Text

Proposed Tech Specs Re Administrative & Editorial Corrections
	ML18102B675
Person / Time
Site:	Salem
Issue date:	11/14/1997
From:	; Public Service Enterprise Group
To:	;
Shared Package
ML18102B674	List: ML18102B673; ML18102B675;
References
	NUDOCS 9711210123
	Download: ML18102B675 (139)
	v • d • e

Document Control Desk

Attachment 3 LR-N970150 LCR S97-08 SALEM GENERATING STATION UNIT NOS. 1 AND 2 FACILITY OPERA TING LICENSES DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 CHANGE TO TECHNICAL SPECIFICATIONS ADMINISTRATIVE AND EDITORIAL CORRECTIVES TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specifications for Facility Operating License No. DPR-70 are affected by this change request: Technical Specification INDEX 2.2.1 B 2.1.1 3/4.1.1.1 3/4.1.1.4 3/4.1.2.6 3/4.1.3.1 3/4.1.3.4 3/4.1.3.5 3/4.2.1 3/4.3.1.1 3/4.3.2.1 3/4.3.3.1 97-11.210123 971114 PDR ADOCK 05000272 p PDR 3 Page I, Ill, V, VI, VIII, IX, XI, XII, XIII, XIV, XV, XVII, and XVIII 2-6 2-7 B 2-2 3/4 1-1 3/4 1-5 3/4 1-16 3/4 1-18 3/4 1-22 3/4 1-23 3/4 1-25 3/42-2 3/4 2-3 3/4 3-11 3/4 3-12 3/4 3-22 3/4 3-34 3/4 3-35 3/4 3-36a 3/4 3-37 3/4 3-38a

Document Control Desk Attachment 3 LR-N970150 LCR S97-08 SALEM GENERATING STATION UNIT NOS. 1 AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 CHANGE TO TECHNICAL SPECIFICATIONS ADMINISTRATIVE AND EDITORIAL CORRECTIVES TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specifications for Facility Operating License No. DPR-70 are affected by this change request: (continued)

Technical Specification 3/4.3.3.7 3/4.3.3.8 3/4.3.3.9 3/4.3.4 3/4.4.1.4 3/4.4.5 3/4.4.6.1 3/4.4.6.3 3/4.4.8 3/4.4.9.3 3/4.4.10.1 3/4.5.3 3/4.5.5 3/4.6.1.1 3/4.6.1.2 3/4.6.1.6 3/4.6.2.2 4 Page 3/4 3-53 3/4 3-56 3/4 3-57 3/4 3-57a 3/4 3-63 3/4 3-68 3/4 3-71 3/4 4"'.3b 3/4 4-8 3/4 4-14 3/4 4-16a 3/4 4-16b . 3/4 4-16c 3/4 4-20 3/4 4-31 . 3/4 4-32 . 3/4 5-6 3/4 5-7 3/4 6-1 3/4 6-2 3/4 6-8 3/4 6-10

Document Control Desk Attachment 3 LR-N970150 LCR S97-08 SALEM GENERATING STATION UNIT NOS. 1 AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 CHANGE TO TECHNICAL SPECIFICATIONS ADMINISTRATIVE AND EDITORIAL CORRECTIVES TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specifications for Facility Operating License No. DPR-70 are affected by this change request: (continued)

Technical Specification Page 3/4.6.3.1 3/4 6-13 3/4.7.9 3/4 7-31 3/4.7.10 3/4 7-33 3/4 7-34 3/4.8.2.3 3/4 8-8 3/4.9.7 3/4 9-7 3/4.9.8 3/4 9-8 3/4 9-8a 3/4.9.10 3/4 9-10 3/4.10.1 3/4 10-1 3/4.11.2 3/4 11-10 B 3/4.4.8 B 3/4 4-5 B 3/4.4.9 B 3/4 4-6 B 3/4.7.9 B 3/4 7-6 5.4.1 5-4 6.12.2 6-28 5 Document Control Desk Attachment 3 LR-N970150 LCR S97-08 SALEM GENERATING STATION UNIT NOS. 1 AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 CHANGE TO TECHNICAL SPECIFICATIONS ADMINISTRATIVE AND EDITORIAL CORRECTIVES TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specifications for Facility Operating License No. DPR-75 are affected by this change request: Technical Specification Index 1.0 2.2.1 3/4.0 3/4.1.1.1 3/4.1.1.3 3/4.1.3.1

3/4.1.3.4 3/4.1.3.5 3/4.2.1 3/4.3.1.1 3/4.3.3.1 3/4.3.3.7 6 Page I, IV, VIII, IX, XI, XII, XIII, XIV, XVII, and XVIII 1-2 2-4 2-6 3/4 0-1 3/4 1-1 .. 3/4 1-4 3/4 1-13 3/4 1-19 3/4 1-20 3/4 1-22 3/4 2-2 3/4 3-7 3/4 3-11 3/4 3-12 3/4 3-38 3/4 3-39a 3/4 3-40 3/4 3-50 3/4 3-51 3/4 3-51 b 3/4 3-52 3/4 3-52a

Document Control Desk Attachment 3 LR-N970150 LCR S97-08 SALEM GENERATING STATION UNIT NOS. 1 AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 CHANGE TO TECHNICAL SPECIFICATIONS ADMINISTRATIVE AND EDITORIAL CORRECTIVES TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specifications for Facility Operating License No. DPR-75 are affected by this change request: Technical Specification Page 3/4.3.3.8 3/4 3-58 3/4.3.3.9 3/4 3-63 3/4.3.4 3/4 3-66 3/4.4.1.3 3/4 4-3 3/4.4.7.1 3/4 4-16 3/4.4.9 3/4 4-23 73/4.4.11

.1 3/4 4-33 3/4.5.3 3/4 5-7 3/4.5.5 3/4 5-9 3/4.6.1.1 3/4 6-1 3/4.6.1.2 3/4 6-2 3/4.6.1.6 3/4 6-8 3/4.6.2.1 3/4 6-10 3/4.6.3 3/4 6-15 3/4.7.1.5 3/4 7-10 3/4.7.9 3/4 7-26 3/4.7.10 3/4 7-28 3/4 7-29 3/4.8.2.1 3/4 8-8 7

Document Control Desk Attachment 3 LR-N970150 LCR S97-08 SALEM GENERATING STATION UNIT NOS. 1 AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 CHANGE TO TECHNICAL SPECIFICATIONS ADMINISTRATIVE AND EDITORIAL CORRECTIVES TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specifications for Facility Operating License No. DPR-75 are affected by this change request: Technical Specification Page 3/4.8.2.5 . 3/4 8-14 3/4.9.7 3/4 9-7 3/4.9.8 3/4 9-8 3/4 9-9 3/4.11.2 3/4 11-10 B 3/4.4.9 B 3/4 4-6 B 3/4.4.10

B 3/4 4-12 B 3/4.10.3 B 3/4 10-1 5.4.1 5-4 8 .' i Unit 1 Mark-Up Pages

DEFINITIONS se:c=;:oN

EF::NED :'ER.MS ACTION . . . ..\X!AL FLUX DIFFEJU:NCI

HANNEL CALIBRATION ClWmEL CHEC'!C .... CHANNEL FUNCTIONAL TEST CONTA!NMEN'l' INTEGRITY CORE ALTERATION . . . . QOSE EQUIVALEN't' I-lll E-AV'ERAGB DISINTEGRATION ENERGY ENGINEERED SAITrY nATUlU R.ESPONSI TIMS FREQUENCY NOTATION . . . . . . . .

FULLY WITHDRAWN . . . . . . . . .

GASEOUS RACWASTB TREATMENT SYSTEM I CENT IF I ED t.EAJCAG* . . . MEMBIJl(Sl or nm POBLIC (0004) OPERABLB OP PHYSICS TESTS PRESSURE BOUNOAJlY LZAltMi* PROCESS CONTROL PROGRAM (PCP) PURGE -PURGING . . . . . .

QUADRANT POWER TILT RATIO RATED THERMAL POWD REACTOR TRIP SYSTIM RS.HONS* TI!G REPORTABLB rJENT SHUTDOWN MARGIN SITE BOUNDARY SOLIDIFICATION SOURCE CHBCJt . STAGGER.ED TUT BASIS THERMAL i'OWD '.JNIOENTIPim>

UNRESTRICTm AJlBA './ENTILATIOll IXllAUST TRDTMKNT SYSTEM 'lENTING -..............

SALEM -t"NIT

1 I l-l l-l l-l l-l l-l l-l l-2 l-2 l-4 l-l l-l l-l l-3 l-l l-3 l-4 l-4 '. l-4 l-4 l-5 l-5 l-5 l-5 l-5 l-5 l-6 l-6 l-6 l-6 l-6 l-6 l-6 l-7 l-7 l-7 l-7 l-7 Amendment.

No. 1.78

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=OHP!61otfS rgg t,,l{Q

t!Ot': § .. s;mg!I ?'Ci lLU ppt,IWXI.!:X 3 14 t wcnyxn cpmoL 3/4.l.l BOIATIOM CONTI.CL

T > 2oo*r Sbucdawn K&r1in

T'vl s 2oo*r .......

&VI Koderacor T..,eracure Coeff1cienc Temperacure for Cricical1cy 3/4.l.2 BOIAl'lOll S"ISTDCS now Paw

Shucdovft now Paw

Operacina Char11nc Pu-.

ShucdoWll Charline "-'

Operacs.na lartc Actcl ?rwfer ,_,.

SlNcdow lor$c Acid Trand'.1r

.....,,

O,eracint loracecl Source*

SlnacdOVll loracacl Source*

Operact.na 3/4. l. l ICVA.11.1 COll?IOL ASSDGUU QrQ\lf Hal.pc Poeicioa Iadicaciq Sy1c ..

Oparaciq Po1icicm Indicaciq Syac ...

Sllucdiawla R.ocl Drott Ti.M ShucdoWll lo4 tnaarcioa Uaic Concrol loci tnearcion

'-Laiu SAlJX

1 ttI *

J/4 0*1. 3/4 l*l 3/4 1.-J J(_, 3/ .. 1 ** 3/4 l*5 3/4 l-6 l*7 3/4 l*I 3/4 l*lO 3/,. l*ll 3/* 1*1:1 e_, 3/la \*U 3/4 1*14 3/4 1*1' 3/4 l*U 3/4 l* l9 3/4 1*20 3/4 l*2l 3/4 1*22 . 3/4 l*23 wr tnsnc 1lo. 101

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECIION 3/4.4 3/4.4.l REACIOR coot.ANT SXSTQ1 REACTOR COOLANT LOOPS JcNS eeet.ldft e?JtetJUCT:l:Olf:>Q_; Hot St&ndby . Hot Shutdown Cold Shutdown 3 I 4 . 4 . 2 . l SAFETY VALVES -SHUTDOWN 3/4.4.2.2 SAFETY VALVES -OPERATING 3/4.4.3 3/4.4.4 3/4.4.5 3/4.4.6 RELIEF VALVES PRESSURIZER . STEAM GENERATORS REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection system Operational Leakage . . . C.OOLAl'tl'Pressure Isolation Valve* DELETED .3/4 4-l .3/4 4-2 .3/4 4-3 .3/4 4-3b .3/4 4-4 .3/4 4-4* .3/4 4-5 .3/4 4-6 .3/4 4-7 .3/4 4-H .3/4 4-15 .3/4 4-l6a 3/4.4.7 3/4.4.8 SPECIFIC ACTIVITY .................

3/4 4-20 3/4.4.9 PRESSURE/TEMPERATURE LIMITS Reactor Coolant Sy*tem Pressurizer . . . . . .

overpre*aur*

Protection Sy*tem8 3/4.4.lO STRUCTORAL INTBGltITY ASHE Code Cl*** l, 2, and 3 Components 3/4.4.ll INTENTIONALLY BLAN1t 3/4.4.12 H!:AD VENTS SALEM -UNIT l v .3/4 4-24 .3/4 4-29 .3/4 4-30 .3/4 4-32 .3/4 4-34 .3/4 4-35 Amendment No.180

Sic;I!ON l/t ' l/4.5.l )/t.5.2 )/4.5.l )/4.5.* )/ *. 5.5 3 Ii ' l/*.S.l )/ *. §.2 )/ *. §.) l/4.6.4 CJ:!IBQRfCX CCU CQOLI!!J SXSTpl (ICC:U, A.Cc:ti'MUt.\

TOU !CC:S

T"' a l50*P SL\L :tMJZc:TIOM rt.OW . R!IUZ?.INO WAT'IJl STOUGI TANK COlfTAINMllQ' SXSIIN coacaimaenc InC99TiCy concaia.enc I.tlaka9e . Conca1nmenc Air t.ocX. Incernal . . A.ir Tetaperacure . . . conca1nmeAc scruccural IAC99TiCy Conca1nmenc vea.cilacioa Sy*c .. concaiD89AC Spray Syac .. spray Mdieive Syac* . . ConcaizmulC C:ool1D9 Syn* COllTAIJMIWI' ISOU.TtOll VALVU lleccric Bydrofec lecOlllbin*r*

SALDI

OMIT l Vt JI* 5-l ll* 5.3 ll* 5*6 l/4 *-1 11* s-2 3/4 5.5 3/4 ,_, 3/4 6-7 l/4 6*1 3/4 S*ia 3/4 S*9 3/4 6*10 3/4 5*1.l 3/4 6-12 3/4 S*lll 3/4 6*U .

LI!'1ITISG CONDITIONS FOR OPERATION A.'iD SL"RVE:LLA..'iCE C\EQL":RE..'iE:S7S SECTION ]/4.8 ELECTRICAL POWER SYSTEMS J/4.8.l A. C. SOURCES Operating

.. Shutdown 3/4.8.2 ONSITE POWER DISTRIBU'l'ION SYSTEMS A.C. Distribution

Operating

.. A.C. Distribution

-Shutdown 125-Volt D.C. Distribution

Operatinl 125-Volt o.c. Distribution

-Shutdown 18-Volt D.C. Distribution

Operating . 28**. -. : D.C. Distribution

-Shutdown J/4.8.3 ELE: !CAL EQUIPMENT PROTZ -:VE DEVICES

SALEM

UNIT l VIII )/lJ 3/4 8-6 J/4 8-7 3/4 8-8 J/4 8-10 J/4 8-ll J/4 8-13 J/4 8-14 Amendment No. 105

L!MIT!NG CONDITIONS FOR OPERAT!ON ANO REQUIREMENTS SECTION 3/4.9 3/4.9.l 3/4.9.Z 3/4.9.3 3/4.9.4 3/* 9.5 3/4.9.5 3/4.9.7 .. 3/4.9.S REFUEt.IN6 OPERATIONS BORON CONCEMTRATtON INSTRUMEMTATtON

DECAY T!ME ** . . . . . . . . . . CONTAINMENT BUILDING ?EMETRATIONS

- COMMUNICATIONS

MANIPULATOR CRAN . . . . . . . . .. . . CRANE TRAVEt. -FUEl. HANDLING AAEA ** RESIOUAL HEAT REMOVAL .I.HO COOLANT CIRCULATION All W*t1r Levels ****************

Lew W*ter Level ***************

3/4.9.9 CONTAINMENT

?URGE AHO ?RESSURE-VAOJUM REt.IEF ISOLATION SYSTait ****** 3/4,9.10 WATER LEVEL -REACTOR VESSEl. 3/4.9.ll STORAGE ?OOL WAT!R LEVEL *** 3/4.9.12 FUEL HANCLING AREA VENTILATION SYSTEM

3/4.10 SP!CtAL T!ST EXCEPTIONS 3/4.10.l SHUTOOWN MARGIN * * *

3/4.10.2 GROUP HEIGMT, INSERTION ANO ?OWER OISTRIBUTtON LIMITS 3/4.10.3 ?HYSICS r::srs )/ol.10.ol . -.a FLOW TE:S'T'S . . . . . . . 3/4 9-l 3/4 9-2 9-3 3/4 9-.1 3/4 9-5 3/4 9-o 3/4 9-7 3/4 9-8 3/4 9-aa 3/4 9-9 3/4 9-10 3/4 9-ll 3/4 9-12 3/4 10-2 3/1 10-3 3/4 .....

SAL:."I -*u.*HT l IX Amendment No. 59 INt>EX BASES SECTION ?AGE -APPlICAIILITY

* * * * * * * * * * * * * * * * * *

a 3/4 0-1 3/4.l REACTIVITY CONTROL SYS'TlMS 3/4.1.1 SQUTION CCNT'ROL ** * * * * * * * * * * * * * * * *

I 3/4 l*l 3/4. l .Z BORATtON SYST!MS ***** * ****** * * * * * *

I 3/4 l*l 3/4.1.l MOY.AILI CONTROL ASSEMBLIES

I 3/4 l-4 3/4.Z. POWER OISTRIIUT!ON LIMITS 3/4.Z.l 3/4.Z.Z lftd 3/4.Z.l 3/4.Z.4 3/4.2.5 .UtAL FlUl I"£R£NCI a l/4 2-1 l M9f GMMNEL eJ\blAL pEAk.JNb EMT14ALl'f HOT OWINIL-HieTS=9 A.NP ******** I 3/4 z.4 ONI PAUICTERS . * * * * * * * * * * * * * * * * ..

.... ll/L/ SALE11

UNIT l XI Amendment No. 59 BASES =========-=-

--. ..=* =============

3/4.3.l o.'fl4 . I 4. 3. 2 PRA'Ni>'IVE

IWS'tRUMBll:t'ATION NEEREO SAFETY FEATURES ll:':SFl 3/4.3.3 3/4.3.4 MONITORING INSTRUMENTATION TURBINE OVERSPEED PROTECTION 3/4.4 R£ACTOR COOLANI SXSDjM 3/4.4.l REACTOR COOLANT LOOPS AND COOLANT CIRCULATION 3/4.4.2 SAFETY VALVES 3/4.4.3 RELIEF VALVES 3/4.4.4 PRESSURIZER 3/4.4.5 STEAM GENERATORS 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.7 DELETED 3/4.4.8 SPECIFIC ACTIVITY 3/4.4.9 PRESSURE/TEMPERATURE LIMITS 3/4.4.lO STRt1CTOR.AL INTEGRITY 3/4.4.ll BLANlt 3/4.4.12 REACTOR VESSEL HEAD VENTS SALEM -UNIT l XII .B 3/4 3 li_..JC-.8 3/4 3-l .8 3/4 38 .8 3/4 3-4 .B 3/4 4-1 .B 3/4 4-la .B 3/4 4-la .8 3/4 4-2 .8 3/4 4-2 .B 3/4 4-3 .B 3/4 4-5 .B 3/4 4-6 .B 3/4 4-17 .B 3/4 4-17 .B 3/4 4-17 Amendment No. 180

nrnos '" ' gpqgs;x cog cooLIHC sxsipq ( !C:C:!) 3/4 . .5.1 ACcmmt.ATOl.S 3/4. 5.2 and 3/4.S . .:. * -asne 9---Sc"L XN:J"&:..,,T10/lf FLDw 3/4.5.5 IUUILIMC \IA )/It 6 t;O'TAINMM mm1 3/4.6.1 PUKA&Y ccKTAINKlllT 3/4.l.2 DUUSSUUZAtlCll AID COOUIG nsTIMI 3/4.6.3 cOlftAilQCllCT ISOLAtIOll VALVIS 3/4.6.4 CC>>mJSTI11.I CAS coartlOI.

SALDI

UlllT l XIII *

I 3/* 5*1 I 3/4 5*1 'B $-2..1 I I 3/4 6*1 I 3/4 6*3 I 3/4 6*3 I 3/4 : :.

Bt\.5ES grI'IQf 3/4.7 3/4.7.1 3/4.7.2 3/4.7.3 3/4.7.4 3/4.7.5 3/4.7.6 3/4.7.7 3/4.7.8 3/4.7.9 '!URBINE CYCI.E * * * * * * * * * * * * * * * *

8 3/4 7-1 LIMITATiaf B 3/4 7-4 FLOOD PRm:x:::nai

* . . . . . a:ff!R)L RXM AIR a::tIDITiamG SYSTEM **** ADXIL!AR:i EIJIIDDG EXfWlS'l' AIR FIIlmAT!af S'iS'1'!Jil

* * * * * * -.......... .

B 3/4 7-4 ** B 3/4 7-4

B 3/4 7-5

B 3/4 7-5 * * *. * *

B 3/4 ******** B

* * * *

B 3/4 7-6 3/4. a EIETPil2L PQQ. S'iSTD6 3/4.8.l am 3/4.8.2 3/4.8.3 A. C. SCllR:D AND oosrm :EaO oISimll1I'Iai SYS'IDtS *

B 3/4 8-1 PRmX!l'IVE CEVIC!S * * *

B 3/4 8-1 SALDI -UNIT l XIV Amm'dnait No. 139 BASES ==============================================================================

SECTION 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION B 3/4 9-1 3/4.9.2 INSTRUMENTATION B 3/4 9-1 3/4.9.3 DECAY TIME B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS B 3/4 9-1 3/4.9.5 COMMUNICATIONS B 3/4 9-1 3/4.9.6 MANIPULATOR CRANE B 3/4 9-2 3/4.9.7 CRANE TRAVEL -SPENT FUEL STORAGE BUILDING .. B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE AND PRESSURE-VACUUM RELIEF ISOLATION SYSTEM 3/4.9.10 WATER LEVEL -REACTOR VESSEL and AND 3/4.9.11 STORAGE POOL B 3/4 9-3 3/4.9.12 FUEL HANDLING AREA VENTILATION SYSTEM B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4 .10 .1 SHUTDOWN MARGIN .... ....... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS 3/4.10.3 PHYSICS TESTS 3/4.10.4 NO FLOW TESTS SALEM -UNIT 1 xv B 3/4 10-1 B 3/4 10-1 B 3/4 10-1 Amendment No. 192 INDEX ::ESIGN FEATURES ==============================================================================

SECTION 5.1 SITE Configuration

........ . Design Pressure and Temperature 5.3 REACTOR CORE fuel Assemblies Control Rod Assemblies 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature Volume . . . . . . . . .

5.5 METEOROLOGICAL TOWER LOCATION 5.6 FUEL STORAGE Criticality Drainage Capacity .. 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT SALEM -UNIT 1 XVII PAGE . 5-4 5-4 5-4 5-4 5-4 5-5 5-5 576'l 5-f bc:t 5-f Amendment No. 59

ADMINISTRATIVE CONTROLS ====================================================================

6.1 6.2 ORGANIZATI.N Onsite and Offsite Organizations Facility Staff .. Shift Technical Advisor 6.3 FACILITY STAFF QUALIFICATIONS 6-1 6.4 TRAINING .........................

6-7 6.5 REVIEW AND AUDIT (THIS SECTION DELETED) . . . . . . . . .

6-8 6.6 REPORTABLE EVENT ACTION 6.7 SAFETY LIMIT VIOLATION 6.3 PROCEDURES AND PROGRAMS 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS 6.9.2 SPECIAL REPORTS 6.10 RECORD RETENTION 6.11 RADIATION PROTECTION PROGRAM 6. 12 HIGH RADI.l\TION AREA . . 6.13 PROCESS CONTROL PROGRAM (PCP) 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) 5.lS MAJOR CHANGES TO RADIOACTIVE LIQUID, GASEOUS A."ID SOLID WASTE TREATMENT SYSTEMS 6.16 ENVIRONMENTAL QUALIFICATION SALEM -UNIT 1 XVIII 6-16 6-16 6-17 6-20 2.5 6-26 Amendment No. 192

FUNCTIONAL lJNU ll. Steam Generator Water Level--Low-Low

14. Steam/Feedwa*.er Flow Mismatch *nd Low Steam Generator Water Level 15. 16. 17. Undervoltage-Reactor Coolant Pumps Underfrequency-Reactor Coolant Pumpa Turbine Trip A. Low Trip System Pressure 8. Turbine Stop Valve Cloaure 18. s t 19. Reactoc Coolant Pump Breaker Position Trip SALEM -UNIT l IA@LE 2,2-1 (Continuedl REACTQR TRIP SYSTEM INSTRQMENTATION TRIP TRIP SETPOJNI 9.0\ of narrow range instrument span-each steam generator so 40l o.f full steam flow at RATED THERMAL POWER coincident with *team generator water level 11 10.0l ot narrow range inatrument

pan--each steam generator 11 2900 volts-each bus 11 56.S Hz -each bus II 45 p&ig so 15\ off full open Not Applicable Not Applicable 2-6 ALLOWMLE VALVES 8.0\ ot narrow range inetrunient span-each steam generator s; 42.5\ of full steam flow *t RATED THERMAL POWER coincident with steam generator water level 9.0\ of narrow range instrument steam generator 11 2850 volts-each bus 11 56.4 -each bus 11 ts peig s 15\ off full open Not Applicable Not Applicable Amendment No. 159 I

c z ....... -I N I where: Tl TABLE 2.2-l (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION ti T "' Indicated ti T at RA TED THERMAL 0 T "' Average temperature, °F T> ;;: Indicated Ta at RATED THERMAL POWER < 577.9°F vg -p ;;: Pressurizer pressure, psig p> ;;: 2235 psig (indicated RCS nominal operating pressure)

Thf' function generated by the lead-lag controller for T dynamic compensation avg l+T 1S _ l+T 2S -Time constants utilized in the lead lag controller for Tavg Tl "' 30 secs, T 2 "' 4 secs. & T2 S -Laplace transform aper

*

BASES !he curves are bas*d on ui enthalpy hot channel factor, , of L.SS and a reference cosine with of l.55 for axial paver shApe.

is includ*d for an incraa.* ln r'1H at reduced power ba*ed on the expre**ion:

r:H

l.55 (1 + O.J(l*P)]

where ? i:3 the fraction These limitin1 h ... t flWl condition*

are hi er th.an those calculated for t th* range of all control rod* rut.LY WITHDRAWN h* sazillua allowable control rod insertion asauaiz.11 the axial power illlbalance is within th* limits of th* f 1 (6I) function ot the Overtemperature trip. Whan the axial power imbalance ii not within the tolerance, the azial pov.r imbalance effect on the Overtemperature AT trip* will reduce the setpoints to provide protection consistent with core safety liait.. 2.1.2 R!ACTOI COOL.Am' SYSTEM PR!SSURI Th* restriction of this Safety Limit protect* the inte1rity of the Ructo?" Coolant Syst* froa ov.rpruaurization and thenby p't"-*1ents the nleue of radionuclide*

contained in the reactor coolant frOll ruchinl tha containmant ac.moaphere.

The reactor p?"Usur* V"essel and pt'uauriz*r an cte.iped to Section III of the ASMI Cod9 for Nuc:l ... r Power Plant vbic:h penli.ta a aazillull tran*ient pressure of 110% (2735 psi1) of de*ip pressure.

Th9 Reactor Coolant System pipin1 and fittinas are d**isned to ANSI I 31.1 1955 Edition vhila th*

are duigned to ANSI B 16.5, MSS*SP-66*1964, or A.SKI Section III-1968, which pet'mit m.a.zi.mua tran*ient pr .. auru of up to 120% (2985 psi1) of component design pressure.

Th* Safety Limit of 2735 pail is therefore con*istent with th* design criteria and associated code requirements.

The .ntir* Ra&ctor Coolant Syst .. is hydrotastad at 3107 psi1, 125% of design pr*s*ure, to demonstrate inte1rity prior to initial operation

. .... Sal .. -Unit l 8 2-.J. Amendment No. 91 .. ,.

*

l(f,t,1 IQl!Attgw cp1rJtQio

- - ,. ..... .:> -* .. -:-*.,.,.

ua aautW.* NllllDI < 1.R ..,.,..,,,, 1Wd1&t*1Y 1alt1au ud C:Oll,i.A1.1*

M.r*iola ac. a JJ na Of & .. lnioe MnUA.al.luJ

,,StO R19

\IACU ""IUU'ed IBUiDOill ICAlllH* 19

... * --. 4.1.1.1.1 ne amJiNIM llUllU ebal.1 i. u.1m*net be a 1.6' ** 111t!WI .. baur an.

f

1.Dapwllbl*

cmtn1 rocSC*) ... c. .... &1: 1-d aa. ,_ 12 ... wllil* tlle nd(*) 1* 1.Aapuu1**

If i.AaperUle coac:ol nd 1* 1-1wui. H uau-1pt81e, t:ae ....,.. rell'li.nd aauw ICUlllZ* *ball N LAcnaMd tty aa 1 ian. a& leuc equal u wont. of tlle U.OV&Ol* or per U "f "W1f1i.ll9 ta. 1iat.'* of a,.c1f .rod s.11 a ##111tia

< i.o 1#1# 1.a1*

k/k cycl* 11 of

3/t 1-1 *

o. 149

guct1v1n comoL sxrrms !10QEMIOI IT.KPQATYU CQ!FllC!W LIMIT1MC CCIDlTlOI FOi OPERATlOM

3. L 1.4 'nle moderator temperature coefficient (KTC) shall be: a. IA** po*itiv* than 0 delta k/k/*r for the all rod.a vithdrava, beginnin&

of cycle life (IOL), hot zero 'raERMAL POVll condition . b.

4 l.Ass negative than *4.4 z 10 delta k/k/*r for the all rod.I vithdravn, end of cycle life (EOL), RATED mUMAL POVD. condition.

APPLICAIILITX:

Specification 3.1.l.4.a

KOD!S 1 and 2* onlye Specification 3.1.l.4.b

MODES 1, 2 and 3 only# ACTION: a. Vith the KTC iaor* positive than the limit of 3.1.l.4.a, above, operation.

in MODES 1 and 2 may proceed provided:

1. ' 2. 'n\* control are maintained vithin th* withdrawal limits established abpve until a subsequent calculation verifies that the KTC ha* been restored to within it* limit for th* all rods vithdravn condition.

3. A Special is prepared and submitted to the Comhs1oa purSUADt to Specification 6.9.2 vithin 10 days, describing the valua of the measured MTC, th* interia control rod vithdraval lt.its and th* predicted average cora bunwf necessary for rastorin&

th* poeitive KTC to within its limit for the all vithdravn condition.

b. the KTC more negative than th*

of 3.1.1.4.b, above, be in HOT SHU'1'!)0\llf vi thin 12 hour*. *\11th Keff greater than or aqUAl to 1.0 #Sae Special Te*t Exception 3.10.l SALEK

UNIT l 3/4 1-s ... Amendment No. 113

RIAC"l'Mn COl!IBQL SXSTIMS aoum WAID sosmgs -OPERATING LIKITillO

<Dm%'1'IOll POil OPERATION

- - - - ----*******----*--*******-*****-------------*

3.1.2.6 a. Aa a minimull, th* followin9 borated water *ourc*(*)

h&ll be OPSRAB.La

- required by apecificationa 3.1.2.l and 3.1.2.21 A boric: acid atoraqe ayat .. vith: l. A contained volume of borated water in accor 3.1.2, 2. A boron concentration in accordance with and A minimwa aolution tilmperatur*

of 6J*r. b. The refuelinq water atoraqe tank with: l. A contained volume of between 36,,500 and 400,000 qallona ot water, 2. A boron concentration of between 2,300 and 2,500 ppm, and 3. A mini.mum aolution temperature of Js*r. AfPLICA8ILITY:

MODES l, 2, 3 and 4. ACTION: a. With th* boric acid atora9* *y*tmn inopera.bl*

and being uaed a* one of th* a.bove required boration water ayatema, r**tore th* *toraqe *y*tem to OPERABLE atatu* within 72 hour* or be in at lea*t HOT STANDBY within th* next 6 hour* and borated to & SHt1TDOW!I MARGIN equivalent to at lea*t 1\

at 2oo*r; re*tore th* boric acid atoraqe system to OPERABLE statu* within th* next 7 day* be in COLO SHUTDOWN within th* next 30 hour*. b. With th* refueling water storage tank inoperable, re*tore the tank to OPERABLE atatua within one hour or be in at lea*t HOT STANDBY within the next 6 hour* and in COLD SHUTDOWN within the followinq 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.l.2.6 Each borated water source shall ba demonatrated OPERABLE:

SALEM -UNIT l 3/4 1-16 Amendment No. 169 REACTIVITY COHTROL SYSTEMS 3;4 .1.3 /l10vA8Lt cOHTROL ASSEMBLIES tiROUP HEIGS4f LIMITING CONDITION FOR OPERATION J.l.3.1 A11 full length (shutdOwn and control) rods, shall be OPERABLE and pasit1oned within+ 12 steps (indicated position) of tnefr step countlf" demand i:iosTt1on within one after rod nation. APPl.ICABILITY:

MODES l* and z* ACTION: a. With one or nare full length rads tnaperabl*

di.le to being fn.avable as a result of txc:.ss1ve friction or 119cn1n1ca1 inttrfer"ence or xnown ta be untripp1ble, det1r"m1ne tn1t th* SHUTOUliift P1ARGIN requ1rell9nt of Specification 3.1.l.l is satisfied within l nour and be in HJ7 STANDBY within 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months "s. b. wt th rrare than one ful 1 length rad tnaper"able Of" m1 s-al i gned from l the CTQUP step ccunter demand pas1t1on by itart than + 12 steps (indicated position), be 1n HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Q c. wt th one ful 1 lengtn rad inoperable du* to other than addressea by ACTION a, abOYe, or m1s*a11gned from its CJ"OUP step counter demand position by nare thin + lZ steps (indicated pas 1 t ion), ?O\IER OPERATION m1y cant1 nut. provided that w1 tni n one rour t1 ther: l. 3. Th* rad fs restor"ed to OPERABLE status w1th1n the above 111 gn1111nt requir1*nts, or The r111111nder*o d 1n the t>Ank *1th the 1naP9rable rod are al 1 gned ta within + 12 staps of the 1naperable rod wh11t a1nt1in1ng the rod it *nee and 1nsert1cn lilllits of Figures 3.l*l andl.3.1=2"!-tne RMAL snall be ,..str1ct*d pursuant to Spec1f1 tion 3.l.3.5 during ,JUbsequent o per at 1 on, or The ed. noperable and the SHUTtlOliift MAiitGIN 1"9qu1re,.nt of Spec1f1cat1on J.l.l.l is OPERATION 11111 then cant1nue provided that: *See Special Test Exc:eptions 3.10.2 and 3.10.3. SALEM

UNIT L 3/4 l-18 Amendmnet No. 73

*

R!AC'l'IV!TY CONTROL SYSTPJotS pgsmQN DmICi\TIQN SYSTEM sHU'ff)()W

°' SHtm:Joiori RDb msef2:t1t>1J uMCT" LIMITING CONDmON FOR OPERATION 3.1.3.4 All ahutdovn rod* shall be FULLY WITHDRAWN.

APPLICABII.I'l'Y:

HODIS l*, and 2*11@ ACTION: With a maximum of one shutdown rod not PULLY WITHDRAWN, except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hour either: a. PULLY WITHDRAW the rod, or, b. Declare the rod to be inoperable and apply Specification 3.1.3.1. SURV!ILLANC!

R!QUIR!MENTS 4.1.3.4 !&ch shutdown rod shall be determined to be PULLY WITHDRAWN by use of the group demand counters, and verified by the analog rod position indicators**:

a. Within 15 minute* prior to withdrawal of any in control bank* A, B, C, or D during an approach to reactor criticality, and I b. At least one* per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter

*See Special Test Exceptions 3.10.2 and 3.10.3 **For power levels bel.ow 50% one hour thermal "soak time" is permitted.

During this soak time, the absolute value of rod motion is limited to six steps. I/With Keff greater than or equal to 1. O @Surveillance is applicable prior tQ withdrawing control banks in preparation for startup (Hoda 2). SAUM -UNIT 1 3/4 1-22 Amendment No. 103 ... I I

3.1.3.5 Figures APPLICABILITY:

ACTION: C..o/\l'T"'R.OL XNSERTJDl'l L.tM.n-S limited in physical insertion as shown in With the control banks inserted beyond the above insertion limits, a.xcept for surveillance tasting purSU&nt to Specification 4.1.3.1.2, either: a. Restore the control banks to within the limits within two hours, or b. Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by th* bank position using the above f iguras, or c. Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . SURVEILLANCE REQUIREM!N'l'S 4.1.3.5 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by use of the dema.nd a.::j *;el:'::..:

.ed ':iy :.::a ar.a.:.og l:'Od po=>.i. -:.i.Jn e '.::.;::;c :.::! time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s*** *See Special Test Exceptions 3.10.2 and 3.10.3 **For power levels below 50% one hour thermal "soak time" is permitted.

During this soak time .* the absolute value of rod motion is limited to six steps. //With Keff than or. equal to 1.0 SALEM -UNIT l 3/4 l-23 Amendment No. 103 _ I

INTENTIONALLY L! COMMISSION AiPROVAL OP

ROD ! INS!ITIOH LIMITS V!ISOS TR!IMAL POW!I. I THU! LOOP OP!IATIOH

SALEM

UNIT l 3/4 l*25 No. 91-.-

I '. 1 . .

UMiiS CONOI7!0N (Continued)

' b. limits

c. THERMAL POWER sha i1 not be ir.cre!sed above of RATED POWER unless the indicated .A.FD has not been of the limits for m:ire than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months pena1ty deviation C'.t.'nOJ1ative d_uri ng the previous Z4 hours.

4.2.l.:t The AXIAL FLUX sha11 be de:ennined to be within its during PO',,'£R o;:iE:RATION above of RATED THEW,L PO',(::R by: a. Monitoring the indicated .A.FU for each O?ERABL::

excore channel* 1. At least once per 7 days when the AF'O Monitor A1arm is OPERABLE:, and Z. At 1east once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AF'O Monitor Alar.:: OPERABLE st!tus. b. Monitoring and log;ing the indfe!!ed AXIAL FLUX OIFF'::RENCE for each OPERABLE excore channe1 at least once per hour for tile 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at per 30 minutes theT"!after, when the AXIAL Fl..UX O!rFER:NCE Monitor A1an: is inoperab1e.

The logged values of the indicated AXIAL FLUX sha11 be !Sswned to exis-: during inter,*a1 precedil"lg each 4.2.1.2 The indicated AF'O shall be considered outside of its licits when at ieast Z of 4 or 2 of 3 O?ERA3LE excore channe1s are indicating the AFD to be outside the 11mits of Specification 3.2.1. ?ena1ty deviation outside of the limits sr.ali be aec1G.1Jlated on a time be.sis* of: a. b. One*minute penalty deviation for each one minute of ou:side of the limits at POWER 1eve1s equal to or above of RAiED TH::?-MAL POt,..!R, and pena1ty each one cf ?O'ri::R O?:'.?J..i'!ON outside of the Hrr.its at THERMAL below so: o i. R.!. i'SD PO'rlER. z..o s;..LS1", -UN!i l 3/4 2-2

, , 11 1' . ! '; Ii *I I. 'I I I * . I ! ! 11 i l , , ; ! I; I * . , ... -=

--. --: ---; ... x Days. The provisions of Soecificaticn are nc:

! . 2. l. 4 :he r , . -,1re.

Cays bJ :=-;:: =* _,

':o at::ova er:::/

i-::e'":.:'.=:*.::-:

... :-:'" ;:r:en: at e:-: :y:'.: 2-3

. ) TABLE 4.l-1 REACTOR TRIP SYSTEM CHANNEL MODES IN WHICH CHANNEL FUNCTIONAL SURVBILLANCB PQNCTIQNA.L UHIT CHBCK TEST RBOUIRED 1. Manual Reactor Tr_ip Switch N.A. Ritt 1, 2, and

I 2. Power Range, Neutron Flux s on1, ,.c:i1 0 1, 2 and 0111 3. Power Range, Neutron Flux, N.A. Reill Q 1, 2 High Positive Rate ... Power Range, Neutron Flux, N.A. R_, 0 1, 2 High Negative Rate 5, Intermediate Range, Neutron Flux s Rllf S/1Jll 1, 2 and

6. Source Range, Neutron Flux sm Riii Q and s/li-11 2, ), 4, 5 and

7. Overtemperature 6T s R 0 1, 2 8. Overpower 6T s R 0 1, 2 9. Pressurizer Presaure--Lov s R 0 1, 2 10. Pre*aurizer Preasure--High s R 0 1, 2 11. Preeeurizer Nater Level--High s R 0 1, 2 12. Lo** of Plow -Single Loop s R Q 1 SALIM -UHlT 1 3/4 3-11 Amendment No.176 I TA8LB 4.3-1 (Continued)

REACTOR TRIP SYSTBM INSTRUMENTATION SURVEILLJ\NCE REQUIREMENTS fUNCTIONAL UNIT 13 . Lose of Flow Two Loops 14. Steam Generator Nater Level--Low-Low

15. Deleted 16. Undervoltage

-Reactor Coolant Pumps 17. Underfrequency

-Reactor Coolant Pumps 18. Turbine Trip a. Low Autostop Oil Pressure b. Turbine Stop Valve Closure 19. Safety Injection Input from ESF 20. Reactor Coolant Pump Breaker Position Trip 21. Reactor Trip Breaker 22. Automatic Trip Logic SALEM -. 'T 1 CHANNEL CHBCK s s N.A. N.A. N.A. N.A. N.A. N.A. N.A. N.A. 3/4 12 CHANNEL FUNCTIONAL TEST N.A. t) R Q R 0 R 0 N.A. s/u<1> N.A. s1u<1> N.A. MC4)(5) N.A. R N.A. M(5)(11)(13) and R<14> N.A. M(5) MODES IN WHICH SURVEILLANCE REQUIRED I 1 1, 2 1 1 1, 2 . 1, 2 ' 1, 2 1, 2 and

1, 2 and * /' ) If * )}

L Amendme To. 170 I I i

ACTION 19 -ACTION 20 -ACTION 21 -* ACTION 22 -ACTION 23 -* SALEM -UNIT 1 TABLE 3.3-3 (Continued)

With the number of OPERABLE Channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . *o. The Minimum Channels OPERABLE require::-.ent.s is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing o: other channels per Specification 4.3.2.l.l.

With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.2.1.1 provided the other channel is OPERABLE.

With the number of OPERABLE channels one less than the Minimum Number of Channels, operation may proceed provided that the inoperable channel is restored to within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . NOT USED With the number of OPERABLE channels one less than the Total of Channels, restore the inoperable channel to status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within rs and in at least HOT SHUTDOWN within the following 3/4 3-22 Amendment No. 191 I

* ** ( 1) TABLE 4.3-2 (Continued) 7ABLE NOTATION Outputs are up to, but not including, the output relays. The of Specification 4.0.4 are not applicable.

Each logic channel shall be tested at least once per 62 days on a STAGGERED TEST BASIS. The CHANNEL FUNCTION TEST of each logic channel shall verify that its associated diesel generator automatic load sequence timer is OPERABLE with the interval between each load block within 1 second of its design interval.

(2) Each train or logic channel shall be tested at east every 62 days on a \.___ staggliX"liQ bi.iii.iii 0 'SIAGGerlED TE'ST 'BA'Sl.S.

* (3) s all i ude exercising the transmitter by applying either a vacuum or pressure to the appropriate side of the transmitter.

(4) NOT USED (5) NOT USED (6) Inputs from Undervoltage, Vital Bus, shall be tested monthly. Inputs from Solid State Protection system shall be tested every 62 days on a STAGGERED TEST BASIS . SALEM -UNIT 1 3/4 3-34 Amendment No.187

3/4.l.3 14QNtrORING IHST'RUP!El!TATIOt!

RAG IAT!ON I?!S'T?!!jME[AT!ON 1.n.mHG CON1Hi!ON FOR OPWTIOf! 3.3.l. l nae "41at1on mn'ft:1r1n9 f

ctwtnels shown fn Table 3.l-f shall b* OPGAIL! tt'1111"'

1l1Ml/trtp

1t2'lfn the spec'ff1td 11a1u.

AP't.!CAU!.m:

As shawl f n iabl t 3. l*f. ACT!O!!: 1. il1tb 1 1"'1d1at1on c:nannel a111'1/:Mp seQQf,,i: f"f tt'le value Sta.ft fn 'i'IClt adjust stt;a1nt :a w1-:ain CM 1 fll'ft w1tt'l1n 4 nau1"'S or declare :na dl&.M1l f,.,.,...u11.

b. W1th one o .. mre 1"'&G1a1:1on mft1tor"'frr9 fnopenalt, taJc1 tfta ACrtON SllNI 1 n Tab l t 3. l*f. c. The P'r0"#1s1ons of Sgec11'1ca1:1ans 3.0.3 and 3.0.& are not *01'1 fcabl1. 4.J.l. t Eii:ti l"ld1at1on

=n1t=-1ng c:tt*nnel d.ans"tntad OP!AABL! b)' t:l*

of :ill and OWtNE!. FtmC11CftAL T'EST tnt =an !nd 1t Cle f1"'tqueftc1ts sna..n fn i1al1 4.l-3

5..it.£!1

UMii t 3/4 3.35 -t\*-

INSTRUMENT

2. PROCESS MONITORS b. Noble Gas Monitors 1) Medium Range Auxiliary Building Exhaust System (Plant Vent) 2) High Range Auxiliary Building Exhaust System (Plant Vent) 3) Main Steamline Discharge (Safety Valves and Atmospheric Steam Dumps) 4) Condenser Exhaust System 3 . CONTROL ROOM a. Air Intake -Radiation Level TABLE 3.3-6 (Continued)

RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS OPERABLE 1 1 1/ MS Line 1 2/Intake##

APPL;ICABLE MODES 1, 2, 3&4 1,2,3&4 1,2,3&4 1,2,3&4 ** ALARM/TRIP SET POINT s3. ox10*2µCi/cm 1 (Alarm only) sl. Ox10 2µCi/cm 1 (Alarm only) slO mR/hr (Alarm only) sl. 27x10 4 cpm (Alarm only) s2. 4Bxl0 3 cpm ## Control Room air intakes shared between Unit 1 and 2. ** ALL MODES and during movement of irradiated fuel assemblies and SALEM -UNIT 1 3/4 3-36a MEASUREMENT RANGE 1-10' mR/hr 1-10 6 cpm toeE AL.

D 1'J g ee£e alteratiens ACTION 23 23 23 23 / C l /} 11 CG Amendment No. 190 I

TABLE 3.3-6 (Continued) TABLE NOTATION ACTION 19 -With the of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . ACTION 20 -With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.l. ACTION 22 -With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE' requirement, comply with the ACTION requirements of Specification 3.9.9. ACTION 23 -With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements, initiate the preplanned alternate method of monitoring the appropriate parameter(s), within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and: l) either restore the inoperable Channel(s) to OPERABLE status within 7 days of the event, or 2) prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status. ACTION 24 -With the number of channels OPERABLE one less than required by the Minimum Channels.OPERABLE requirement, restore the inoperable channel(s) to OPERABLE status within 7 days or initiate and maintain operation of the Control Room Emergency Air Conditioning System (CREACS) in the pressurization or recir.culation mode of operation.

CORE ALTERATIONS and movement of irradiated fuel assemblies will be suspended during operation in the recirculation mode. A CTIDN 2.-£ -o channels OPERABLE in a Control Room air intake, immed"ately initiate and maintain operation of the CREACS in the press rization or recirculation mode of operation.

CORE ALTE TIONS and movement of irradiated fuel assemblies will be ended during operation in the recirculation.mode.

SALEM -UNIT l 3/4 3-37 Amendment No.190

TABLE

RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH CHANNELS SOURCE CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECKS CHECKS CALIBRATION TEST REQUIRED 2. PROCESS MONITORS b. Noble Gas Effluent Monitors 1) Medium Range Auxiliary s M R Q 1, 2, 3 & 4 Building Exhaust System (Plant Vent) 2) High Range Auxiliary s M R Q 1, 2, 3 & 4 Building Exhaust System (Plant Vent) 3) Main Steamline s M R Q 1, 2, 3 & 4 Discharge (Safety Valves and Atmospheric Dumps) 4) Condenser Exh. Sys. s M R Q 1, 2, 3 & 4 3. CONTROL ROOM a. Air Intake -Radiation Level s M R Q ** CJ) 72.E AL 'T'£ t2.I\ T 10N.S ** ALL MODES and during movement of irradiated fuel assemblies and duri eore SALEM -UNIT 1 3/4 3-3Ba Amendment No.190

ACCipgrt MONITORING INSIRtJMtNJ'AIION LIMITING CONDITION FOR OPERATION 3.3.3.7 Th* accident monitoring lnscrumentation channels shown ln !able 3.3-ll shall be operable.

MODES l, 2, and 3. ACTION: a. A.£ shown in Table 3.3-ll. b. The provisions of Specification 3.0.4 are not applicable.

SURVEII.Lt.NCE REQUIREMENTS 4.3.3.7 Each accident monitoring in.1 on c ann* all be demon*trated OP perfot'll&nc*

of th* CHANNl:L CHANNEL CALIB TIONVoperation.1 ac th* frequencie*

1hovn in Table 4.3-ll. t\YlJ CHANNEL. l==iJNC...TIDNA.l..-

!EST .. 3/4 3.53 Amenda8nC No. 117 SALEM

OHlT l ""l*

ACTION 1 TABLE 3.3-11 (continued)

TABLE NOTATION With the number of OPERABLE accident monitoring channels less than the Required Number of Channels shown in Table 3.3-11, restore the inoperable channel(s) to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ACTION 2 With the number of OPERABLE accident monitoring channels less than the MINIMUM Number of Channels shown in Table 3.3-11, restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT ii SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . I) w c,hG\11.1" w ACTION 3 deleted I ACTION 4 With the number of OPERABLE channels one le the Required Number of Channels shown in T proceed provided that an OPERABLE Steam Gen Steam Generator el. of OPERABLE Requir roceed prov that Steam nd the following

'red PERABLE to provide an a ystem subcooling margin: a. Reactor Coolant Outlet (Wide Range) channels less than the shown in Table 3.3-11, operation may Tables are available in the Control Room in Table 3.3-11 are ate means of calculating Reactor Coolant b. Reactor Coolant Pressure (Wide Range) '----------

SALEM -UNIT 1 3/4 3-56 Amendment No. 147

. 4. 3-11 SURVEILLANCE REQUIREMENTS FOR .ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHANNEL INSTRUMENT CHECKS CALIBRATION

1. Reactor Coolant Outlet Temperature

-THOT (Wide Range) 2. Reactor Coolant Inlet Temperature

-TcoLD (Wide Range) 3. Reactor Coolant Pressure (Wide Range) 4. Pressurizer Water Level 5. Stearn Line Pressure 6. Stearn Generator Water Level (Narrow Range) 7. Stearn Generator Water Level (Wide Range) 8. Refueling Water Storage Tank.Water

9. deleted 10. Auxiliary Feedwater Flow Rate 11. Reactor Coolant System Subcooling Margin Monitor Level M M M M M M M M #Auxiliary Feedwater System is used on each startup R ('.)f\ll*t R R R R R R R 11 11nh

CHANNEL FUNCTIONAL

..llkNA* , . l' -H<< N. A. .llk N. A.

The instruments used to develop RCS subcooling margin are calibrated on an 18 month cycle; the monitor will be compared quarterly with calculated subcooling margin for known input values. SALEM -UNIT 1 3/4 3-57 Amendment No. 147 TABLE 4.3-11 (Continued)

SURVEILLANCE REQUIREMENTS FOR ACCIDENT MONITORING INSTRUMENTATION INSTRUMENT

12. PORV Position Indicator

13. PORV Block Valve Position Indicator

14. Pressurizer Safety Valve Position Indicator

15. Containment Pressure -Narrow Range 16. Containment Pressure -Wide Range 17. Containment Water Level -Wide Range 18. Core Exit Thermocouples

19. Reactor Vessel Level Instrumentation System (RVLIS) CHANNEL CHECKS M M M M M M M M

Unless the block valve is closed in order to meet the requiremenrs of Action b, or c in specification 3.4.3. SALEM -UNIT 1 3/4 3-57a R CHANNEL FUNCTIONAL TEST R R -Hfr N.A. tl.A. .uA-N. A. ...blA N. A. -Hf\ N. A. Amendment No.

T"e HNCTlONAL

".'ES7 s:ia11 also

':"at a:.Jtorna-:*c of ':his pathway and control rool'l alarm annunciation occurs 1f any 0f -:rie following conditions exist: l. !nstrulT'l!nt indicates l'l!asured levels at or aoove the alarm/trio set point. 2. Circuit failure. (Loss of Power) 3. !nstrul"ent indicates a downscale failure. (IndicHion on instl"'urrent drawer in Control Equipl'l!nt only) 12) "'.'he CHANNEL FUNCTIONAL TEST stial 1 al so r!erTCnstrat.e that :antral room al ar"ll annunciation occurs if any of following conditions Pxist: ( 3) ,. ,. l. !nstrul"ent indicates l"@asured levels at or above the alarm/trip set point. 2. Circuit failure. (Loss of Power) 3. Instr\11'1!nt inrticat.es a cX>wnscale failure. (Indication on instrurrent drawer in Control Equiprrent Room only) 4, rnstrul"ent controls not set in operate rmde. (On instrufT'f!nts equippP.1 with operate rTCde switches only) 7'ne initial CHANNEL CALIBRATION was using appropriate liquid or gaseous cal i oration sources obtained from reputable suppliers.

The activity of the calibration sources were reconfirrT"ed using a rrulti-criannP.1 analyzer which was calibrated using one o.-.-:ure standards.

CHANNEL CHEC)( shall consist of verifying indication of flow during per;ods of release. CHANNEL CHEC)( shall be Made at least once per 24 nours on cays on wnich continuous, periodic, or batch releases are made. During liquid additions to the tank. If tank level indication is not providP.t1, vertification will be !1one hy visual inspP.ctit')n, cha nel is an in-line channel wnich requires periodic decontamination.

Any count rate indication aMve 10,000 cpl"'I cnnstit 1 i<:.es a for col'l'T'Cliance purposes

3/4 3-63 Amendment SJ

TABLE 4 . 3 -13 RADIOACTIVE GASEOUS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL

MODES IN WHICH RCE CHANNEL FUNCTIONAL INSTRUMENT CK CALIBRATION TE.ST REQUIRED 1. WASTE GAS HOLDUP SYSTEM a. Noble Gas A-::tivity Monitor -Providing P p R ( 3) Q (1) Alarm and Automatic Termination of Release b. Oxygen Monitor D N.A. Q(4) M R ( 3) Q ( 1) 2. CON'.'AINMENT PURGE AND PRESSURE -VACUUM RELIEF a. Noble Gas Activity Monitor -Providing P Alarm and Automatic Termination of Release 3 PLANT VENT HEADER SYSTEM# a. Noble Gas Activity D M I<( 3) Q (2) b. Iodine Sampler w N.A. N.A. N.A. c. Particulate Sampler w N.A. N.A. N.A. D N.A. R N.A.

d. Flow Rate Monitor e. Sampler Flow Rate Monitor w N.A. R N.A. # The following process streams are routed to the plant vent where they are effectively monitcned by the instruments described: (a) Condenser Air Removal Systeill (bl Auxiliary Building Ventilation System (c) Fuel Handling Building Ventilation System (d) Radwaste Area Ventilation System (e) Coutairunen::

Purges SALEM

UNIT l 3/4 3-68 No. 157

*** SllaV!IUh'C!

u:QUIUKENTS (Continu.ed) 4.3.4.3 The abov. required turbine over*p**d protection

Y*tea *hall be damonatrated OPEIA!LE:

- b. Ac l***t once per 18 month* by perfor11&nce of a CHANNEL CALIBllATIOH on th* turbine overspeed protection sy*c ....

AC 1 p* 40 monct\9 by di* .. *eabling ac le .. c one of each of t *

ab*ove va vu and performing a visu.al and surface 1mpection of v v , disk.8 and see .. and verifying no unacceptable flaws or corrosion.

If unacceptable flav* or exc***ive corro*ion are found, all other valves of that typ* *hall be 1n8pected.

4.3.4.4 Verify th* t**C frequency 11&incaina the probability of a mis*ile ejection incident vithi.n Nl.C guidelirw*

by revievina th* *thodology pnHnted in WCAP-11525:

a. At le .. t once every two refu.eliq ouca1**. b. After aodificationa to th* 11&in turbine or turbia. ov*r*p**d protection valve* . SALEK -UNIT l 3/4 3-71 Aaandment No. 115

COO SYSTEM COLD SH UT OOWN LIMITING CONDITION FOR OPERATION 3.4.1.4 Two# residual heat removal loops shall he OPERABLE*

anrl at least one RHR loop shall be in operation.**

APPLICABILITY:

MO IE 5. H ACT ION: a. With less than the above required loops operable, inrnediately initiate corrective action to return the required loops to OPERABLE status as soon as possible.

b. With no RHR loop in operation, suspend all operations involving a rerluction in boron concentration of the Reactor Coolant Syste1ft and i111T1ediately initiatP.

corrective action to return the rP.quired RHR loop to operation.

SURVE I Rt.OUIREMENTS 4.4.1.4 At least one residual heat removal loop shall he verified to he in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . I One RHR loop may be inoperable for up to two hours for surveillance testing, provided the other RHR loop is OPERABLE and in operation.

Additionally, four filled reactor coolant loops, at least two steam generators with their secondary side water levels greater than or equal to 51 (narrow range), may be substituted for one residual heat re"10val lnop. operahility may be excepted as follows: a. The nonnal or emergency power source may he inoperable.

b. One service water hearler may be out of service provided the equipl'M!nt listed in Table 3.4-3 is OPERABLE.

- The residual heat removal pumps may bP. de-energized for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided 1) no operations are permitted that would cause dilution of thP. reactor coolant system boron concentration, and 2) core outlet te!Tlperature is maintained at least 10°F below saturation te1ftPerature.

SALEM -UNIT 1 3/4 4-3b AMENDMENT NO. 7 2

c. :he selected as the second and third samcles (if reouired 1able 4.4-2) each ;nservice inspection may be subjected part1al tube inspect1on provided:

l. The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with found. 2. The inspections include those portions of the tubes where imperfections were previously found. The results of each sample inspection shall be classified into one of the fol lowing three categories:

Category Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 1% of the total tubes inspected are d£ ..*. ive, or between 5\ and 10% of the total tubes inspected arn 1egraded C-3 More than 10\ of the total tubes inspected are degraded tubes or more than 1\ of the inspected tubes are defective.

Note: In all inspections, previously degraded tubes must exhibit significant

(>10\) further wall penetrations to be included in the above percentage calculations.

4.4.5.3 Inspection Frequencies

-The above reQuired inservice inspections of steam generatoi tubes shall be performed at the following freQuencies:

a. b. The first inservice inspection shall be performed 6 Effect;*:e Full Power Months but within 24 calendar months of initial criticality.

SubseQuent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection.

If two consecutive inspections following service under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional has occurred, the inspection interval may be extended to a maximum of once per 40 months. rf the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40 month intervals fall in Category C-3, the inspection freQuency shall be increased to at least once per 20 months. The increase in inspection freouency shall apply until the subseQuent inspections satisfy the criteria of Specification 4.4.5.3.a; the interval may then be extended to a maximum of once per 40 months. SALEM -UN IT 1 3/4 4-8 Amendment No. 118

REACTOR COOLANT SYSTEM 3 / 4. 4. 6 REACTOR COOLANT SYSTDt LEAKAGE LEAKAGE DffiCTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System leakage detection systems shall be OPERABLE:

a. The containment atmosphere particulate radioactivity monitoring system. b. The containment sump level monitoring system, and c. Either the containment fan cooler condensate flow rate or the containment atmosphere gaseous radioactivity monitoring system. APPLICABILITY:

MODES 1, 2, 3 and 4 . ACTION: With only t'1IO of the above 11akage detection systems OPERABLE, operation may continue for up to 30 days provided grab of the containment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the required gaseous and/or particulate radioactivity monitoring system is 1noperablei otherwise.

be in at laast HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.4.6.1 The 111kage detection systmns shall be desnonstrated OPERABLE by a. Containment atmosphere particulate and gaseous ( ;f being monitoring systems-perfonnance of CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the fn<1uenc1 es S'.)(.)t<ct U\ErJ() specified in Table 4.3-3. b. Containment st111p level and containment fan cooler condensate flow rate {if being used) monitoring systems-perfonnance of CHANNEL CALIBRATION at least once per 18 months

SALEM -UN IT 1 3/4 4-14 -*--_ ... --****-------* . * ..

SYSTP1 COOLANT

!SOLAilON VALV:S CONDITION FOR i! 3.4.6.3

?ressure Isoiaticn Valves sha11 be operational.

a. The inte in Tabl pre>>sure isolation valves iisted have been except Vaive leakage shall not exceed d. b. In the event that*the

If nei tr.er CJndit.ion "a" nor "b" can be met, orderly shutdown shall be initiated within one hour and the reactor sha11 be in at least HOT STAtH'3Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . 4.4.5.3 each va1ve 1isted in plant is placed COLD SHUTDOWN condi:icn for refue1 i ng. 2. Each tir:ie the plant is placed in COLD SHUTDO\.'N conci:icn for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months if testing has not accomp1ishec in the preced5ng 12 months. (a) :*'io:Jr operated valves shall be placed in the closed position and power supplies deenergized. . . ( io sa:isfy 1eakage may be :i':easured indirect1y (as of pressure 1ncicators) if in

-' d

d "' * *

h . ** ..i.. ;.;i':h a;:i;:::-cvec an wY co::ipu .. a .. ions S owin; .. na: *.. e ::-.e::iod is ca;n':lie of ce-::;cnstra";in; valve c::7o;:i1iance wi':h the leakage cri:eria.

3/t.

Qrder dated I I I I I

3. ?ricr valve to fo1lo* * .-1ng re?air, or rc?1ace!:'.ent on

4.

of scecific!ticn 4.0.C is not for entry into ,l.\ode 3 or .4. b. isolation valve integrity of ... ,, i each high pressure line having a and recorded daily. In additicn, the po

a of one oth<:r valve located in the high press*.;rE>

line shall recorded dai1y. Order dated i ?I"'. -' .. .--! -: : '

- (a)i. Leaka;e less than or to 1.0 s;::i:".'!

are cor.sicered ac:c:epta:le.

for initi!1 tests, or tests following valve or rr.ent, leakage rates less than or equal to 5.0 gpm are considered ac:c:eptable.

Z. Leakage rates sreater than i .O g;rn but less than or to 5,0 g;:r.: are ac::eptable if the measured rate has not exc:eeded the rate the test by an amount that reduc:es the between leakage rate and maximum rate of 5,0 by sc: or gre!ter, 3, Leaka;e rates greater than 1 .0 but 1ess than or to 5,0 are consicered ur.a:c:ept!ble if the latest measured rate exceeded the rate by t!':e previous test an that reduces the rr.arcin bet..,*een*

1eaka:e

!:id the

iem'fssible of 5.o or greater

(o) u* *

. ,

,,,eren .. 1a .

l test sure sha11 not be 1 ess than 1 SO psid. 3/4 4-16C Order dated Aoril

WCj'l'QI COOLMT SXSTQ SPJCtrtc ACjTIVlTX LIMITillG CCllDITIOR FOi OPll'ATIOR

- - - - ----***********************g***********a***c*a*******************

J.4.8 Th* *pecific activity of the priaary coolLnt *hall be limited toz ** l.0 >1-Ci/9ru DOSI IQUIVALIRT I-lll, and APPLICA1ILITJ'1 MODll 1, 2, 3, 4 and 5 ACTIQ11 KODIS 1, 2 and 3* a. With the *pecific activity of the primary coolant > 1.0 >'C1/9ram DOSI IQOIVALIM'l' I-131 for more than 41 hour* durift9 one continuoue time interval or *xc*edin9 th* limit line *bown on Piqure J.4-1, be in at lea*t HOT STA!ft)IY with T < soo*r within 6 hour*. *-.v b. With th* epecif ic activity of the prill&ry coolant > 100/1 >1-Ci/9ram, be in at lea*t HOT STABDIY with T < SOO*r within 6 hour*. *YCJ MODll 1, 2, 3, 4 and 5 a. With the *pec:ific activity of tJle,_,,...,.._..'Y DOSI IQOIVALmn' I-131 or > 100/ analy*i* requirement*

of it .. activity of th* primary coolan SOltVSILLMC:S UQOIJlDD'l'I

........................................

4.4.1 Th* epec:ific acti*ity of th* primary coo t 1 be dete.rmin*d

o b* within the liait* by performance of the H11Plin9 and an&ly*i* pr09ram ot Table 4.4-4. *with T 2 soo*r. ave; SAI.IH -OllI'l' 1 3/4 4-20 Amendment

-<o. '. 33 1

RF.ACTOR COOWfI SYSTEM OVERPRESSURE rROTECIION SYSTEHS SURVEIUANCE REQUIREMENTS

a. rformance of a CHANNEL FUNCTIONAL TEST on the POPS actuation channel, but valve within 31 days prior to a condition in the POPS is lea.st OYl6t. f'W 31 fl,,e.ye..o:

fhL 'PoPS ,*s N?..iu*rd OPeRABtE.

Performance of a CHANNEL CALIBRATION on the POPS accuation channel at lea o ce per 18 montha. c. Verifying the POPS isola e is open hours when the POPS is being used for overpressure

d. Testing pursuant to Specification 4.0.5. 4.4 .. 9.3.2 The RCS vent(s) shall be verified to be open at least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s* when vents(s) is being used for overpressure protection.

Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days. SALEM

UNIT 1 J/4 4-31 Amendment No. 150

REACTOR COOLANT SYSTEM 3.4.10 S1RUCTURAL INTEGRITY ASME CODE CLASS l , 2 and 3 COMPONENTS LIMITING CONDITION FOR OPERATION 3.4.10.1 The structural integrity of ASME Code Class 1, 2 and 3 ponents sha11 be maintained in accordance with Specification 4.4. 1 . 1 .1. APPLICABILITY:

ALL MODES ACTION: a. With the structura 1 integrity of ar . .1 .-. .3ME Code Cl ass 1 ponent ( s) not conforming to the above restore the structural of the affected component(s) to* wittiin its 1 imit or* isci1ate tt.e ilffected component(s) prior to inrreasing the Reactor Coolant Systen1 ten1perature.more thar1 50°F above t!le minimum temperature required by NOT considerations.

b. With the structural integrity of any ASME Code Class 2 ponent(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate tlte affected component(s) prior to increasing the Reactor Coolant System temperature above 200°F. c. With the structura 1 integrity of any ASME Code Cl ass 3 component(s) not confonning to the above requi'"!ments, restorP the structura 1 i ty of the affected compone:.-1t ( s) to within its limit or isolate the affected component(s) from service. d. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.10.1 .1 The structural integrity of ASHE Code Class 1, 2 and 3 components shall be demonstrated:

a. Per the :equirements of Specification 4.0.5, and *-b. Per the requirements of the augmented inservice inspection program in Specification 4.4.10.l.2.

1 SALEM -UNIT 1 3/4 4-32 Amendment No. 24

--EMEBGENCY CORE COOLING SYSTF.MS ECCS SUBSYSTEMS

-Tave <3so*r LIMITING CONDITION FOR OPERATION

3. 5. 3 a one ECCS subsystem comprised of the following shall be OPERABLE:

a. One OPERABLE centrifugal charging pwap and associated flow path capable of taking suction from th* refueling water storage tank and transferring suction to the residual heat removal pump diacharge piping and; 1. Diacharging into each Reactor Coolant Systea (RCS) cold leg. b. One OPERABLE residU&l heat removal pump and aaaociatad reaidual heat removal heat exchanger and flow path capable of taking suction from the refueling water storage tank on a safety injection signal and transferring suction to the containment sump during the recirculation phase of operation.

and; 1. Discharging into each RCS cold leg, and; upon manU&l initiation, 2. Discharging into two RCS hot legs. APPLIGABILIIX:

MODE 4. ACTION: a. With no ECCS aubayatea OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within l hour or be in COLD SHUTDOWN within the next b. OPERABLE because of the inoperability of removal heat exchanger or residual heat mov , reatore at least on* ECCS subayatea to OPERABLE status or maintain the Reactor Coolant Sy*tea T lesa than 3so*r by use of alternate heat removal 11ethoda.

avg c. In the event the ECCS 1a actuated and inject* water into the Reactor Coolant Syatea, a Special Report shall be prepared and submitted to the co ... i*aion puraU&nt to Specification 6.9.2 within 90 days deacribing the circumstance*

of the actuation and the total accumulated actuation cycle* to date. # A maximum of one safety injection pump or one centrifugal charging pump shall be OPERABLE in MODE 4 when the temperature of one or more of the RCS cold leg* is leas than or equal to 312.F, Mode 5, or 6 when the head is on. the reactor vessel. SALEM -UNIT 1 3/4 5-6 Amendment No. 150 *-.

EMERGENCY COAE COO LING SYSTEJ14S LE LING WATER STORAGE TANK LIM ITrNG CON OIT ION FOA OPERAT !ON 3.5.5 Tt'te refueling w1t1r storage tank (RWST) shalt a. A contained volume of b*twe*n 364,500 and 4 b. A boron concentration of betw11n 2,300 and c. A water of 35-F. APPLICABILITY:

MOCES 1, 2, 3 and 4. ACTION: th: With the r*fueling water storag* tank fnotJ*rabl*, restore the tank to OPERABLE status within 1 titour or be fn at lent HOT STANl!Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and fn COLD SHUTCDWN within the following 30 t'tours. S LRVf I LL,,-NCE REQUIREMENTS 4.5.5 The RWST shall be daonstrattd OPERABLE:

a. At least once per 7 days by: 1. Vtr1fying the water level in the tank, Ind 2. Verifying the boran concentration of the water. b. At least Oftca per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by ver1 fying th* RWST ttfl'CJ*rature when the outside 11r t1111P*r1tur1 fs < 35-,:. SAL.EM

UNIT 1 3/4 5-7 Amendment No. 83

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION

==================================================================

3.6.1.l Primary CONTAINMENT INTEGRITY shall be APPLICABILITY:

MODES 1, 2, 3 and 4. ACTION: Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS

==================================================================

4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a. At least once per 31 days by verifying that: 1. All penetrations*

not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except for valves that are opened under )1-" administrative control as permitted by Specification 3.6.3.1_/, and 2. All equipment hatches are closed and sealed. b. By verifying that each containment air lock is OPERABLE per Specification 3.6.1.3. *Except vents, drains, test connections, etc. which are (1) one inch nominal pipe diameter or less, (2) located inside the containment, and (3) locked, sealed, or otherwise secured in the closed position.

These penetrations shall be verified closed at least once per 92 days. SALEM -UNIT 1 3/4 6-1 Amendment No. 189 CONTAINMENT

CONTAINMENT

...E.AKAGE LIMITING C8NDITION FOR OPERATION

==================================================================

3.6. 1.2 Ccintair.:nent leakage rates shall be limited to: a. An :::i*.;eraL integrated leakage rate of s; La, 0. :o percent by weight 0: the air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at design pressure, (47.0 psig) . b. A leakage rate of s; 0.60 La for all penetrations and valves subject to Type B and C tests, when to Pa. APPLICABILITY:

MODES 1, 2, 3 and 4. ACTION: With either (a} the measured overall integrated containment leakage rate exceeding 0.7S La, or (b} with the measured combined leakage rate for all penetrations and valves subject to Types Band C tests exceeding 0.60 La, restore the leakage rate{s) to within the limit(s} prior to increasing the Reactor Coolant System temperature above 200°F. SURVEILLANCE REQUIREMENTS

===============================================

-====== 4.6.1.2 ows: a. Type A tests shall be in accordance with lOCFR SO.S4 (0) in conformance with Appendix J of lOCFR SO, Option B, using the methods and provisions of Regulatory Guide 1.163, September 199S as modified by approved exemptions.

b. Type B and C tests shall be conducted in conformance with Appendix J of lOCFR SO, Option A, with gas at design pressure (47.0 psig) at intervals no greater than 24 months except for tests locks. c. Air locks shall be tested and demonstrated with Appendix J of lOCFR SO, Option A, per survei lance Requirement 4.6.1.3. SALEM -UNIT 1 3/ 4 6-2 Amendment No. 189

CONTAINMENT SYSTEMS CONTAINMENT STRUCTURAL INTEGRITY LIMITING CONDITIONS FOR OPERATION

===========================================================

3.6.1.6 The structural integrity of the containment shall b maintained at a level consistent with the acceptance criteria in Specificati n 4.6.1.6.1.

APPLICABILITY:

MODES 1, 2, 3 and 4. ACTION: With the structural integrity of the containment not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200°F. SURVEILLANCE REQUIREMENTS

===========================================================

4.6.1.6.1 Containment Surfaces The structural integrity of the exposed accessible interior and exterior surfaces of the containment, including the liner plate, shall be determined by a visual inspection of these surfaces.

This inspection shall be performed prior to the Type A containment leakage rate test (reference Specification 4.6.1.2) to verify no apparent changes in appearance or other abnormal degradation.

If the Type A test is performed at 10 year intervals, two additional inspections shall be performed at approximately equal intervals during shutdowns between Type A Tests. 4.6.1.6.2 Reports Any abnormal degradation of the containment structure detected during the above required inspections shall be evaluated for reportability pursuant to 10CFR50.72 and lOCFRS0.73.

The evaluation shall be documented and shall include a description of the condition of the concrete, the inspection procedure, the tolerances on cracking, and the corrective action taken. SALEM -UNIT 1 3/4 6-8 Amendment No. 184 SPRAY A.DDIIIVE SYSIE,!1 LIMITING CONDITION FOR OPERATION 3.6.2.2 The spray additive system shall be OPERABLE with: a. A spray additive tank containing a volume of between 2568 and 4000 gallons of beeween 30 and 32 percent by weight NaOH solution, and b. Tvo spray additive eductors each capable of adding NaOH solution from the chemical additive tank to a containment spray system pump flow. APPLICABILITY:

MODES 1, 2, 3 and 4. ACTION:

the spray additive system inoperable, restore the system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT ST**m'y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the spray additive system to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEil.LANCE REQUIREMENTS 4.6.2.2 The spray additive emonstrated OPERABLE:

erifying th&t each valve a. At least once per 31 (manual, power operate locked, sealed, ic) in the flow path that is not:

in position, is in its correct posit:ion.

b. At least per 6 months by: l. Verifying the solution level in the tank, and 2. Verifying the concentration of the NaOH solution by chemical analysis.

c. At least once per 18 months during shutdown, by verifying that .ac.n e11rcw.etic v&lv* in flow path acttwr .. t.o it.a correct position on a Containment High-High pressure test signal. d. At least once per 5 years by: SALEM

UNIT l l. Verifying a NaOH solution flow rate of 12 +/- 3 gpm from the spray additive tank through sample valve lCS61 with the spray additive tank at 2.5 +/- 0.5 psig and 3/4 6-10 Amendment No. 122. '

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

==================================================================

4.6.3.1.2 Each containment isolation valve shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by: a. Verifying that on a Phase A containment isolation test signal, each Phase A isolation valve actuates to its isolation position.

b. Verifying that on a Phase B containment isolation test signal, each Phase B isolation valve actuates to its isolation position.

c. Not used. d. Verifying that on a Containment Purge and Pressure-Vacuum Relief isolation test signal, each Purge and Pressure-Vacuum Relief valve acruates to its isolation position.

e. Verifying that the Containment Pressure-Vacuum Relief Isolation valves are limited to 5: 60° opening angle. 4.6.3.1.3 At least once per 18 months, verify that on a main steam isolation test signal, each main steam isolation valve actuates to its isolation position.

4.6.3.1.4 The isolation time of each power operated or automatic containment isolation valve shall be determined to be within its limit when tested pursuant to Specification 4.0.5. 4.6.3.1.5 Each containment purge isolat' demonstrated OPERABLE within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after eac closing o he valve, e cept when the valve is being used for multiple 'ngs, then at least one per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , by verifying that when the measured lea age rate is added to the e kage rates determined pursuant to Specifica all other Type B and C penetrations, the combined leaka e rate isjless than or equal to .60La. Gz.. b] 4.6.3.1.6 A pressure drop test to identifytess1ve degradatio of resilient valve seals shall be conducted on e: a. Containment Purge Supply and once per 6 months. b. Deleted. SALEM -UNIT 1 3/ 4 6-13 s at least Amendment No. 189 .,..

SYS:::?-!S SURV!II.Ll.NC!

R.!QU!R.EM!N'!'S (Continued)

3. Snubber ralease rat*. where required, is within th* specifiad rang* in compression or tension. For snubbers specifically required not to displace under continuous the ability of the snubber to withstand load without displacement shall be verified.

f. Snubber Service Lif* Monitorin1 A record of the s*rvice life of each snubber, the at which d**i1nat*d s*rvic* life commances and the instal maintanance record* on which the de1i1n&t*d 11f*v>s shall be required by Specificatipn 6.l0.2lr\

sed ConcurTent with th* first inservic*

vi.su&l one* per l8 months thereafter, the installation and m.a\.a.t'anance recor" for each snubber shall be reviewed to verify that th* indicated service life has not been exceeded or vill not be exceeded prior to th* nest scheduled snubber service 11fe review. If the indicated service life will be uceedad prior to the nezt scheduled snubber service life review, th* snubber service life shall be reevaluated or th* snubber shall be replaced or reconditioned so as to extend i u Hrvic* life beyond th8 d&te of th* nut scheduled service life review. This reevaluation, replac ... or reconditionin1 shall be indie&ted in th8 recorct.

Jli! ... SAL.D1

UNIT l / 3/4 7*31 Amendment No. 93

PLANT SYSTEMS 3/4.7.10 CHILLED WATER SYSTEM -AUXILIARY BUILDING SUBSYSTEM LIMITING CONDITION FOR OPERATION sa:aa****&a***--***********************Q=**m*m===*=====*****z=====

3.7.10 The chilled water system loop which services the safety-related loads in the Auxiliary Building shall be OPERABLE with: a. Three OPERABLE chillers b. Two OPERABLE chilled water pumps APPLICABILITY:

ALL MODES and during movement of irradiated fuel assemblies.

ACTION: MODES 1, 2, 3, and 4 a. With one chiller inoperable:

1. Remove the appropriate non-essential heat loads from the chilled water system within 4 2. Restore the chille us within 14 days or; 3. Be in at least HOT 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following

b. With two chillers inoperable:

c. l. Remove the appropriate non-essential heat loads from the chilled water system within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and; 2. Align the control room emergency air copditioning system (CREACs) for single filtration operation using the Salem Unit 2 train within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and; 3. Restore at least one chiller tus within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or; 4. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . hours. ter pump inoperable, restore the chilled water atus within 7 days or be in at least HOT STANDBY ours and in COLD SHUTDOWN within the following 30 SALEM -UNIT l 3/4 7-33 Amendment No. 199

PLANT SYSTEMS LIMITING CONDITION FOR OPERATION ACTION: MODES 5 and 6 or during movement of irradiated fuel assemblies.

a. With one chiller inoperable:

1. Remove the appropriate non-essential heat loads from the chilled water system 2. Restore the chiller 14 days or; 3. Suspend CORE ALTERATIONS and movement of irradiated fuel assemblies.

b. With two chillers inoperable:

1. Remove the appropriate non-essential heat loads from the chilled water system within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and; 2. Align the control room emergency air conditioning system (CREACs) for single filtration operation using the Salem Unit 2 train 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and; 3. Restore at least one chiller t atus within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months c. or; 4. Suspend CORE ALTERATIONS and movement of irradiated fuel assemblies.

With pump SURVEILLANCE REQUIREMENTS

============-===============z==========z================

and 4.7.10 The chilled water loop which services the safety-related loads in the Auxiliary Building shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each manual valve in the chilled water system flow path servicing safety related components that is not locked, sealed, or otherwise secured in position, is in its correct position.

b. At least once per 18 months, by verifying that each automatic valve actuates to its correct position on a Safeguards Initiation signal. c. At least once per 92 days by verifying that each chiller starts and runs.

During Modes 5 and 6 and during movement of irradiated fuel assemblies, chilled water components are not considered to be inoperable solely on the basis that the backup emergency power source, diesel generator, is inoperable.

SALEM -UNIT 1 3/.4 7-34 Amendment No. 1 9 9 .......

.: 5 -*:::::...:

J C. :!STR!9L'"T!ON

-

CONDITION FOR OPERATION 3 a : 3 :allowing D.C. bus trains shall be OPERABLE and energized:

.A consisting of 125-volt D.C. bus No. :A. 125-volt D.C. battery :A and battery charger lAl.

B consisting of 125-volt D.C. bus No. lB. 125-volt D.C. battery No. :a and battery charger :Bl. consist::..ng No. lC and batte bus No. lC, 125-volt D.C. battery 1. A!?L:cAB::::-Y:

l, 2, 3 and 4. AC':':ON:

a. With one 125-volt D.C. bus inoperable or not energized, restore the inoperable bus to OPERABLE and energized status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be ::..n at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHtrrDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . b. With one 125-volt D.C. battery charger inoperab:e, restore the inoperable charger to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or connect the backup charger for no more than 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . c. With one or more 125-volt D.C. batteries with one or more battery cell parameters not within the Category A or B limits of Table 4.8.2.3-1:

' .. ... 3. 4. d. ".-Jith cell w:..th Verify within l hour, that the electrolyte level and float voltage for the pilot cell meets Table 4.8.2.3-1 Category C :..::..mits, and '/erify within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , that the battery cell parameters of all connected cells meet Table 4.8.2.3-1 Category C limits, and Restore battery cell parameters to Category A and B limits of Table 4.8.2.3-1 within 31 days, and If any of the above listed requirements cannot be met, comply with the requirements of action f. one or more 125-volt D.C. batteries with one er more battery parameters not within Table 4.8.2.3-1 Category C values, comply the requirements of action f. e. With average electrolyte temperature of representative cells less than 65°F, comply with the requirements of action f. f. Restore the battery to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SALEM -UNIT 1 3/4 8-8 Amendment No. 177

REFUELING OPERATIONS CRANE TRAVEL -FUEL HANDLING AREA LIMITING CONDITION FOR OPERATION 3.9.7 Loads in excess of 2200 pounds shall be prohibited from travel over fuel assenm11es

.. 1n c.ne storage pool

APPlICABILITY:

With fuel assemblies in the storage ACTION: wtth the requirements of the above specification mt satisfied, place the crane load 1n a safe condition.

The provisions of Spec1f'icat1on 3.0.3 are not applicable.

SURVEILLANCE RE()JIREMENTS SALEM

UNIT l ...... Ls crane travel with loads in excess of del!1lnstrated OPERABLE within 7 days 7 days thereafter during the crane 3/4 9-7 77 .......

J/4. 'B f$MJ>YAL AN!::> lDoLArIT Ql2WU.TJD/'{.

GQOW!I crpc 7'LAUC?f' ALL: =\PPLIC.\BILITI:*

MODE 6. ACTION: a. Wlth less than one resldual heat removal loop ln operation, a.s provided ln b. below' suspend an operations lnvolving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant Sy.stem. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . b. Th* residual heat removal loop may be remov*d from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs. c. The provisions of Specification 3.0.3 are not applicable.

REQUIR.EMETS 4.9.8. l At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months one R.HR loop shall be verified ln operation and circulating coolant at a flow rate of: a. greater than or equal to 1000 gpm, and maintain the RCS temperature at less than or equal SAUM . UNIT l 3/4 9-8 Amendment 109

REFUELlNG OPERATlONS

LOW \#. TER LEV EL LIM1TIH6 CONOlTIQH.FOR OPERATION 3.9.8.2 Two independent Residual Heat Reno val (RHR) loops shal 1 be OPERABLE.*

APPllCASlLITY:

MOOE 6 wnen water level above the top of' the reactor pressure*

vessel flange is less.than 23 feet. ACTION: a. wt th less than the requ1 red RHR loops o 1n1t1at.-co-rrect1ve action ta return OPERABLE status as soon as poss1bl* b. The provisions of Spec1f1cat1on

3. SURVEILLANCE to 4.9.8.2 The required Resi"dual Heat Renoval loops sna\l be deterinint:\A Ql)ERABLE per Specification 4.u.s *

Systems supporting RHR loop operability may be excepted as follows: a. The normal or em1rg1ncy po.-r source may be inoperable.

b. One service water header may be out of service provided the equip1111nt listed in Table .3.4*3 is OPERABLE.

SALEM -UM lT l 3/4 9-8a Amendnient No. 72 : ...

I OPERATION:>

LEVEL

VESSEL be maintained over the top APPLICABIL 9 rrovement of fuel assemblies or control M)ds within the reactor pressure vessel while in MODE 6. ACTION: With the requirements of the above specification not satisfied, suspend al1 operations involving movement of fuel assemblies or control rods within the pressure vessel._ The provisions of Specification 3 .0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.10 The water level shall be determined to be least its minimum required depth within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter during movements of fuel assemblies or control rods. 1 SALEM

UNIT 1 3/4 9-10 AMENDMENT NO. 28 .-:

3 / lJ SPECIAL TES7 EXCEPTIONS SHOTpQWN M;RGIN LIMITING CORDITION FOR OPERATION 3.10.1 The SHOTDOWN of Specification 3.l suspended for measurement of control rod worth and shutdo the reactivity equivalent to at least the highest estima d control ro worth is available for trip insertion frc:xn OPBRABLB control rtsl* C Ai'PLic:ABILITX:

MODE 2. ACTION: a. With any full length control rod not fully in*erted and with less than the above reactivity equivalent available for trip inaertion, immediately initiate and continue boration at 33 gpm of a solution containing I 6, 560 ppm boron or its equivalent until the SHOTDOW!I MARGIN required by I Specification 3.l.l.l is restored.

b. With all full length control rods inaerted and the reactor su.bcritical by less than the above reactivity equivalent, immediately initiate and continue boration at 33 gpm of a solution containing

ii!: 6, 560 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.l.l.l is restored.

SURVE ILLANCB RB QUI REMBNTS 4.10.l.l The position of each full length and part length rod either partially or FULLY WITHDRAWN shall be determined at lea*t once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . 4.10.l.2 Bach full length rod not fully in*erted shall be demonstrated capable of full insertion when tripped from at lea*t the sot withdrawn position within 24 hour* prior to reducing the SHOTDOW!I to less than the limits of Specification 3.l.l.l . SALBM -IJNIT l 3/4 10-1 Amendment No. 145

3 *

-,, LLD cefioed 4,11,I Sl-1,lf -.1 The principal ganna emitters for which the LLQ specification aoolies o:?xclusively are the following radionuclides:

Kr-d7, Kr-a8, t.e-133, Xe-l33m, Xe-135, Xe-138 for gaseous emissions Fe-59, Co-Sa, Co-60, Z.n-65, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions.

This 1 ist does rlJt rrean that only these nucl ides are to oe detected and reported.

Other peaks that are measurable and identifiaole, together with the above nuclictes, shall also be identified and reported.

c. Sampling and analysis shall also be performed following snutctown, startup or a THERMAL POWER change that, within one hour, exceeds 15 percent of KATED THERMAL POWER unless: l. Analysis snows that the DOSE ECUtv** :.H I-131 concentration in the primary cool ant has l"Qt increased 1TCre than a factor of

2. The l'llble gas activity 1TCnitor snows that effluent activity has rot increased by 1TCre than a factor of three. ct. Tritium grab samples snall be taken at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months wnen tne refueling canal is flooded. e. Tritium grab shall oe taken at least once per 7 days from the ventilation exhaust from the spent fuel pool area whenever spent fuel is in the spent fuel pool. r f, The ratio of the samQle flow rate to the sampled stream flow shall be ltnown for tne time period covered by each ci>se or ci>se rate calculation in accordance with Specifications 3.ll.2.l, 3.ll.2.2 and 3.11.2.3.

SALEM -UNIT l 3/4 ll-10 Arrendtrent

-lo. 64

*

R!:ACIOR COOLJ.NT Slll'.IH BASIS 3/4.4.8 SPECIFIC ACTIVITY The limitation*

on the *pecif ic activity of th* primary coolant en*ure that the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months do*** at the *ite boundary will not exceed an appropriately

mall fraction of Part 100 limit* following a *team generator tube rupture accident ir conjunction an a**wned *teady *tate primary-to-*econdary

team generator leakage rate of l.O GPM. Th* value* for the limit* on *pecific activity repr***nt interim limit* ba*ed upon a parametric evaluation by the NRC of typical *it* location*.

Th*** value* are con**rvative in that *pecific *it* parameter*

of the Salem *ite, *uch a* *ite boundary location and meteorological were not con*idered in this evaluation.

Th* NRC i* finalizing

it* *pecific criteria which will be used a* the ba*i* for th* reevaluation of the *pecif ic activity limit* of thi* *ite. Thi* reevaluation may re*ult in higher limit*. generator tube below the to <SOO*F f'/X-vf'" ure"dnce of *urveillance requ.fJ.ement

/ vent* th* rel*a** of activity *hould a steam *aturation pr***ur* of the primary coolant i* tmospheric

team relief valv*** The id* adequate ***urance that exc***iv*

specific activity level* in th* primary coolant will be detected.

in *ufficient time to take corrective action. Information obtained on iodine *piking will be used to a***** the parameter*

a**ociated with *piking phenomena.

A reduction in frequency of i*otopic analy*** following power change* may be permi**ible if ju*tified by th* data obtained

SALEM -UNIT 1 B 3/4 4-5 Amendment No. 133

R.EAC'!'OR COOI.JJrr BASES 3/4.4.9 PR!SSUR!/T!MP!RATI.JR!

LIMITS Th* c..,.rature and pr***ure chan1** durin1 h .. tup and cooldovn are 11-ited to b9 con*i*tant vith th* requir ... nt* 11ven in the ASK! Boiler and Pressure Vessel Code,Section III, Appendix :. l) 'nle reactor coolant t .. perature and pr***ure and *Y*t .. heatup and cooldovn rate (vith the e.zception of th9 pr***uri&er) 1hall b9 lillited in accordance vith Piaur** 3.4-2 and 3.4-3 for the **rvic* p.riod specified thereon. a) Allowable coabination.1 of pru9Un and t...,_ratun for 1pecific t-.peratura chan1* rat** are b.lov and to the risbt of tha limit lin** 1hovn. Limit lin** for cooldovn ratu b9tv..n tho** pr***nted obtained by interpolation.

b) Piauru 3.4-and 3.4-3 dafin* U.aiu to U9Un pnvation of "°"'dvJilt-11H*neH-.'*

ilun only. Por norMl operation, other inherent pbnt charactari1 ic1, **I** pump heat addition ...S p\'988\&rilar hiaatar c cit y liai t the hutup and cooldown ratu that can b9 achieved over certain pre11ura*t..-:-*-

ature ran .... 2) TMH liait linH shall b9 calculated perloclically uain1 *thoU provided b9lov. 3) 'n\e Hcondary Iida of the 1tU. 1anarator mat not b9 prunr15ad above 200 p*il if tha t.-paratur*

of tba 1taaa 1.aerator i* b9low 70 r. 4) TM 0 pre*nrizer h9atup and cooldovn ratu *ball not ezcaecl l00°P/hr and 200 1/hr, rupact1'19ly.

'n\e 1pra7 *hall not b9 used if the t-.perature difgeranca b9tve9n the pressurizer and the 1pra7 fluid i* sr-.tar than 320 '* 5) Sy1t* pruarTica hydrotut1 and in*Hrvic*

leak and hydrotut1 shall be perfol'Md at pre11uru in accordance with the of ASK! Boil'er .iaad Va11al Coda,Section XI. The fracture touatm***

properti**

of the fr-:itic .. t*rial9 in the reactor ve1 .. 1 are det*raift*d in accordance vith tba NIC Standard leviav Plan, ASTM !185-82, and in accordance vith additional reactor ve***l requiremnt*.

These properti .. are tblD evaluated in accordance with G of tis. 1976 Su.mmer Addenda to Section III of th* ASH! Boiler and he18ure V*18*l Code and th* c&lculaUoa

thoda d .. crib9d in WCA.P-7924*A, "Ba.ii for Beatup and Cooldovn Limit April 1975". Heatup and cooldovn liait curve* are calculated usin1 tba moat liaitin1 value of the ail*ductility reference t.aperature, ITllD'r!_

at the end of 15 effective full power year* of service life. The 15 **rvic* life period ii chosen 1uch that th* li.mitin1 RTNnT at tha l/4T location in th* core re1ion i* 1reater than th* RT of th-a-liaitin1 unirradiated material.

The selection of such RTNDT a1*ures that all components in the Reactor Coolant Sy1t* vill y-o,.rated conservatively in accordance vith applicable Cod* requirements.

SAL!H -UNIT 1 B 3/4 4-6 .Amendment No. 108

?LA.NT S':'STSMS BAS::s ==============================================================================

3/4.7.9

.;11 s:: -..:bbe rs a re re qui red OP EAABLE to ensure that the st ruct u ra l integrity of the reactor coolant system and all other safety related systems is maintained during and following a seismic or other event initiating dynamic loads. Snubbers excluded from this inspection program are those installed on nonsafety related systems and then only if their failure or failure of the system on which they were installed, would have no adverse effect on any safety related system. A list of individual snubbers required to be operable per the technical specifications with detailed information of snubber location and size and of system affected shall be available at the plant in accordance with Section 50.7l(c) of 10 CFR Part 50. The accessibility of each snubber shall be determined and approved by the Station Operations Review Committee.

The determination shall be based on the existing radiation levels and the expected time to perform a visual inspection in each snubber location as well as other factors associated with accessibility during plant operations (e.g., temperature, atmosphere, location, etc.) and the recommendations of Regulatory Guide 8.8 and 8.10. The addition or deletion of any snubber shall be made in accordance with Section 50.59 of 10 CFR Part 50 . . The visual inspection frequency is based upon maintaining a constant level of snubber protection to systems. Therefore, the required inspection interval varies inversely with the observed snubber failur determined by the number of inoperable snubbers found during an ins I spections . b performed before that interval has elapsed may e used s re rence point to determine the next inspection.

The inspection are perfo med for each category of snubbers.

The snubbers are categorize b ibility (i.e., accessible or inaccessible during reactor operation).

The next visual inspection for each category may be twice, the same, or reduced by as much as two-thirds of the previous inspection interval.

This interval depends on the number of unacceptable snubbers found in proportion to the total number of snubbers in each category from the most recent inspection.

Intervals may be increased up to 48 months if few unacceptable snubbers are found in these inspections.

The visual inspection interval will not exceed 48 months. However, as for all surveillance activitie3, unless otherwise noted, allowable tolerances of 25% are applicable for snubbers.

Table 4.7-3 establishes three limits for determining the next visual inspection interval corresponding to the population of each category of snubbers.

For a category that differs from the representative sizes provided, the values for the next inspection interval may be found by interpolation from the limits provided in Columns A, B, and C. Where the limit for unacceptable snubbers in Columns A, B, or C is determined by interpolation and includes a fractional value, the limit may be reduced to the next lower integer. The first inspection interval determined using Table 4.7-3 shali be based upon the previous inspection interval as established by the requirements in effect before amendment (161). Any inspection whose results require a shorter inspection interval will override the previous schedule.

SALEM -UNIT 1 B 3/4 7-6 Amendment No. 161

DESIGN FEATURES ==============================================================================

DESIGN PRESSURE AND TEMPERATURE 5.2.2 reactor containment is designed and shall be maintained for a internal pres*sure of 47 psig. Containment air :emperatures

'..lp to 351.3°? are acceptable providing the containment pressure is accordance with that ln the UFSAR. S.3 RE.ACTOR '.:ORE fUEL ASSEMBLISS 5.3.l The reactor shall contain 193 fuel assemblies.

Each assembly shall consist of a matrix of zircaloy or ZIRLO clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide as fuel material.

Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used. fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 53 full length and no part length control rod assemblies.

The full length control rod assemblies shall contain a nominal 142 inches of absorber material.

The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing. 5.4.1 The reactor coolant system is designed and shall be maintained:

SALEM -UNIT 1 5-4 Amendment No. 198

-* area. 6 .13 fBXEiS a:mB)L m:x;RAM CPCPl 6.13.1 'lhe PCP shall be the Omnissicri prior to iq>lementaticri.

6 .13. 2 Lioensee initiated charqes to the PCP: 1. Shall be sul::mitted to the in the SEln:l.anruU .Radioactive Effluent Rel.ease Rsport for the period in lllhidl the chan;Je (s) was made. 'Ibis sul::mittal shall ocmtain: a. SUfficiently detailed infODDat.icri to totally S11AXJrt the :ratiaial.e for the chan;J8 wittoit benefit of aatitiaial or --suwlemental infonaticri: -_ __ __ _ _ b. A detem:inaticri that the c:::hm;Je did net redtnt the averall. a:mfcmnara!

of the solidified wasts pz:cduct to exi..st.irg

-criteria for solid wastes: an:! c. Documentaticri of the fact that the dlaD)9 bas bem reviewed ani founcl aoceptabJ A 'ty the SCH:. 2. Shall beoane effective upcri review an:! accept:anca by the SCH:. SAUM -UNIT 1 6-28 Amel alma 1t No. l

Unit 2 Mark-Up Pages

DEP!HlTIONS SEC'TION l . 0 DBFUlITIONS DEFINED TERMS ACTION ....... . AXIAL FLUX DIFFERENCE CHANNEL CALIBRATION CHANNEL CHEC!t CHANNEL FUNCTIONAL TEST CONTAINMENT INTEGRITY CORE ALTERATION . . . . Q:OSB EQUIVALENT I*l3l E*AVERAGB DISINTEGRATION ENERGY ENGINEERED SAFETY FEATORB RESPONSE TIMB FREQUENCY NOTATION . . . . . . . .

FULI..Y WITHDRAWN . . . . . . . . .

GASEOUS RAOWASTB TREATMENT SYSTEM IDBNTI PIED L.EAJCAGB . . . MEMBBR(S)

OP THB PUBLIC OFFSITB DOSB CAL OPERABLB

OP PHYSICS TESTS PRESSURE BOUNDARY t.BAXAGB PROCESS CONTROL PROGRAM (PCP) PURGB *PURGING . . . . . .

QUADRANT POWER TILT RATIO RATED THERMAL POWER REACTOR TRIP SYSTEM RBSPONSB TIM8 R.EPORTABLB iVBNT SHUTOOWN MARGIN SITE BOUNDARY SOLIDIFICATION SOURCE CHEClt . STAGGERBD TUT BASIS THBltMAI.

SIOWD UNIDBNTinm r..uuas UNRBSTRICDD ARD VENTII.ATlOll' UBAOST TRBATMENT SYSTEM VENTING SALEM -UNIT 2 I l-l l-l l-l l-l l-l l-l l-2 l-2 l-2 l*l l*l l-3 l-3 l-3 l-3 l-4 l-4 l-4 l-4 l-5 l-5 l-5 l-5 l-5 l-5 l-6 l-6 l-6 l-6 l-6 l-6 l-6 l-7 l-7 l-7 l-7 l-7 AmendmenC NO* 1 59

3/4.2 3/4.2.l 3/4.2. 3/4.2.4 3/4.2.5 3/4.3 3/4.3.1 3/4.3.2 3/4.3.3 3/4.3.4 AXIAL FilJX *********************************

3/4 2-1 2-5

'l'IIlI' RATIO *****************************

3/4 2-13 Dm 0 **************************************

3/4 2-16 'mIP SYS'l'!X

*****************

3/4 3-1 INS'JJllmfrATIClf

3/4 3-14 Radiaticl'l M:ni:t:cri.n; IrBt1'umBnt:ati

3/4 3-38 Irxx:::IE9 ******************************

3/4 3-42 :Rll!llct:e Instz:umant:atiai

3/ 4 3-43 Accider1t Jlt:rlitari.n; Instzuaattaticl'l

3/4 3-50 Radioactiw Liqdd Ettl\B'!t Jlt:rlitarin; ******* ******************************

3/4 3-53 Radioactiw Gaseous Ettl\B"Jt

!b'lit.crin:J ******* 0 ******************************

3/4 3-59 'nJRBDm ŽPE&> PIC11l!X:!fiCll

3/4 SAUM -UNIT 2

==================================================================

?AGE 3 . 3 3 I 4 o 3 * :. .:... :::. soc::<.c::s Ope::a:i:<g

.............................................

3/P\ sr.utjo;.i:<

..............................................

ONs::::

.:i...*:.

Jistribution

-Operatir.g

.........................

3/4 3-5 .:i...c. Jiscribut.:.:::in

-Shu:.down

..........................

3/4 8-9 2.25-Volt J.C. Distributior.

-Ope::atir.g

................

3/4 3-lJ J.C. Jistributior:

-Shu:down .................

3/4 3-12 29-Vol: D.C. Llistribution Cpe::ating

.................

3n 3-l3 25-'lolt D.C. Distribu::.:.on

-Sh'Jtdowr.

..................

3/4 3<5 3/4.8.3 E:ECTRI:AL EQUI?MENT

?ROTECTIVE DEVICES

?enetrat.:.on C .:ductor Overcurrent Pro:ect i ve Devices ....................................

314 8-16 .SALSM -c::: T 2 VIII Amendrr.ent No. 25 SJC't!OI l.8.2 3/4.9.1 3/4.9.2 3/4.9:3 1HDll LDUTDIC cgaptnon*rqa orQAttQI AID SlJIVl!WJfCI uau1u:mrrs WI ll'VJL11!Cj OPJIATIOHS BOROS COllCENTIATIOH

.********************************..

3/4 9

l INSTl.UM!HTATlotl

.******.***.....***************..*.....

3/4 9

2 . DECAY TIKI .....*..........**.***....****.....**........

3/4 9

l 3/4.9.4 CONTAINKlllT IUIIJ>IMC P!MITIATIOllS

- ...****..*.*.*.*...

3/4 9-4 3/4.9.5 COMMUMICATIOMS ._ ********...*........

3/4 9*5 3/4.9.6 MAMIPOL\TOR CUli& 3/4.9.7 CIANI TIAVIL *FOIL HA!IDLDIC A1lEA .*.**.******.**......

3/4 9.7 3/4.9.I ll!SIDUAL HEAT UKOVAL AND COOLAHT Ciactn.ATIOll All V&tar IAv*la .........

o ****************************

3/4 9*1 Water I.ave,l **.*.*.************* ************.****.

3/4 9-9 3/4.9.9 COHTAINMDT 1'\mGI AND PUSSUU*VAClrull all.Ill' ISOLATIOll SYST!'.11

3/4 9-10 3/4. 9.10 llAT!ll L!VIL

UACTOR V!SHL . *:; :-:* **. * *****.**.*****...

3/4 9-Il. 3/4.9.11 STOIAGI POOL llAT!a I.EVIL ***********.**.**..*****.*....

3/4 9*12 3/4.9.12 FU!L IWQ)LlllC AllA Vllrt'll.ATIOll SYSTDI ***********.*.**.

3/4 9-13 3/4.10 SPICW. TIU IXClltlOlfS 3/4 .10 .1 MAl.Glll ..****.....****..*******....***...*.*..

3/4 io -1 3/4.10.2 GaOU1 HIIGHT, INSEaTIOM ARD POVll. DISTllltJTlOI LIMITS ........*....*.*...................

3/4 10

2 3/4 .10. :S PllY'SICS TUTS **...........*.*.*...**..*....*.*........

3/4 10 .4 3/4.10.4 NO n.ov T!STS .........................................

3/4 10*5 ,. SALEK

UNIT 2 IX AaendMnc No. 78

tMCEX -BASES PAGE -SECTION APPLICABILITY

* * * * * * . . . . . . . . . . . . . . B 3/4 O*L 3/4.0 3/4.l REACTtVtif CONTROL SYST!l'41 3/4.1.1 3/4.l.2 3/4.l.l 3/4.Z 3/4".2.L 3/4.Z.2 and 3/4.2. 3/4.Z.4 3/4.Z.5 SOAATtON CONTROL * * * * * * * * * * * * * *

l!QlATtON SYSTiMS * * * * * * * * ** . . . . . l'ICVAIU CONTROL ASSE!eLIES . . . . . . . . . .

OIST'llI!UTION LIMITS 0 NI P AAAMrmlS * * * . . . . . . . . . . . .

SALEH

UNIT 2 XI * * * * . . . . 8 3/4 L*L 8 3/4 l-3 * * * *. 8 3/4 L-4 . . . . . . . . . . Amandm9nt ZS

BASES __ ;.,_ _______ --SECTION ) '4 ' 3 . l Arq) 3 ,'4 . 2 3/4.3.3 3/4.4

..........................

a AND ENGINEERED SAFETY FEATUU:S (ESFl rNSTRllMENTATION

...... . MONITORING INSIRtJMENTATION

............................

B REA,CTOR COOt..yrr SXSIEM 3 I 4 . 4 . l REACTOR COOLANT LOOPS AND COOLANT CIRCULATION

...........................................

B 3/4.4.2 and SAFETY VALVES ... : ......................................

B 3/4.4.3 3 I 4. 4. 4 PRESSURIZER

...........................................

B 3 I 4 . 4 . S RELIEF VALVES .........................................

B 3 I 4 . 4 . 6 STEAM GENERATORS

......................................

B 3/4.4.7 REACTOR COOLANT SYSTEM LEAKAGE ........................

B 3/4.4.B DELETED 3/4 3*t_JJ._,

3/tv 3/4 4-l 3/4 4-2 3/4 4-2 3/4 4-2 3/4 4-3 3/4 4-4 3/4.4.9 SPECIFIC ACTIVITY .....................................

& 3/4 4-6 3/4.4.lO PRESSORE/TEMPERATURB LIMITS ...........................

B SALEM -UNIT 2 XII 4 4-yf /8 g 3/4 Amendmanc No. l6L

3ASC:.S ==============================================================================

!? AG:: 3 . .; . :

.......................................... a 3/4 3, 4. s. 2

ccs

.......................................

3 3 / : -3/4.S.3

' ....... --..... 8 o/4 S-2 3/4.S.5 .-,--**-r

"*'A---, . .,. *-,*1s-1 :::i 3;*' --;3 :"..:..: '...;:..._.

__ h...J '"' !.:..:'\.

> .. v:'\Ml..J':..

.:"\1.lr\

1 :'\.t' * .: , . *.,,.,, .. ,, .. ,,., :::::> --: :,-3/4.6 3/4.6.l

...................................

B 3/4 6-1 3/4.6.2

..................

3 3/4 6-3 3/4.6.3

SOLATICN

..........................

8 3/4 6-4 3/4.6.4 CCM3CST:3LE GAS CON!ROL ...............................

3 3/4 6-4 :c::

i.

m:::x * ==============================================================================

?AG:: ' 1 -"1. ' : 3 : c::. *= c:..:: . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3 314 .., -

............................................

3 1_4 3,* 4 -3/4. 7. 4 s:::Rv:c::: "ti.:;:::::i..

..................................

3 3/4 7-4 3/4.:.s :::..coo ?Ro::::cT:::ON

......................................

3 3/4 .., -,-J 3/4.7.6 CCNTROL ROOM :::MERGENCY A:R

...................................

3 3/4 .., -I -J 3/4.7., AUXIL:ARY 3UILDING AI?.. :IL7?...Zl.TICN SYSTSM .....................................

B 3/4.7.8 SEALE.u SOU?..CE.

CONTAMINATION

...........................

B 3/4. 7. 9 S>JC33E.RS

..............................................

B 3/4.7.:0

-

SYS7::M ............................. 3 3/4 -:-3 3n. 8.: ,; . C. S*JCRCES ..........................................

3 3/4 :3

3/4.3.2 PCWE.R

.....................

3 3/4 3-l 3/4.3.3 ::::E.CT?..ICAL EQUIPMENT

...............

3 31 ., "-

-

T 2 X!'i Amendir.ent No.

INDEX DESIGN FEATURES ==============================================================================

SECTION PAGE 5.1 SITE ion ea .............................................

5-1 ow Population Zone ........................................

5-1 AllSAS l=oR.. f<.l'.l>lO.O.C.T1v*

G.At'E:OUS AND Eii=FLuf:/llTS 5-1 5 Configuration

..............................................

5-1 Design Pressure and Temperature

............................

5-4 5.3 REACTOR CORE Fuel Assemblies

............................................

5-4 Control Rod Assemblies

.....................................

5-4 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature

............................

5-4 Volume .....................................................

5-4 5.5 METEOROLOGICAL TOWER LOCATION ..............................

5-5 5.6 FUEL STORAGE Criticality

...............................................

5-5 Drainage ...................................................

5-5 Capacity ..................................................

5-50.., 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT ........................

5-Sb SALEM -UNIT 2 XVII Amendment No. 28 Kl:::-n:-.;r ST?..AT!VE CONTROLS =======================================================================

S"E:CTION 6.l ? ..

..............................................

6-1 5.2 ORGANIZATION Onsite and Offsite Organizations

...........................

6-1 Facility Staff .............................................

6-1 Shift Tech...'1ical Advisor . . . . . . . . . . . . . . . . . . . . . . . . . . . .

....... 6-7 6.3 FACILITY STAFF QUALIFICATIONS

................................

6-7 6.4 TRAINING ...................................

'. ...............

6-7 6.5 REVIEW AND AUDIT (THIS SECTION DELETED) ....................

6-8 6.6 REPORTABLE EVENT ACTION 6.7 SAFETY LIMIT VIOLATION 6.8 PROCEDUP .. ES AND PROGRAMS 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS 6.9.2 SPECIAL REPORTS 6. 10 RECORD RETENTION

.......................................... . 5.11 RADIATION PROTECTION PROG? .. AM 6.12 RADIATION AREA ....................................... . 5 . 13 PROCESS CONTROL PROGRAM (PCP) ............................. . 5.14 CFFSITE DOSE CALCULATION MANUAL fODCMl 6.15 CHANGES TO RADIOACTIVE LIQUID, GASEOUS AND SOLID WASTE TREATMENT SYSTEMS ........................ . 6-16 6-16 17 2D 6-;4 2f 6-'¥ 2..,S 6-;A 21 SALEM -UNIT 2 XVIII Amendment No.175

DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when: 1. 8 1.7. l All penetrations required to be closed during accident conditions are either: a. Capable of being closed by an OPERABLE automatic isolation valve system, or 1.7.2 All equipment hatches are 1.7.3 Each air lock is OPERABLE pursuant to Specification 3.6.1.3, 1.7.4 The containment leakage rates are within the limits of Specification 3.6.1.2, and 1.7.5 The sealing mechanism associated with each penetration (e.g., welds, bellows or 0-rings) is OPERABLE . NCT USED CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

DOSE EQUIVALENT I-131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The SALEM -UNIT 2 1-2 Amendment No. 172 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2. l The reactor trip system instrumentation setpoints-shall be set consistent with the Trip Setpoint values shown in Table 2.2-1. APPLICABILITY:

As shown for each channel in Table 3.3-1. ACTION: trip system instrumentation setpoint less conservative than the o the Allowable Values column of Table 2.2-1, declare the.channel in e ao e a apply the applicable ACTION statement requirement of t on 3.3. ljunti the channel is restored to OPERABLE status with its trip etpoint adjus ed consistent with the Trip Setpoint value. SALEM -UNIT 2 2-4 FUNCTIONAL UN l T 13. Steam Generator Water Level--Low-Low

14. Deleted 15. Undervoltage-Reactor Coolant Pumps 16. Underfrequency-Reactor Coolant Pumps 17. Turbine Trip A. Low Trip System Pressure B. Turbine Stop Valve Safety Injection Input from SSPS E:s...C 19. Reactor Coolant Pump Breaker Position Trip :;/\LEM -UN IT 2

REACTOR THI P SYSTEM INSTRUMENTATION TRIP SETPOINTS TR!P SETPOINT ? 9.0: of narrow range instrument span-each steam generator 2 2900 volts-each bus 2 56.5 Hz -each bus 2 45 psig s 15t off full open Not Applicable Not Applicable ALLOWABLE VALUES 2 8.0* of narrow range instrument\

steam generator 2 2850 volts-each bus 2 56.4 Hz -each bus 2 45-psig s off full open Not Applicable Not Applicable Amendment No. 154

CONDITION FOR OPERAT:ON J.0.l Compliance wlth the llmltlng Conditions for Operation contalned succeeding speclfications is required during the OPERATIONAL.

MOOES or conditlon*

thereln; except that upon failure to the Condltlons for Operation, the associated ACTION ahall be 3.0.2 Noncompliance with & *p*cification sh&ll *xi*t when the r*qulr*ments o: th* Limiting Condition for Operation and a**ociated ACTION requirements are not met within the specified tim* interval*.

If the Limiting conditlon

or Operation i* r**tored prior to expiration of the *pecified time completion of the ACTION requirement*

i* not required.

J.O.J When a Limiting Condition for Operation i* net met except a* provided in th* a**cciated ACTION requirement*, within one hcur action *hall be initiated to place the unit in a MODE in which th* *p*cif ication do** not apply by placing it, a* applicable, in: l. At lea*t HOT STANDBY within the next 6 hcur*, 2. At lea*t HOT SHUTOOWN within the following 6 hour*, and J. At lea*t COLD SHUTDOWN the sub*equent 24 hour*. Where corrective m*a*ur** are completed that permit operation under the ACTION requirement*, the ACTION may be taken i.1 accordance with the *pecif ied tlme limit* a* measured from th* time of failure to meet th* Limiting Condition

or Operation.

Exceptions to these requirements are stated in th* individual specifications.

J.0.4 Encry inco *n OPERATIONAL KODE or ocher *P**itied (a) shall not be mad* when the conditions of the Limiting Condition for Operation are net met and the associated ACTION require* a shutdown they are not met within a specified time interval. (b) cn.ay be made in accordance with ACTION requirements when conformance to th.a permit* continued operation of the facility for an unlimlted period of time. Thi* provi*ion shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirement*.

Exception*

to th*** requlrements are stated in the individual specifications.

SAL.EM -UNIT 2 3/4 0-l 1 "*\ Amendment: ... ......

H1. l ru:ACT:::VI"rY CONTROL

! 14...J. l BORATION CONTROL MARGIN ... > 200°F . avg i * .1.MITING FOR OPERATION 3.l.l.l The SHUTDOWN MARGIN shall be greater than or equal to l.6t delta k/k. AfPLic:ASILIT"f:

MOCSS l, 2*, 3, and 4. ACTION: With the SHUTDOWN MARGIN less than l. 6t delta k/lc, immediately inii:.,ate and continue boration at 33 gpm of a solution containing 6,560 ppm boron or equivalent until the requ*red SHUTDOWN MARGIN is restored.

iURVE tLl.A?lCE REQUIR.EME?n'S 4 . l . l. l . l The SHUTDOWN MARGIN shall* be detentined to .be greater than or equal to l.6t delta k/k: a. Within l hour after aetectiou of an inoper&ble

.:o."ltrol rod(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod(s) is inoper&ble.

If the inoper&.ble control rod is immov&ble or untripp&ble, the above required SHCTDOWN MARGIN shall be increased bf an amount least equal to the withdrawn of the immov&ble or untripp&ble control rod(sl. b. c. When in MOCB 2 with K ff les* than l.O, within 4 hour* prior to achieving crificality by verifying that the predicted critical control rod position is within the limits of Specification 3.l.3.5. *See Special Test Exception 3.lO.l SALEM

UNIT 2 3/4 l*l No. 133 RF.ACIIVIIY CONnOL SYSTOO MOQWIOR IOOF.:RAnJU COEFIICI£Nl LIMITING CONDITION FOR OPERATION 3.1. t.3 a. b. 'nle moderator temperature coefficienc (KTC) shall be: Less positive than 0 delu k/lt/9F for the all rods vithdrawn, beginning of cycle life (SOL), hot z..:ro IBERXAL POIJER cond:

-4 Less negative than -4.4 x 10 delta k,11t/*r for the all rod.a vithdravn, end of cycle life (EOL), RATED THERMAL PO\Jl:R condition.

APPLICABILITY:

Specification 3.1.l.3.a

-MODES 1 and 2* only# Specification 3.1.1.3.b

-MODES 1, 2 and 3 only# ACTION: a.

the MTC more positive than the limit of 3.1.1.3.a, above, operations in 1 and 2 may proceed provided:

1. Control rod withdrawal limits are and maintained sufficient to restore the KTC to less positive than 0 delta k/lt/"F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be S SY in next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . These withdrawal lim s shall be in Bditi o the insertion limit.s of Specific&

ion 3.1.3.6.

3, I. 3,5'. 2. 'nle control rods are maintained within the withdrawal liJlits established a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods vithdrawn condition.

3. In lieu of any othet report required by Specification 6.9.1, a Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 vithin 10 days, describing the value of the measured KTC, the interim control rod vithdrawal limits and the predicted average core burnup necessary for restoring the positive KTC to within its limit for the all rods vithd.rawn condition.

b. Yith the KTC more negative than the limit of 3.1.1.3.b, above, be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Keff greater than or equ&l to 1.0 #See Special Test Exception 3.10.3

UNIT 2 3/'* 1-4 Amendment No. 94 REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOl/ABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING .CONDITION FOR OPERATION 3.1.3.l All full length (shutdown and control) rods, shall be OPERABLE and within = 18 steps (indicated position) when reactor power is s 85\ RA7ED POWER, or +/- 12 steps (indicated position) when reactor power is > 85\ RATED 7HERMAL POWER, of their group step counter demand position within one hour rod motion.

MODES l* and 2*

a. With one or more full ,length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.l is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . b. With more than one full length rod inoperable or mis-aligned from the group step counter demand position by more than +/- 18 steps I (indicated position) at s 85\ RATED THERMAL POWER or +/- 12 steps (indicated position) at > 85\ RATED THERMAL POWER, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . c. With one full length rod inoperable due to causes other than addressed by ACTION a, above, or mis-aligned from its group step counter demand position by more than +/- 18 steps (indicated position) at s 85\ RATED THERMAL POWER or +/- 12 steps (indicated position) at > 85\ RATED THERMAL POWER, POWER OPERATION may continue provided that within one hour either: 1. rod is restored to OPERABLE status within the above requirements, or 2. 3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.l is satisfied.

POWER OPERATION may then continue provided that: *see Special Test Exceptions 3.10.2 and 3.10.3. SALEM -UNIT 2 3/4 1-13 Amendment No. 183

REACTIVITY CONTROL SYSTEMS POSITION HffHOti'fIOH 9Y9'fEM 9H"J'f1'eW e.. SuuTOOCA)tJ

(<DO UMIT LIMITING CONDITION FOR OPERATION 3.1.3.4 All shutdown rods shall be FULLY WITHDRAWN.

APPLICABILITY:

MODES l*, and 2*#@. ACTION: With a maximum of one shutdown rod not FULLY WITHDRAWN, except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hour either: a. FULLY WITHDRAW' the rod, or, b. Declare the rod to be inoperable and apply Specification 3.1.3.1. SURVEILLANCE REQUIREMENTS 4.1.3.4 Each shutdown rod shall be determined to be FULLY 'WITHDRAWN by use of the group demand counters, and verified by the analog rod position indicators**:

a. Within 15 minutes prior to withdrawal of any rods in control banks A, B, C, and D during an approach to reactor criticality, and b. At least once per 1.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thereafter.

See Special Test Exceptions 3.10.2 and 3.10.3. ** For power levels below 50% one hour thermal "soak time" is permitted.

@ # During this soak time, the absolute value of rod motion is limited to six steps. Surveillance 4.1.3.4.a is applicable prior in preparation for startup (Mode 2). With Keff greater than or equal to 1.0. to withdrawing any control banks Note: This page effective prior to startup from fifth refueling outage scheduled to begin March 1990. Letter dated Jan. ll, 1990. SALEM -UNIT 2 3/4 1-19 Amendment No. 66

REACTIVITY CONTROL SYSTl.HS pgamQN nmIGATION SYS'nlf sHU'fDOWN

C.Of'JTML Rot) "INSER-r10N UMrrS limited in physical u shown it. ACTION: With the control ban&. iMarted beyond the above iM*rtion limits, ucapt for surveillance t .. tina pursuant to Specification 4.1.3.1.2, either: a. R .. tora the control ban&. to within the limit* within two hours, or b. Reduce TH!RMAL POWER within two hours to l89s than or equal to that fraction of RATED TH!RMAL POWEi which i* allowed by bank position usina the above figur .. , or .. .r-c* c. Be in at least Har STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . SURV!IU.ANCI R!QUIR!K!NTS 4.1.3.5 Th* position of each control bank shall be determined to be the insertion limit* at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by use of the 1roup demand counters and varif iad by the analo1 rod poaition indicators**

ucapt during ti.ma interval*

when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod poaitica.

at least once per 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s**. *S** Special T .. t Exception.

3:10.2 and 3.10.3 **Por power level* below 50% one hour thermal "soak time" is permitted.

Durina this soak time, the absolute value of red motion is limited to six stapa. I/With IC.aff greater than or equal* to 1.0 SALEM -UNIT 2 3/4 1-20 Amandmant No. 80

K lllSERTIOH VG.SUS A THRE LOOP OPWTIOH SAL.EM

UHIT Z 3/4 l-2.%

L!M!iS b. 111 r1ot be

&Cove 90S of llAT'ED 11"1d1c:atad AFa 1s w1tilin tM +6, -9t. target band and Vt . c:. 11 i r1ot be ll"ICl"Hstd

&Cove s= of lATtn THEmL unless Ule 1nd1cat.ad AFa b**n outside of tile +6, -9t.

band for .,,.. 1 "°ur penalty dev11tion during Z4 Power 1nc,..11ts .Cove !OS of llATttl THER)IAL PO'tiU do r1ot wiU'lin the b&nd p.-ov1d1d Ult 1c=.,.ul1tivt penalty deviation 11 not viol1t.ad.

4. Z. 1. 1 The indic:attd AXIAL FLUX sn111 be d1t1T"!l1ned to be within its limits during ?O'tt'ER OPER.ATION .Cove 151 of RAT!l) TMERMAL PO'efER by: 1.

U'l1 1nd1cat.ad AFC for Heh OPERABLE 1.xco" channel: 1. At 111st onct ptr 1 days when the AFO Ala,.. is anc z. At 111st onc1 ptr "°ur for tJ'I* f1""t 24 noul"1 aftar ?"Wstorin; Ult AFO Monitor Alaf"9 to OPEWL! IUtus. "4onitorin; and lo;;1n; tJ\e 1nd1catad AXIAL FlUX OIF1tRENCt for c:nanntl at. 111st one* ;>er for the f1r1t. 24 hour1 and at lust onCI p1r 30 ainut.as tne1*ufur, wf'l1n U\e AXIAL Fl.UX A1af"9 is The va1uet of ind1c1t.ad AXIAL OIFFtl!MCE 1n.ll be 111.aied ta exist inta,.._11 ucl'l

'* z. 1. z The il'ldic:aud AFu shall be outJ1dt of 1u .0, -9i target band wf'ltn at hut z OT' ac" OPEAAll.!

uco" d\Mfteh '"

t"9 AFO to be outside t.1"11 band. Ptnalty deviation outside of tM +6., .gi target band ,n111 on a t.1 .. basis of: 1. One air1ut.a pef'alty deviation for each one 1inut* af autsidt of Ult b&ftd u. TI4E.R*L ,..,.ts equ.1 t4 01" ICOVI SOS of UTED llOW'ER, Ind ti.

a1nut.a de.,1at1on for 11cl'l one 1inui. of POWER out..sidt af Ult UrtJlt band It , * .,.is of Ur-ED PQlw'ER .

2 l/4 Z*Z Amendment

'40. S

ACTION lO -ACTION J.l -ACTION l2 -ACTION 13 -ACTION 14 -*

P-6 P-7 SALEM -UNIT 2 J.J-l !Continued)

With the number of OPEJV.BLE Channel* one l*** than the Minimum Channel* OPERABLE requirement, r**tore the inoperable channel to OPERABLE *tatu* within 6 hour* or be in at l*&*t HOT STANDBY in the next 6 hour*; however, one channel may be bypa***d for up to 4 hour* for *urveill&nce t**tinq per Specification 4.3.l.l.l provided th*

channel i* OPERABLE.

With l*9* than the Minimum Number of Channel* OPERABLE, operatLon may continue provided the inoperable channel i* placed Ln the tripped condition within 6 hour*. With th* number ot channel* OP!RABLJ:

one l*** than required by the Hinimwn Channel* OPERABLE requirement, r**tore th* inoperable channel to OPERABLE *tatu* within 48 hour* or be in HOT STANDBY within th* next 6 hour* and/or open the reactor trip breaker*.

With the number ot OPERABLE channel* one l*** than the Minimum Channel* OPERABLE requirement, r**tore the inoperable channel to OP!RASLZ *tatu* within 48 hour* or open the reactor trip breaker* within the next hour. With one of the diver** trip feature* (Undervoltaqe or *hunt trip attachment) inoperable, r**tore it to OP!RASLZ *tatu* within 48 hour* or declare the breaker inoperable and be in at lea*t HOT STANDBY within 6 hour*. Th* breaker *hall not be bypa**ed while one of the diver** trip feature* i* inoperable except tor th* time required for performinq maintenance to r**tor* th* breaker to OP!RASLZ *tatu*. R!!ACTOR TRIP SYSTEM INTIJU.OCM COND III ON MD S!TP9INT

MCTION With 2 of 2 Intermediate Rani!i Neutron Flux Channel* < 6x10 amp*. 2 of 4 Power Ranqe Neutron 18 ;e: ll\ of RATED POWER or l of 2 TYrbin* l** chamber pr***ure channel* ;e: a pr***ure equivalent to 11\ of RATZD THEJUllAL POWER. ._ 3/4 3-7 P-6 prevent* or def eat* th* manual block of *ource range reactor trip. P-7 prevent* or d*f* the automatic block 1 reactor trip ona Low flow in more than one primary coolant loop, reactor coolant pwDl>> undervoltaqe and under-frequency, pre**uri:*r low pre**ure, pre**uriz*r hiqh level, and th* openinq of more than one reactor coolant pump breaker. Amendment No.121 TABLE 4.3-1 REACTOR TRIP SYSTEM CHANNEL FUNCTIONAL UNIT CHECK 1. Manual Reactor Trip Switch N.A. 2. Power Range, Neutron Flwc s 3. Power Range, Neutron Flux, N.A. High Positive Rate 4. Power Range, Neutron Flux, N.A. High Negative Rate 5. Intermediate Range, Neutron Flux s 6. Source Ranl!{e, Neutron Flux 50l 1. Overtemperature 6T s 8. Overpower 6T s 9. Pressurizer Preaeure--Low s 10. Pressurizer Preseure--High s 11. Pressurizer Water Level--High s l 12. Lose of Flow -Single Loop s SALEM -UNIT 2 3/4 l-11 CHAN.l'JEL N.A. o<21, tt1>* and 0"1 a111J R"I atbl Rl61 R R M R R R Q TEST R,,,, 0 0 0 s /tJll and S/tJ 11 0 0 Q Q 0 Q I i MODES IN WHICH SURVEILLANCE REQUIRED 1, 2, and

1, 2 l, 2 l, 2 l, 2 and

2, 3, 4, 5 and

l, 2 l, 2 l, 2 l, 2 l, 2 1 Amendment No. 1 I TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS FUNCTIONAL UNIT 13. Loss of Flow Two Loops 14. Steam Generator Water Level--Low-Low

15. Deleted 16. Undervoltage

-Reactor Coolant Pumps 17. Underfrequency

-Reactor Coolant Pumps 18. Turbine Trip a. Low Autostop Oil Pressure b. Turbine Stop Valve Closure 19. Safety Injection Input from ESF 20. Reactor Coolant Pump Breaker Position Trip 21. Reactor Trip Breaker 22. Automatic Trip Logic SALE!' NlT 2 CHANNEL CHECK s C!iANNEl CALIBRATION R s R N.A. R N.A. R N.A. N.A. N.A. N.A.

N.A. N.A. N.A. N.A. N.A. N.A. N.A. N.A. I -12 CHANNEL FUNCTIONAL TEST N.A. Q Q Q S/U'" S/U'" wt101 R Ml)J11111111 and R 1141 M1)1 MODES IN WHICH SURVEILLANCE REOUIBED 1 l, 2 l 1 l>k-A-. 1 Z, l, 2 l, 2 l, 2 I/ * >l l Amt:ua 1 and and *

Nu 1'1/

KAJIATIJN MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.:

instrumentation shown in Table 3.3-6 shall be OPERABLE with their alarm/trip setpoints within the specified As shown in Table 3.3-6. ACTION: a. With a radiation monitoring channel alarm/trip setpoint exceeding the value shown in Table 3.3-6, adjust the setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.

b. With one or more radiation monitoring channels inoperable, take the ACTION shown in Table 3.3-6. c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.1 Each radiation monitoring instrumentation channel shall be strated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-3. .3.:'.l.L.E:*l

-\.:NIT 2 3/4 3-38 ""'It'*

TABLE 3. Continued RADIATION ING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARM/TRIP INSTRUMENT OPERABLE MODES SETPOINT 2. PROCESS MONITORS b. Noble Gas Effluent Monitors 1) Medium Range,Awciliary 1 1,2,3&4 :s;3. Ox10-2µCi/cm 1 Buiiding Exhaust System (Alarm only) (Plant Vent) 2) High Range Awciliary 1 1,2,3&4 :s;l. Oxl0 2µCi/cm 1 Building Exhaust System (Alarm only) (Plant Vent) 3) Main Steamline 1/ -1,2,3&4 10 mR/hr Discharge (Safety MS Line (Alarm only) Valves and Atmospheric Steam Dumps) 4) Condenser Exhaust 1 1,2,3&4 :s;7 .12x10* cpm System (Alarm only) 3. CONTROL ROOM a. Air Intake -2/Intake##

- :s;2. 48xl0 1 cpm Radiation Level ## Control Room air intakes shared between Unit 1 and 2. ** ALL MODES and during movement of irradiated fuel assemblies SALEM -UNIT 2 3/4 3-39a MEASUREMENT RANGE 10-1-10 1 µCi/cm 1 10-1-10' µCi/cm 1 1-10* mR/hr 1-10 6 cpm 10 1-10 7 cpm C.ORE A.LTEJ2,<\il'Df'lS coJ<e alterat:ions fl--ACTION " 26 26 26 26 Amendment No. 173 TABLE 3.3-6 (Continued)

TABLE NOTATION ACTION 23 -With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . ACTION 24 -With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.7.1 . . ACTION 25 -With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.9. ACTION 26 -With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements, initiate the preplanned alternate method of monitoring the appropriate parameter(s), within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and: 1) either restore the inoperable Channel(s) to OPERABLE status within 7 days of the event, or 2) prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status. ACTION 27 -With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel(s) to OPERABLE status within 7 days or initiate and maintain operation of the Control Room Emergency Air Conditioning System (CREACS) in the pressurization or recirculation mode of operation.

CORE ALTERATIONS and movement of irradiated fuel assemblies will be suspended during operation in the recirculation channels OPERABLE in a Control Room air intake, diately initiate and maintain operation of the CREACS in the pr ssurization or recirculation mode of operation.

CORE ALTERATIONS and movement of irradiated fuel assemblies will be during operation in the recirculation mode. SALEM -UNIT 2 3/4 3-40 Amendment No.173

ACCIDENT LIMITING CONDITION FOR OPERATION

--3.3.3.7 The accidenc monicoring channels shown in Table 3.3-11 shall be operable.

APPLICABILITI:

M.ODES 1, 2, and 3. ACTION: a. As shown in Table 3.3-11. b. The provisions of Specification SURVEILU\NCE REQUIREMENTS 4.3.3.7 Each accident m t ring instrumentation channel shall be demonstrated O ERA.BLE by performance of the CHANNEL CHANNEL CALIB TION(operations at the frequencies shown iA Table 4.3-11. C>>AN.NEL.

F()N vTtoNI\ 1-TES't SALEM -UNIT 2 3/4 3-50 Amendment No. 95

TABLE 3. 3-11 ACCIDENT MONITORING INSTRUMENTATION INSTRUMENT

1. Reactor Coolant Temperature

-THot (Wide

2. Reactor Cooiant Inlet Temperature

-Tcow (Wide Range) *3. Reactor Coolant Pressure (Wide Range) 4. Pressurizer Level 5. Steam Line Pressure 6. Steam Generator Water Level (Narrow Range) 7. Steam Generator Water Level (Wide Range) 8. Refueling Water Storage Tank Water Level 9. deleted 10. Auxiliary Feedwater Flow Rate 11. Reactor Coolant System Margin Monitor 12. PORV Position Indicator SALEM -UNIT 2 L ______ _ REQUIRED NO. OF CHANNELS 2 2 2 2 2/Steam Generator Generator) 4 (!/Steam Generator) 2 2/valve**

3/4 3-51 MINIMUM NO. OF CHANNELS ACTION l 1, z .i' l 1, 2 1 1, 2 1 1, 2 l/St*eam "Generator 1 I 2 l/Steam Generator 1, 2 3 (l/Steam Generator) l I 2 1 1, 2 3 (!/Steam Generator) 4, 6 1 1, 2 l 1, 2 Amendment No. 125

TABLE 3.3-11 (continued) 7.".BLE 1 l"lith t!1e :-ii..:mbe.:::

of C?ERA3LE accident rnonitor.:.r:g

-:a.nnels less than the ?.eq:.:i.:::ed Nu::-be.:::

of Channels shown in "?a8le 3. 3-11, restc:-e the inoperable channel ( s) to O?ERABLE status wi :'.-:ir: 7 days, or be .'..:-. HOT SHUTJOWN

he next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2 Wi ::.h the nu:r.ber of O?ERABLE accident rnoni toring :::-.a::r:els less than the Number of Channels showA in Table 3.3-11, resto:-e t!:e inoperable channel (s) to OPERABLE status wi::.'."'.i.:;

48 ::curs or "::::e i.n HOT within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ACTION 3 deleted ACTION 5 Tab.:._ 3.3-1, ay oceed provide an OPERABLE Ste channel is available as an alternate means of indication Steam Generator with no OPEEABLS Auxiliary feedwater flow Rate Channel. number of O?ERJ..BLE channels less than the Requi.:::ed Number Channels shown in Table 3.3-11, operation may proceed Tables a.:::e available in the Ccr:trol Room and fcllowir.g

_ . .:.nnels in Table 3. 3-ll a.:::e OPEPJ*.3LS tc provide a:: al:e.:::r.a:e o: calculating Reac::.o:-

Coolant System subcooling margin: a. Reac:or Coolant (Wide Range) b. Reacto.:::

Coolant Pressure (Wide Range) SALEM -UNIT 2 3/4 3-Slb Amendment No. 125

.I. *: *

TABLE 4.3-11 SURVEILLANCE REQUIREMENTS FOR ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHANNEL INSTRUMENT CHECKS CALIBRATION

1. Reactor Coolant Temperature

-THOT (Wide Range) 2. Reactor. Coolant Inlet Temperature

-TcoLD Range) 3. Reactor Coolant Pressure (Wide Rangel 4. Pressurizer Water Level 5. Steam Line Pressure 6. Steam Generator Water Level (Narrow Range) 7. Steam Generator Water Level (Wide Range) 8. Refueling Water Storage Tank Water 9. deleted 10. Auxiliary Feedwater Flow Rate 11. Reactor Coolant System Subcooling Margin Monitor Level M R M R cJ-.,,:t .. f M R M R M R M R M R M R R

CHANNEL FUNCTIONAL Ji 11 ll \\" ->>KN. A. ...NK N. A. .)Ur' /\/

A * ...NtVr-/\I. A * /\J. A, Ail<<

JJA ,,.J.,.A *

Auxiliary Feedwater System is used on each startup and flow rate indicacion is verified at that time. *The instr11ments to develop RCS subcooling margin are calibrated on an 18 month cycle; the monitor will be compared quarterly with calculated subcooling margin for known input values. SALEM -UNIT 2 3/4 3-52 Amendment No. 125 r

TABLE 4.3-11 (Continued)

SURVEILLANCE REQUIREMENTS FOR ACCIDENT MONITORING INSTRUMENTATION INSTRUMENT

12. PORV Position Indicator

13. PORV Block Valve Position Indicator

14. Pressurizer Safety Valve Position Indicator

15. containment Pressure -Narrow Range 16. Containment Pressure -Wide RaPge 17. Containment Water Level -Wide Range 18. Core Exit Thermocouples

19. Reactor Vessel Level Instrumentation System (RVLIS) CHANNEL CHECKS M M M M M M M M

Unless the bluck valve is closed in order to meet the requirements of Action b, or c in specification 3.4.5. 3/4 3-52a SALEM -UNIT 2 JiJA:N.A. -NK N. A. .NK-N. A, R R

t H.ANNEL FLNCTl-')NAL TEST R Q* R ...AAN.A.

..NA-N .. A,. -N1r N. A. .l>IA-N.. A. N. A, Amendment No. 136

* ( l) TABLE 4.3-12 (Continued)

TABLE NOTATION Tne CHANNEL FUNCTIONAL TEST shall also dermnstrate that auto1111t1c fsolation of tnis pathway and control roo11 alar11t annunciation occurs 1f any of the fol lowing cond1tions exist: l. Instruawnt ind1cates 119asured lt.,.ls at or abo'4 the 1llr*/trip S@tpo int. 2. Circu1t failure. (Loss of Po*r) (Autoll9t1c Isolation only) 3. Instrument ind1cates a ct>wnscalt failure. (2) The CHANNEL FUNCTIONAL TEST shall also dtimnstrata tl't1t 1*00* 1llrin annunchtion occurs ff any of tht following conditions tx1st: l. Instru ... nt fnd1cltts "91SUrtd lt.,.ls at or aba¥t*th9 1l1rll/trip sttpo f nt. 2. Cf rcuft fa11urt. (Loss of Po*r) (Indic1t1on only) 3. Instru,,.nt 1nd1c1tes 1 downscllt failure

4. InstruNnt controls nat stt in operate rmdt. (3) Tht fnitfal CHANNEL CALIBRATION was perfor!IWd using apprapr11tt liquid or gaseous calibration sources obt1i"9d reput1blt supplitrs.

Tht activity of th* e111br1tion sourcts were reconfir"9d using a nult1-ch1nntl analyzer wnfch was c1lfbr1ttd using Ont or flt:>rt NBS standards

. (4) CHAHNEi. CHECK shall consist of "9r1fying indication of flow during periods of release. CHANNEL CHECK shall be lftldl at least onca per 24 "->urs on days on wnicn contf nuous, per1od1c, or bitch releases ll't *dt*

Durfng lfqufd 1dd1t1ons to the tank. .. If tank lt'lel 1nd1cat1on 1s not providtd, .,.rt1f1cat1on wt11 be ct>nt by visual 1nspect1on.

fs an off-lfne channel whfch requires periodic decontami natfon Any count rate indication abo Y9 l0,000 cp"' ccnst1 tutu a CltAHNELeCHECK f ,. coiapl i ance purposes.

SALEM -UNIT 2 3/4 3-58 Amendment No. 36

?

TABLE 4 . 3 -13 RADIOACTIVE GASEOUS EFFLUENT MQNITORING INSTRVMEtffATION SURIJEILI.ANCE REOUIREMBtffS CHANNEL RCE CHANNEL FUNCTIONAL INSTRUMENT CK CALIBRATION TEST 1 . WASTE GAS HOLDUP SYSTEM a. Noble Gae Activity Monitor Providing p p R (3) Q(l) Alarm and Automatic Termination of Release b. Oxygen Monitor D N.A. Q(4) M 2. C01ITAINMENT PURGE AND PRKSSURE -VACUUM RELIEF a. Noble Gae Activity Monitor Providing p p R (]) Q(l) Alarm and Automatic Termination of Release 3. PLANT VENT HEADER SYSTEM# a. Noble Gae Activity Monitor D M R (3) Q (2) b. Iodine w N.A. N.A. N.A. c. Particulate Sampler w N.A. N.A.

N.A. d. Flow Rate Monitor D N.A. R N.A. e. Sampler Flow Rate Monitor w N.A. R N.A.

IN WHICH SURVEILLANCE REQUIRaD * ** *** * * * * * # The following process streams are routed to the plant vent where they are effectively monitored by the instruments described: (a) Condenser Air Removal System (b) Auxiliary Building Ventilation System (c) Fuel Handling Building Ventilation System (d) Radwaste Area Ventilation System (e) Containment Purges SALEM

UNIT 2 3/4 3-63 Amendment No. 138

IHSIRtJKEN'IATION SUllV!ILLAMC!

8.EQUIR!MENTS (Continu.d) 4.3.4.3 'Ibe above required turbine overspeed protection syst .. shall be demon.trated OPERABLE:

a. Ac le .. t one* per 18 month* by performance of a CHANNEL CALIBRATIOK-oll th* turbine overspeed protection syste ... b. At le 40 months by* disassembling at least one of each

valves and performing a visual and 1urface in.sp* o valve seats, di1k.s and st* .. and verifying no unacceptable flaws or corrosion.

If unacceptable flaws or excessive corrosion are found, all other valves of that type shall be inspected.

4.3.4.4 Verify th* test frequ.ncy maintain.a th* probability of a missile ejection incident within guidelines by reviewing the .. thodology presented in WCAP-11525:

a. At least once every two refueling outage*. b. After modification.a to th* 11.&in turbine or turbine overspeed protection valves . SALEK

UNIT 2 3/4 3-66 Amendment No. 9 7 I

IEACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION.

FOR OPERATION 3.4.1.3 a. b. APPLICABILITY:

ACTION: At least two of the coolant loops listed below shall be OPERABLE:

1. Reactor Coolant Loop 21 and its associated steam generator and reactor coolant pump,* 2. Reactor Coolant Loop 22 and its associated steam generator and reactor coolant pump,* 3. Reactor Coolant Loop 23 and its associated steam generator and reactor coolant pump,* 4. Reactor Coolant Loop 24 and its associated steam generator and reactor coolant pump,* 5. Residual Heat Removal Loop 21, 6. Residual Heat Removal Loop 22. At least one of the above coolant loops shall be in operation.**

MODE 4. a. With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; be in COLD SHUTDOWN within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months . b. With no coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

SALEM -UNIT 2 not be started with one or more of the RCS 3/4 4-3 s may be de-energized that would cause core REACTOR COOL.ANT SYSTEM

3/4. 4. 7 REACTOR COOL.AHT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.7. 1 Th* following Reactor Coolant System leakage detection syste*s shall be OPERABLE:

a. The contain11ent ataosphere particulate radioactivity 1110nitoring sys ta, b. The containment pocket SIJllP level monitoring system, and c. Either the containment fan cooler condensate flow rate or th* ment atmosphere gaseous radioactivity monitoring system. APPLICABILITY:

MODES 1, 2, 3 and 4. ACTION: With only two of the above required leakage detection systells OPERABLE, tion may continue for up to 30 days provided grab samples of the contairment at.osphere are obt.ained and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the required gaseous and/or particulate radioactivity 1110nitoring system fs inoperable; othentise, be in at least HOT STAHDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLO SHUTOO'tm within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS

4. 4. 7. 1 a. b. The leakage detection syst111s shall be de90nstrated OPERABLE by: ContainMnt at.asphere particulate and gaseous (if beina used) lsoor<.Gt Gl-!E:GK 110nitoring syst*s-perforunce of CHANNEL CHECK,i'CHANHEL CALIBRATION C ::_ and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3. Cont.aitw1ent pocket level and cont.ai1111ent fan cooler condensate flow rate (if used) 110nitoring syste11s-perforunc1 of CHAHHEL CALIBRATION at least once per 18 111anth1 .

SALEM -UNIT 2 3/4 4-16

MACTQB COOLANT SXSTIM 3/4.4.9 SPICIPIC ACIIYITX LIMITIRCI CO!Q)ITION FOR OPERATION

- - ----------**********************************************------------

3.4.9 The *pecific activity of th* primary coolant *hall be limited toz a. .s l.O /jCi/qr:rua DOSI: !QOIVALEMT r-131, b. ,S 100/1 /jCi/qr:am.

APPLICABILITXz l, 2, 3, 4 and 5. ACTIONz MOD!S l, 2 and 3* a. With th* *p.cific activity of th* primary coolant > l.O µCi/qr:am DOSS !QUIVALZN'T I-131 for: mor:* than 48 hour:* during on* continuou*

time interval or: exce*ding th* limit lin* *hown on Pic;ur:* 3.4-1, be in at leaat HOT STANJ)IY with T < SOO*P within 6 hour*. a99. b. With th* *pecific activity of the primary coolant > 100/1 µCi/gram, be in at lea*t HOT STANDBY with T < soo*p within 6 hour*. avg MODIS l, 2, 3, 4 and 5 a. With th* *pecific activity of th* P, coolant l.O µCi/qr:am DOSI !QOIVAL!!N'T I-131 or > 100/1 i/;r ... , perform t * *amplinq and analy*i* requirement*

of item 4a of Tahle 4 4 th* specific activity of the primary coolant

within t* limit*. SURVJ:ILLANCZ RJ:QUIREMENTS

-*********************************************

- - - -**-***

4.4.9 The *pecific activity of the primary coolant *hall be determined to b* within the limit* by per:formanc*

of th* aampling and analy*i* proqram of Table 4.4-4. *with T > soo*r. avq SALEH -UNIT 2 3/4 4-23 Amendment No. 112 ...... .-

REACTOR COOLAHT SYSTEM 3.4.11 STRUCTURAL "INTEGRITY ASME CODE CLASS 1, 2 and 3 COMPONENTS LIMIT1HG CONDITION FOR OPERATION 3.4.11.1 Th* structural intagrity of ASME Code Cla 1, 2 and 3 omponents shall b* maintained in accoraaiice with Specificati n

APPLICABILITY:

ALL MODES. 414-11./ ACTION: a.

the structural integrity of any ASME Cod* Class 1 companent(s) not c:onfonting to th* above requirements, restore the structural fnt.grity of affected component(s) to within its limit or isolata tn. affected component(s) prior to increasing the Reactor Coolant System t1111Perature more than S0°F above the minimum temperature required by NOT considerations.

b. With th* structural integrity of any ASME Code Class 2 companent(s) not conforming to the above requirements, restore the structural fntagrity of th* affected component(s) to within its limit or isolate th* affected component(s) prior to increasing the Reactor Coolant Syst.811 tamperature above 200°F. c. With th* structural integrity of any ASME Code Class 3 companent(s) not conforftlfng to the above requirements, restore the structural fntagrity of the affected component(s) to within its limit or isolate th* affected companent(s) fro* service. d. The provisions of Specification 3.0.4 are not SURVEILLANCE REQUIREMENTS 4.4.ll.l In addition to the requirements of Specification 4.0.5, each Reactor Coolant Pump flywheel shall be inspected per the recoanendations of Regulatory Position C.4.b of Regulatory Guide 1. 14, Revision 1, August 1975. 4.4.ll.2 Augmented Inservfce Inspect1on Program for Steam Generator Channel Heads -The Ho. 21 Steam Generator channel head shall be ultrasonically in'SPicted in a selected area during each of the first three refueling outages using the same ultrasonic inspection procedures and equipment used to generate the baseline data. These inservice ultrasonic

nspections shall verify that the cracks observed in the stainless steel cladding to operation have not propagated into the base 111aterial.

SALEM -UNIT 2 3/4 4-33

f.HE&GENCY CORE COOLING SYSTF.HS ECCS SlJASYSTEHS

-Iav1 <3so*F LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a. # One OPERABLE centrifugal charging pump and associated flow path capable of taking suction from the refueling water storage tank and transferring suction to the residual heat removal pump discharge piping and; l. Discharging into each Reactor Coolant System (RCS) cold leg. b. One OPERABLE residual heat removal pump and associated residual heat removal heat exchanger and flow path capable of taking suction from the refueling water storage tank on a safety injection signal and t=3nsferring suction tc the containment sump during the recirculation phase of operation and; l. Discharging into each cold leg, and; upon manual initiation, 2. Discharging into two RCS hot legs. APPLICf.BILity:

MODE 4. ACTION: a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to statu. within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUTDOWN within the next OPERABLE because of the inoperability of removal heat exchanger or residual heat emov , restore at least one ECCS subsystem to OPERABLE status intain the Reactor Coolant Sy*tem T lea* than 350.F by use of alternate heat removal methoda. avg c. In the event the ECCS is actuated and injects water into the Reactor Coolant Sy*tea, a Special Report shall be prepared and submitted to the Coamiasion pursuant to Specification 6.9.2 within 90 days de*cribing the circumstance*

of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70. # A maximum of one safety injection pump or one centrifugal charging pump shall be OPERABLE in MODE 4 when the temperature of one or 110re of the RCS *cold legs is less than or eqU&l to 312*F, Mode 5, or Mode 6 when the head is on the reactor vessel. SALEM -UNIT 2 3/4 5-7 Amendment No. 130

CORE COOLING REF!.( LING WATER STORAGE TANK LŽITING CONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RWST) shall be OPERABLE w1th: a)\J a. A contained volun19 of between and 400,000Aof borated water. b. A boron concentration of between 2,300 and 2,500 ppm, and c. A water tenperature of 35°F. APPLICABILITY:

MO !ES 1, 2, 3 and 4. ACTION: With tht refueling water storage tank inoperable, restore the tank to OPERABLE status w1th1n 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be 1n at least HOT STANCBY within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and 1n COLD SHUTIDWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . S l.RVE I Ll.AHCE REQUIREMENTS 4.5.5 The RWST shall be dlfl!Onstrated OPERABLE:

a. At least once per 7 days by: 1. Ver1fy1ng the water level fn the tank, and 2. Verifying the boron eoncentrat1on of th* w1t1r. b. At least once per 24 hour-1 by ver1fy1ng the RWST tefl!Perature when the outside 11r t_.perature 1s <

SAU:M -UNIT 2 3/4 5-9 Amendment No. SS

3/4.6 CONTAINMENT SYSTEMS 3/4.6,l PRIMARY CONTAINMENT CONTAINMEN'i'""INTEGRITI( LIMITING CONDITION FOR OPERATION 3.6.l.l Primary CONTAINMENT INTEGRITY' shall be maintained.

APPL_ICA8ILITY':

MODES l, 2, and 4 . ACTION: Without primary CONTAINMENT INTEGRITY', restore CONTAINMENT INTEGRITY within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHOTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.6.l.l Primary CONTAINMENT INTEGRITY' shall be demonstrated:

a. At least once per 31 days by verifying that all capable of being closed by OPERABLE containment valves and required to be closed during accide closed by valves, blind flanges, or deactiva ed autcmatic secured in their positions, except for valve th& under Administrative control as permitted by c1 and all equipment hatches are closed and sealed. b. By verifying that each containment air lock is OPBRABLB per Specification 3.6.1.3. c. After each closing of a penetration subject to Type B testing, except containment air locks, if opened following a Type A or B test, by leak rate testing the seal with gas at Pa d verifying that when the measured leakage rate for thee eals is adde to the leakage rates determined pursuant to Specificat"on all other Type B and c penetrations, the canbined l akage rate ss or equal to 0.60 La..

Except vents, drains, test connections, etc. which are (1) one inch nominal pipe diameter or less, (2) located inside the containment, and (3) locked, sealed, *or otherwise secured in the closed position.

These penetrations shall be verified closed at least once per 92 days . SALEM -UNIT 2 3/4 6-1 Amendment No. 1 7 2

N":";..:NMEN7 SYS7EMS CQN1'AINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 a. b. Containment leakage rates shall be limited to: An overall integrated leakage rate of less than or equal to La, 0.10 percent by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at (47.0 A combined leakage rate of less than or equal to 0.60 La for penetrations and valves subject to Type B and c tests 1 1-;e.when pressurized to Pa. APPLICABILITY:

ACTION: With either (a) the measured overall integrated containment leakage rate exceeding 0.75 La, or (b) with the measured combined leakage rate for all penetrations and valves subject to Types B and c tests exceeding 0.60 La, restore the leakage rate(s) to within the limit(s) prior to increasing the Reactor coolant System temperature above 200°F. SURVEILLANCE REQUIREMENTS 4.6.1.2 a. b. c. The containment leakage rates shall be demonstrated as follows: Type A tests shall be in accordance with lOCFR SO.S4(0) in conformance with Appendix J of lOCFR SO, Option B, using the methods and provisions of Regulatory Guide 1.163, September 1995 as modified by approved exemptions.

Type B and C tests shall be conducted*

in conformance with Appendix J of lOCFR SO, Option A, with gas at design pressure (47.0 psig) at intervals no greater than 24 months except for tests involving air locks. Air locks shall be tested and demonstrated OPERABLE in conformance with Appendix J of lOCFR SO, Option A, per Surveillance Requirement 4.6.1.3. SALEM -UNIT 2 3/4 6-2 Amendment No. t6 6, 1 7 2

STRUCTURAL

NTEGRITY

===========================================================

integrity of the containment shall :eve: :onsistent with the acceptance criteria in Specifica

.;??::..::ABE..:TY:

-1CDES :, 2, 3 and 4. ACTI'.:N:

maintained at a 4.6.1.6.f.

he structural integrity of the containment not conforming to :he above :equi:ements, restore the structural integrity to within the limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be i:1 at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHuTDCWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS

===========================================================

4.6.1.6.1 Containment Surfaces The structural integrity of the exposed accessible interior and exterior surfaces of the containment, including the liner plate, shall be determined by a visual inspection of these surfaces.

The inspection shall performed prior to the Type A containment leakage :ate :est (reference Specification 4.6.1.2) to verify no apparent changes in appearance or other abnormal degradation.

If the Type A test is performed a: 10 ;ear intervals, two additional inspections shall be performed at equal intervals during shutdowns between Type A tests. 4. 6.:.. 6. 2 ?.eoorts Any abnormal degradation of the containment structure detected during the above required inspections shall be evaluated for ieportability pursuant to lOCFRS0.72 and lOCFRS0.73.

The evaluation shal: be and shall include a description of the condition of the concrete, the procedure, the tolerances on cracking, and the corrective action :a:-:er. . SALEM -UNIT 2 3/4 6-8 Amendment No.

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION

==================================================================

3.6.2.1 Two independent containment sptay systems shall be OPERABLE with each spray system capable of taking suctio4 from the RWST and transferring suction to the RHR pump discharge.

APPLICABILITY:

MODES 1, 2, 3 and 4. ACTION: With one containment spray system restore the inoperable spray system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within. the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the inoperable spray system to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS

==================================================================

4.6.2.1 Each containment spray system shall be demonstrated OPERABLE:

a. least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

b. By verifying, that on recirculation flow, each pump develops a discharge pressure of greater than or equal to 215 psig when tested pursuant to Specification 4.0.5. c. At least once per 18 months during shutdown, by: l. 2. d.

At actuates High-Hig Verifyi Contai High-High pr in the flow path on a Containment starts automatically on a test signal. 1. .Performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed . SALEM -UNIT 2 3/4 6-10 Amendment No.1 44 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.6.3.2 Each containment isolation valve shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING*MODE at least once per 18 months by: a. Verifying that on a Phase A containment isolation test signal, each Phase A isolation valve actuates to its isolation position.

b. Verifying that on a Phase B containment isolation signal, each Phase B isolation valve actuates to its isolation position.

c. NOT USED d. Verifying that on a Containment Purge and Pressure-Vacuum Relief isolation test signal, each Purge and Pressure-Vacuum Relief valve actuates to its isolation position.

e. Verifying that the Containment Pressure-Vacuum Relief Isolation valves are limited to s 60° opening angle. 4.6.3.3 At least once per 18 months, verify that on a main steam isolation test signal, each main steam isolation valve actuates to its isolation position.

4.6.3.4 The isolation time of each power operated or automatic containment isolation valve shall be determined to be within its limit when tested pursuant to Specification 4.0.5. 4.6.3.5 Each containment purge isolation OPERABLE within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months aftereach closing of ex the valve is being used for multiple cycling east once per hours, by verifying that when the measured le age rate is added to the leakage rates determined pursuant to Specifi all Type B and c penetrations, the combined leak ge rate is less than or equal to 0.60La.

4.6.3.6 A pressure drop test to identify exce resilient valve seals shall be conducted on the: degradation o a. Containment Purge Supply and Exhaust IsoI tion Va least once per 6 months. b. Deleted. SALEM -UNIT 2 3/4 6-15 Amendment No. t 5 3 , 1 7 2

>!AIN STE;..M LINE ISOLATION VAL'/ES :..:MITING CONDITION FOR OPERATION

==================================================================

3. 7. l. 5 Each main steam line isolation valve shall be OPERABLE.

..??!...: CABII...I":"Y:

MODES 1, 2 and 3. ACTION: MODES 1 -With one main steam line isolation valve inoperable, POWER OPERATION may continue provided the inoperable valve is either restored to OPERABLE status or closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; Otherwise, be in MODE 2 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . MODES 2 -With one or more main steam line isolation valve(s) inoperable, and 3 subsequent operation in MODES 2 or 3 may proceed provided;

a. The isolation valve(s) is (are) maintained closed, and b. The isolation valve(s) is (are) verified closed once per 7 days. Otherwise, be in MODE 3, HOT STANDBY, within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and MODE 4, HOT SHUTDOWN, within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . SURVEILLANCE REQUIREMENTS

==================================================================

4.7.1.5 Each main steam line isolati OPERABLE by verifying full closure wi Specification 4.0.5. The provisions of applicable.

demonstrated n 5 seconds w en ested pursuant to 4 . o . are not s;..:..EM

-L'NIT 2 3/4 7-10 Amendment No. 170 ...

.. ANCE 1 Continued)

J. Snubber release rate, where required, is specified range in compression or tension. For snubbers specifically required not to displace under continuous load, the ability of the snubber to withstand load without displacement shall be verified.

f. Snubber Service Life A record of the service life of each snubber, th which the designated service life corT111ences and the :naintenance records on which the designated s ice Life is shall maintained as required by Specificat on Concurrent with the first inservice ins action and at least once per 18 months thereafter, the insta lati maintenance records for each snubber shall be re that the indicated service life has not been exceeded or will not be exceeded prior to the next scheduled snubber service life review. :7 I the indicated service life will be exceeded prior to the next scheduled snubber service life .review, the snubber service life shall be reevaluated or the snubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next scheduled service life review. This reevaluation, replacement or reconditioning shall be indicated in the records. SALEM -lJNIT 2 3/4 7-26 Amendment .68 L PLANT SYSTEMS 3/4.7.10 CHILLED WATER SYSTEM -AUXILIARY BUILDING SUBSYSTEM LIMITING CONDITION FOR OPERATION 3.7.10 The chilled water system loop which services the safety-related loads in the Auxiliary Building shall be OPERABLE with: a. Three OPERABLE chillers b. Two OPERABLE chilled water pumps APPLICABILITY:

ALL MODES and during movement of irradiated fuel assemblies.

ACTION: MODES 1, 2, 3, and 4 a. With one chiller inoperable:

1. heat loads from the 2. Res.tore the chiller atus within 14 days or; 3. Be in at least HOT STANDBY-within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . b. With two chillers inoperable:

1. Remove the appropriate non-essential heat loads from the chilled water system within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and; 2. Align the control room emergency air conditioning system (CREACs) for single filtration operation using the Salem Unit 1 train within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and; 3. Restore at least one chiller tus within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or; 4. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . c. With pump withi hours SALEM -UNIT 2 ter pump inoperable, restore the chilled water atus within 7 days or be in at least HOT STANDBY ours and in COLD SHUTDOWN within the following 30 3/4 7-28 Amendment No. 182 PLANT SYSTEMS LIMITING CONDITION FOR OPERATION ACTION: MODES 5 and 6 or during movement of irradiated fuel assemblies.*

a. With one chiller inoperable:

1. Remove the appropriate non-essential heat loads from the Chilled Water System 4 2. Restore the chiller 14 days or; 3. Suspend CORE ALTERATIONS and movement of irradiated fuel assemblies.

b. With two chillers inoperable:

1. Remove the appropriate non-essential heat loads from the chilled water system within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and; 2. Align the control room emergency air conditioning system (CREACs) for single filtration operation using the Salem Unit 1 3. train within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and; Restore at least one chiller or; atus within* 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 4. Suspend CORE ALTERATIONS and movement of irradiated fuel assemblies.

c. SURVEILLANCE REQUIREMENTS ter pump inoperable, restore the chilled water atus within 7 days or suspend CORE ALTERATIONS and ted fuel assemblies.

4.7.10 The chilled water loop which services the safety-related loads in the Auxiliary Building shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each manual valve in the chilled water system flow path servicing safety related components that is not locked, sealed, or otherwise secured in position, is in its correct position.

b. At least once per 18. months, by verifying that each automatic valve actuates to its correct position on a Safeguards Initiation signal. c. At least once per 92 days by verifying that each chillers starts and runs.

During Modes 5 and 6 and during movement of irradiated fuel assemblies, chilled water components are not considered to be inoperable solely on the basis that the backup emergency power source, diesel generator, is inoperable.

SALEM -UNIT 2 3/4 7-29 Amendment No. 182 ELECTRICAL POWER SYSTEMS 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS LC. DISTRIBUTION

-OPERATING LIMITING CONDITION FOR OPERATION

3. 8.2. l The following A. C. electrical busses shall be OPERABLE and energized from sources of power other than the diesel generators:

4 kvolt Vital Bus I 2.A 4 kvolt Vital Bus I ZB 4 kvolt Vital Bus I ZC 460 volt Vital Bus ' 2.A and associated control centers 460 volt Vital Bus I 28 and associated control centers 460 volt Vital Bus ' zc and associated control centers 230 volt Vital Bus ' 2.A and associated 230 volt Vital Bus ' 28 and associated co 230 volt Vital Bus ' 2C and associated enter 115 volt Vital Instrument Bus ' 2.A and

nverter *" 115 volt Vital Instrument Bus I ZB and verte:-1fC. 115 volt Vital Instrument Bus ' zc and APPLICABILITY:

MODES l, 2, 3 and 4. ACTION: a. With less than the above complement of A.C. busses OPERABLE or energized, restore the inoperable busses to OPERABLE and energized status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOW'N within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . b.

one inverter inoperable, energize the associated A.C. Vital Bus within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; restore the inoperable inverter to OPERABLE and energized status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVE!LLANCE REQUIREMENTS 4.8.2. 1 The specified A.C. busses and inverters shall be determined OPERABLE and energized from A.C. sources other than the diesel generators at least once per 7 days by verifying correct breaker alignment and indicated voltage on the bus se's. *An inverter may be disconnected from its 0.C. source for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for pur?ose of performing an equalizing charge on its associated battery bank proviaec (l) its vital bus is OPERABLE and energized, and (2) the vital busses with the other battery banks are OPERABLE and energized.

SALEM -UNIT 2 3/4 8-S

*

E!.EC7iUCAL rora SYSIF.MS SURVEII..Lt.NC!

R.!QU!R!MINTS (Continued)

2. Th* pilot cell specific gravity, corrected to 77*r, and full electrolyte level, 1s greater than or eqU&l to l.200. 3. The pilot cell voltag* is greater than or eqU&l to 2.08 VO l.t*. The overall battery voltage is greater than or eqU&l to 27 volt1. b. At lea1t once per 92 by verifying that: l. The voltage of each connected cell i1 greater than or equ.l to 2.13 volt1 under float charge and ha1 not decreased more than 0.27 volt1 from the value ob1ervad during the origiral acceptance test. 2. ob1erve than or the value 3. The electro yt* level of each connected call 11 between the mini.llUA and m&ximu.11 level indication 11&rka. c. Ac l*a*c once per 18 month* by verifyin1 thac: l. 2. The cell*, call plat** and battery racka 1hov no vi1U&l indication of phy1ical d&aage or abnormal deterioration.

Th* call*to*cell and t*t'11inal conneccion*

are clean, eight, and coated with anci-corro1ion 11&terial.

3. Th* battery charger will 1upply ac le .. c lSO amperes at 28 volt* for at lease 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . d. Ac lea1c once per 18 monch1, durin1 1hucdovn, by verifying that cha baccery capacity 11 adeqU&ca co 1upply and 11&intain in OPERA!t.!

1cacua all of the actU&l or 11.mulaced

... rgancy load.I for the de1ian ducy cycle when the battery i* 1ubjeccad to a battery orvica t**C. e. AC l***C once per 60 months, during *hucdovn, by verifying thac cha baccary capacity 11 ac lea*c SO* of cha aanufaceurar's racin1 vhan 1ubjecced co a perfot'11Anca discharge ta1c. Sati1factory completion of thi1 perfor11&nce di1char1*

te1c 1hall al10 1atisfy th* requirement*

of Specification 4.8.2.S.2.d if the parfor11&nc*

discharge ce1c 11 conducted during a 1hucdovn vhara that ce1t and the'baccery 1ervice teat would both ba required

SAI..EM

UNIT 2 3/4 8-v.

Mo. 92

REFUELtNG OPERATIONS CRANE TRAVEL

FUEL HANDLING AREA LIMITING CONDITION FOR OPERATION 3 .9 .7 Loads in excess of 2200 pounds shall be prohibited from travel over fuel assembl 1es in the storage pool. APPLICABILITY:

With fuel assembl fes in the storage pool. ACTION: With the requirements of the above spec1fication not satisfied, place the crane load in ii safe condition.

The provisions of Specification 3.0.3 are not appl i cab'-= _ SURVEILLANCE RECl.JIREMENTS 4.9.7 The cutoff which prevent of 2200 pounds over fuel assemblies 7 days prior to crane use and at le crane operation.

ane travel with loJds in excess derronstrated OPERABLE within pe 7 days thereafter during the SALEM -UNIT 2 3/4 9-7 Prnen;jment No. 51

IC.. ? B rOR OPERATION 3 1 3 l At one residual heat removal loop shall be in operation.

" :'?:..: :.l.3::.

6. a.

less :han one residual heac removal loop in operation.

except as provided in b. below, suspend all operations involving an increase in the reactor decay heat load or a reduction ln boron concentration of the Reactor Coolant System. Ciose all contai:".menc providing direc: access fr:im the contairunent atmosphere to the oucside within.:.

hours. b. The residual heat removal loop may be removed from operation for up to l hour per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the perfoniance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs. c. The previsions of Specification 3.0.3 are not applicable.

.:. . 9. 8. l At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months one RHR loop shall be *1erified in operation and circulating coolant at a flow race of: a. greater than or equal to 1000 gpm, and '""' maintain the RCS temperature ac less chan or ! SALE..'i

L:1IT 2 3/4 9-8 Amendment

'lo. 87

OPERATIONS LOW i..ATER LEVEL LIMITING CONDITION FOR OPERATION 3..9.8.2 T;o,o independent Residual Heat Re1ooval (RHR) loops shall be OPEl{ABLE

.* APPLICABILlTY:

MOOE 6 when water level above the .top of the reactor pressure vessel fl an91 is 1 ess than 23 feet. ACTION: a* to b. The provisions of Spec1 f1cat1on SURVEILLANCE REQJlREMENTS 4.9.8.2 The required Rasidual Heat Re111>val OP*RABLE per Specification 4.0.5.

Systems supporting RHR operability may be as follows: a. The normal or emeryency power source may be 1 noperabl e. b. One service water header may be out of service provided the equipment listed in Table 3.4-3 is OPERABLE.

SALEM -UNIT'Z 3/4 9-9 Amendment No. 46

a. b. c. :he LLD is defined 4,ll-2 (Continued)

TABLE NU TAT lON The principal ganma emitters for which the LLD specification applies are the following radionuclide;:

Kr-87, l<r-88, :te-133, .r.e-135, for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, l'o-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions.

This list ooes not rrean that only these are to be detected and reported.

Other peaks that are rreasurable and identifiable, together with t.he above nuclides, shall also be identified and reported.

Sampling and analysis shall also be perforrred following shutdown, startup or a THERMAL POWER change that, within one hour, exceeds 15 percent of RATED THERMAL POWER unless: 1. Analysis shows that the DOSE EQJIVALENT I-131 concentration in the primary cool ant has not increased rTCre than a factor of three. 2. The noble gas activity rTCnitor shows that effluent activity has not increased by rmre than a factor of three. d. Tritium grab shall be taken at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the refueling canal is flooded. e. Tritium grab shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel pool area whenever spent fuel is in the spent fuel pool. f. The ratio of the flow rate to the sampled stream flow rate shall be known for the time period covered by each oose or cilse rate calculation made in accordance with Specif1cat1ons 3.11.2.l, 3.11.2.2 and 3.11.2.3.

SALEM -UNIT 2 3/4 11-10 Amendn'ent No. 36 r--RQCTOll CQOLAIT sxsrg u.su

.. **anLJ*acccmccL 1;*c:t:.;c;a:c..i1.L naaa 3/4.4.9 lrlCIFIC AC'l'IVITX Th* limitation*

on th* *pacific activity of th* pri.m&.ry coolant *n*ure that th* r**ultin9 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months do*** at the *ito boundary vill not exceed an appropriately mmall fraction of Part 100 limite tollowin9 a *t*aa 9enerator tube rupture accident in conjunction vitb an ***Wied *teady *tat* primary-to-*econdary

t*aa generator leaka9* rat* of 1.0 QPK. Th* value* for the limit* on *pecitic activity repr***nt iuteria limit* baaed upon a parametric evaluation by th* of typical *it* locatiorw.

Th*** valu** are con**rvativ*

in that *pecific *it* parameter*

of the .. *it*, *uch a* *it* boundary location and meteoroloqical condition*, were not con*idered in thi* evaluation.

Th* HRC: i* tinalidn9 dte *pecitic criteria vhich vill be u*ed a* th* ba*i* tor th* r .. valuation of th* *pecitic activity limit* of thi* *ite. Thi* r .. valuation may re.ult in higher limit*. Reducing T av9 *t*am generator tube ru coolant i* below the vent* th* r*l*a** of activity *hould a th* *aturation pr***ur* of th* primary th* atmo*pheric

t*aa relief valv*** Th* *urveillanc*

adequate ***urance that exce**iv*

pecif ic activity level* in th* primary coolant vill be detected in *uff icient time to take correcti.e action. Information obtained on iodine *piking will be u*ed to a***** th* parameter*

- - ociated vith *pikin9 phenomena.

A reduction in frequency of i*otopic analy*** followin9 power chan9** may be permi**il>l*

if ju*tif ied by th* data obtained.

SALO -UNIT 2 B 3/4 4-6 Amendment No. 112 1-, * *

3ASES F::1a::;. :1ew ru:e wh:=h addre,ses the metal temperature of the =:osure flange regions is considered.

Tt1s lOCFRSO rule states that the -:-.etal of the clos'clre flange regions must exceed the material RTNDT by at

c'F normal operat:or the pressure exceeds percent of hydrostat1=

test pressure (621 psis :or Salem). Table 33/4.4-1 indicates that the limiting RTNDT of 28°F in the closure head flange of Salem Unit 1, and the minimum allowable temperature of this region :s l48°F at pressures greater than 621 These limits do not affect Figures 3.4-2 and 3.4-3. Although the pressurizer operates in temperature ranges above those for which there is.reason for concern of non-ductile failure, operating limits are provided to assure compatibility of operation with the fatigue anG:ysis performed in accordance with the ASME'Code requirements.

or an RCS opening of than 3.14 hat the RCS will be protected from transients imits of Mppendix G 10 CFR Part 50 when one or more less than or equal to 312°F. Either POPS has adequate ility to prote.:.t the RCS from

,;hen the transi limite to either (1) the start of an idle RCP with the secondary

ture of the steam generator less than or equal to 50°F above the RCS cold leg temperatures, or (21 the start of an Intermediate Head Safety Injection pump pump and ltS injection into a water solid RCS, or the start of a High Head Safety Injection pump in conjunction with running Positive Displacement pump and its injection into a water solid RCS . SALEM -UNIT 2 B 3/4 4-12 December 22, 1994 .... .*

,-J/4.10 SPECIAL T!ST E:XCEPTIONS 3/4.10.l SHUTDOWN MARGIN This 1pecial test exception provides that a m1nU11WD amount of control rod worth is 1DID9diately available for reactivity con.trol when tests are performed for control rod worth measurement.

This special test exception is required to permit the periodic verification of the actual versus predicted core reactivity condition occurring as a result of fuel burnup or fuel cycling operations.

3/4.10.2 GROUP HEIGHT, ISSERTION, AND POWER DISTRIBtrrION LIMITS This special test exception permits individual control rods to be positioned outside of their normal group heights and insertion limits during the performance of such PHYSICS TESTS as those required to l) measure control rod worth, and 2) determine the reactor stability index and damping factor under xenon oscillation conditions.

3/4.10.3 PHYSICS TESTS This special test exception permits PHYSICS TESTS to be performed at less than or equal to 5% of RATED THERMAL POWER with the Reactor Coolant System T slightly lower than normally allowed so that the fundamental nuclear cRX,acteristics of the reactor core and related instrumentation can be verified.

In order for various characteristics to be accurately it is, at tirues, necessary to operate outside th* normal restrictions of these Technical Specifications.

For instance, to measure the moderator temperature coefficient at BOL, it is necessary to position the various control rods at heights which may not be allowed by Specification which ma

in turn, cause the RCS T to fall slightly below the minimum temperature of Specification 3/4.10.4 NO FLOW TESTS This special test exception permits reactor criticality under no flow conditions and is required to perform certain startup and PHYSICS TESTS while at low THXRMAL POWER levels. SAL.EM -UNIT 2 B 3/4 10-l Amendment No, 66 DESIGN FEATURES ==============================================================================

DESIGN PRESSURE AND TEMPERATURE 5.2.2 The r3actor containment is designed and shall be maintained for a maximum internal pressure of 47 psig. Containment air temperatures up to 351.3°F are acceptable providing the containment is in accordance with that described in the UFSAR. 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 193 fuel assemblies.

Each assembly shall consist of a matrix of zircaloy or ZIRLO clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide as fuel material.

Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting ccre regions. CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 53 full length and no part length control rod assemblies.

The full length control rod assemblies shall contain a nominal 142 inches of absorber material.

The nominal values of absorber be 80 percent silver, 15 percent indium : percent cadmium. All control rods shall be clad with stainless steel tubing. PERATURE coolant system is designed and shall be maintained:

a. In accordance with the code requirement specified in Section 4. 1 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b. For a pressure of 2485 psig, and c. For a temperature of 650°F, except for the pressurizer which is 680°F. VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 12,811 +/- 100 cubic feet at a nominal T,vq of 581.0°F. SALEM -UNIT 2 5-4 Amendment No.

ML18102B675

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