zUTS Steel Alnalson Asist 0.60xY 0.80rY 0.8037 0.80rY ANSI B31.1 Reading 1.0xS h 1.15xS h 1.20:Sh 1.2sF but (0.7 l 7

xUTS Tith the exception of the Inclined Fuel Transfer Systes, reiseling and servicing equipsest use industry standards to define design ilmits, such ss AISC code.

ASMB Subsection hY, or crane design standards. The ASME NC and ND sections have been referenced to be used as a backup whom other codes do not apply or where probablistic allowables are not available. Part of the IFTS (the Containment Isolstica Assembly) la classified as ASkE Code Class NC and the corresponding service levels apply. The balance of the tube is classified as ANSI B31.1.

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TABLE 3 SRV ACTUATION FOR VAR!.OUS EVENTS a m

4 z2 KEACTOR( )

Em CE0ilPS Of SRV ACTUATED BUILDING h VI6 RATION z p ILANT EVENTS m '

ONE TIM) Alli lim SET ADS A1.L REV 2-[

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X X X $,

if Or X X

X X X X X gQ L; 10r X X O

NLF X X X X X X X X M e.

(Inside) X -

X X O Inadv. Ans (ShA)

I LBA (Feedwater) (1) X (2) X (1) X X LBA (Inside. Containment)

X X X X X LliA (Cutside Containment)

X X X (2) X

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IBA (Ins!de Containment)

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(1) One SRV is assumed to fail open for containment analysfu only.

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i (2). AD!. Actuation may coincide with main vent chug ging af ter the peak dynamic loads f rom -

the break have dissipated.

'I (3) DilA and hRV events cause a vibration which mt st be considered in the Icading on all cosaonent., in the reactor building, y

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NUCLEAR ENanov '"" A $" * "

bullNEIS OPERATIONS GENERAL h ELECTRIC My 2 TABLE 3.1 IUTAL NUMBEk 0F RI2.TEF VALVE ACTUATICNS IN 40 MARS Numbes cf Valves Lifting BFt/4 DYR/ 4 IIvR/ 5 BTR/ S BTR/ 6 Slauttaneossir Israet Rock gg Crosby (2) Lo-Lo se t(3) All othere All Masy 1100 1100 753 (220)(4I 1100 220 (1; )I4I 2 1550 85 0 -

85 0 -

1 8750 4750 2550 4750 1580 (1) Shorsk a, Feral-2, Hatch-2, Linarick 1 & 2 Hope Creek 1 & 2 (2) Caorso. Chinskaa 1 & 2, Snoquahanna 1 & 2 i

(3) La Salle 1 & 2 (4) Values in () are for vessel and piping values i

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s um31 s!= sa SENER Ath ELECTRIC NUCt. EAR FNERGY DIVISION pg 2 TABLE 4.1 LOADS FOR NORMAL OPERATION (NO) MIL INFREQUENT OPERATIONAL TRANSIENTS (IOT)

.ORMAL OPERATION ior D _L P I R FIV ShV f cgl) ,

REACTOR BUILDING GENEPAL X X X X X X SUPPRESU 7N POOL REGION X X X X X X RV SHIELL WALL ANNULUS REGION X X X X X VEIR VALL ANNULUS REGION X X X X X CONTAINMENT POOL REGION X X X X X CONTAINXENT X X X X X SHIELD BUILDING X X X X X F1EL BUILDING GENERAL X X X X h SPENT FUEL POOL REGION X X X X X X AUXILIARY BUILDING GENERAL X X X X X X STEAM TIJNNEL REGION X X X X X X C0h7ROL BUILDINO C:".iERAL X X X XlX L RADWASTE DUILJING I GENERAL X X X X X X  !

DIESEL GCIERATCR BUILDINGS GCiERAL X X X X X l t

(

General Note: 'In addition, for buildings on co= son basenats, additional loads r.ay be imposed on any particular 1uilding irom another building on the common baser.at.

(1) Fed = f rces resulting fr a cask drop.

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NUCLEAR ENERGY DIV.SiON nev. 2 TABLE 4 2 LOADS F0h 01'EluTIOFAL TR.VSIENTS 0'ERATIONAL TRANSIENTS D L!P, T, R, SRV S FIV RVC0 RSV REACIOR BUILDING CENEAAL 1 X X X X X X X X SUPPn4SSION POOL REGION X X X X A X i X

RY SHIELD WALL ANNULUS REGICN X X X X X '

X VEIR WALL ANNULUS RECION X X X X X X CONTAINMDrr POOL REGION X X l; X X X CC NTAISME.Yr X X X X X X X SHIELD BUILDING X X X X X X X FUEL EUILDING CENERAL X X X X X SPENT FUEL POOL RECION XX X X X AUXILIARY BUILDING CENERAL X X X X X X STEAM TUNNEL RECION X X X X X X X RADWASTE BUILDING CE'IERAL X X Y Y X. X DIESEL CENERATOR BUILDINGS CD ERAL X X X X X (1) See Table 3 for events resulting te SRV actuation.

(2) Applicable only to mechanical components a:*d electrical components in a flow system.

(3) See Note (0), Table 3.

Genera.L Nota In addition, for buildings on cucumen basemats, additions 1 loads may be imposed on any particular building'from another building on the common basemat.

Thu RPMS loaditig condition is not applicable to cl4e Bf'R 6/ Mark III

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GEN ER AL @ ELECTRIC NUCLEAR ENERGY DIVlslON 2 TABLE 4.3 LOAD 3 FOR ENVIP.ONMENTAL. SEISMIC, AND TEST CONDITIONS ENVIRONMENTAL SEI3MIC TEST W W' PMF & H CBE SSE SL T P,

REACTOR BUILDING CE:lERAL X X SUPPRESSION POOL X X X ,

TV 3HIELD WALL ANNULUS X X WEIR WALL AN.4ULUS X X CONTADetENT X X X X SHIELD BUILDING X X X X X X X CONTAINMENT POOL X X X TUEL DUILDING GENERAL X X X X X X X SPENT ITEL POOL REGION X X X

, AUTILIARY BUILDING G E""* * * ~

X X.  ; ;; X X

. CONTPOL BUILDING GENERAL X X X X X X X RADWASTE BUILDING CENERAL X X X X X X X DIESEL CENERATOR BUILDINGS LCfERAL X X X X X X X General Note: In addition. for builcings on cocoon basemats, additsonal loads may be imposed on sny particular building f rom another building-en the com:non t ascsat.

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b TABLE 4.4 LOADS FOR DESIGN BASIS ACCIDENTS N DES 1CH BAS 1S ACC1 DENT (DBA)

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T a

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Y j Y e

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VLC CO QiUG REV

' L SD WCO SL PS fg3 REACTOR BUILDING I 3',

j r=

X X X 3 1 X X X X iX r.

CENERAL SUPPRESSION POOL RECION X X X X X X X X X X X I4)X S' RV SHIELD WALL ANNULUS REGION L 1 X I, X X X ,, ., I X X $

WEIR WALL ANNULUS REGION 1 X X X X X X X X X(4)

Xg4)X g%

X X $ syg CONfA110'DIT POOL REGION I X X I X X X X X X X X 2O CONTA120iENT X X E Sil1 ELD BUILDING O

FUEL BUILDING CENERAL X X X X X X SPENT FUEL REGION X X X AUXILIARY BUILDING X X X X X X X{3)Xg4)

CENERAL CONTROL BUILDING CENERAL RADWASTE BUILDING g)

CENERAL X X X X X X X ^q w

n

DIESESL CENERATOR BUILDINCS X X X X X X X CENERAL 1 =

(1) See Note (3), Table 3 (2) See Table 3 for events resultina in SRV actuation

0) Applicatle to Main Steam System only y

^

(4) Applicable to tanks with a frpe water surface j Note: -Tim sig and loads are obtained from the Containment Load Report. Reference A42-5400 y General Note: In addition, for buildings on co:acion basents, additional loads say be l 4 im;>osed on any particular buildia.g from another building on the common

. basemat.

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I GENER ALM ELECTRIC nev. 2 FINE NUCLEAR ENERGY OlVISION TABLE 4.5 LO/TJS FOR SMALL B'. TEAK LOCAs SBL CHU; SRV P, T, VLC VC0 l RE[

react 0R 2UILDING X X X GTNERAL X X X X SilPPRISSION POOL REGION X X X X X 3.' SKEILD WALL /.NNULUS REGION X X X X X X bEIR WALL Ah3ULUS RECION I X X X X

%NTAlh?1VI POOL REGION i X X X X CONTAlhM.E.VT REGION X X X !X SHIELD BUILDING l

i

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TJEL BUILDING GDEPA. X X (

X X , p.

SPUT TUEL POOL REGION AUXILIARY BUILDING X

.X X CU ERAL i

f CONTROL BUILDINO i GDERAL RA.%'AST E GD ERAL X X DIESEL GCERATOR BUILDINGS X ~X GENERAL II) Applicable to Kain Stean. System only, i General %ote: Additional loads may. be transmitted f rom other butidingn.

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