ML19332C610
| ML19332C610 | |
| Person / Time | |
|---|---|
| Site: | 05000605 |
| Issue date: | 11/17/1989 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19332C606 | List: |
| References | |
| NUDOCS 8911280344 | |
| Download: ML19332C610 (927) | |
Text
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23A6100AP Sinndard Plant Rev.A CHAPTER 16 7 O TABLE OF CONTENTS Section 11111 East i 16 TECHNICAL SPECIFICATIONS -16.1 1 16.1 CON'IEN'il USE AND APPLICATIONS 16.1 2 16.1.1 Table of Contents 16.12 16.1.2 1.0 Use and Application 16.1 9 . 16.2 2.0 SAFEW LIMITS 16.2 1 163 3& APPLICAll!LITY 163 1 16.4 3.1 REACTIVIW CONTROL SYSTEMS 16.4 1 16.4.1 3.1.1 Shutdown Matgin 16.4 1 16.4.2 3.1.2 Control Rod Operability 16.4-5 16.4 3 3.13 Control Rod Seram Times 16.4 10 V 16 4.4 3.1.4 Control Rod Scram Accumulators 16.4 10 16.4.5 3.1.5 Cuntrol Rod Drive Coupling 16.4 16 16.4.6 3.1.6 Rod Pattern Control 16.4 18 16.4.7 3.1.7 Standby Liquid Control System 16.4 20 16.5 3.2 POWER DISTRIItUTION LIMITS 16.5 1 16.5.1 3.2.1 Average Planar Linear Heat Generation Rate 16.5-1 16.5.2 3.2.2 Minimum Critical Power Ratio 16.5-2 16.5 3 3.23 Linear Heat Generation Rate 16.5-3 16.6 3.3 INSTRUMENTATION 16.6-1 16.7 3.4 REACTOR COOLANT SYSTEM 16.7 1 16.7.1 3.4.1 Recirculation Pumps Operating 16.7 1 16.7.2 3.4.2 Safety / Relief Valves 16.7-4 fr v 16-il Amendment 9
ABWR DA6100AP Standard Plant Rev.A ,
- CHAPTER 16 i Section TABLE OF CONTENTS (Continued) hf 31 tic East l 16.73 3.43 Operationalleakage 16.7-6 l
16.7.4 3.4.4 Specific Activity 16.7-8 1 16.7.5 3.4.5 Residual Heat Removal Shutdown 16.7 10 16.7.6 3.4.6 Reactor Coolant System Pressure / Temperature Limits 16.7 12 16.7.7 3.4.7 Reactor Steam Dome Pressure 16.7 15 16.8 3.5 EMERGENCY CORE COOLING SYSTEM 16.8-1
- 16.8.1 3.5.1 ECCS Operation 16.8-1 16.8.2 3.5.2 ECCS Shutdown 16.8 7 16.9 3.6 CONTAINMENT SYSTEMS 16.9 1 ,
16.9.1 3.6.1 Primary Containment Systems 16.9 1 16.9.2 3.6.2 Suppression Pool ~ 16.9-19 l 16.93 3.63 Hydrogen Control 16.9-29 16.9.4 3.6.4 Secondary Containment Systems 16.9-31 16.10 3.7 PLANT SYSTEMS 16.10-1
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( 16.10.1 3.7.1 Reactor Building Cooling Water (RCW)/ Reactor Building Senice Water (RSW) System Operating 16.10-1 l 16.10.2 3.7.2 Reactor Building Cooling Water (RCW)/ Reactor Building Senice Water (RSW) System Shutdown 16.10-5 16.103 3.73 Control Room HVAC Emergency Recirculation System 16.10-8 , 16.10.4 3.7.4 Main Condenser Offgas 16.10-14 16.11 3.8 ELECTRICAL POWER SYSTEMS 16.11-1 16.11.1 3.8.1 A.C. Sources-Operating 16.11-1 16.11.2 3.8.2 A.C. Sources-Shutdown 16.11 20 16 ...m Amendment 9
MM . 23A6100AP Standard Plant unu CHAPTER 16 l ( ,]T TABLE OF CONTENTS (Continued) . b Section 31tle East . 16.113 3.83 Diesel Fuel Oil 16.11 22 [ 16.11 26 i 16.11.4 3.8.4 D.C. Sources-Operatiry, i
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16.11.5 3.8.5 D.C. Sources Shutdown 16.11 31 16.11.6 3.8.6 Battcry Electrolytc 16.11 33 16.11.7 3.8.7 A.C. Power Distribution Systems-Operating 16.11 37 16.11.8 3.8.8 D.C. Power Distribution Systems-Operating 16.11 40 16.11.9 3.8.9 A.C. and D.C. Power Distribution Systems-Shutdown 16.11 42 16.12 3.9 REFITELING OPERATIONS 16.12-1 16.12.1 3.9.1 Refueling Equipmont Interlocks 16.12 1 16.12.2 3.9.2 Refuel Position One Rod-Out Interlock 16.12 2 + 16.123 3.93 Control Rod Position 16.12-4 16.12.4 3.9.4 Control Rod Position Indication 16.12 5 16.12.5 3.9.5 Control Rod OPERABILIT% Refueling 16.12 7 16.12.6 3.9.6 Water level-Reactor Pressure Vessel 16.12 8 16.12.7 3.9.7 Water Level Spent Fuel Storage Pools 16.12 9 16.12.8 3.9.8 ResidualIIcat Removal liigh Water Level 16.12 10 16.12.9 3.9.9 Residual 11 cat Removal-Low Water Level 16.12 12 16.13 3.10 SPECI AL OPERATIONS 16.13 1 16.13.1 3.10.1 Insenice leak and 11ydrostatic Testing Operation 16.13-1 16.13.2 3.10.2 Reactor Mode Switch Interlock Testing 16.13 3 16.12 3 3.103 Control Rod Withdrawal liot Shutdown 16.13-5 16-iv l Amendment 9 i L
23A6100AP Standard Plant an. A i CHAPTER 16 , TABLE OF CONTENTS (Continued) gl Section 31tle Eage 16.12.4 3.10.4 Control Rod Withdrawal-Cold Shutdowc 16.13-8 l 16.12.5 3.10.5 Control Rod Removal-Refueling 16.13 11 ! 16.12.6 3.10.6 Multiple Control Rod Withdrawal Refueling 16.13 14 16.12.7 3.10.7 Control Rod Testing-Operating 16.13 16 i f i i O 16 v Amendment 9 O l
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M 23A6100AP Standard Plant a,m x l SECTION 16.1
/ l CONTENTS .QJ Section Ihlt East 16.1.1 Table of Contents 16.1 2 16.1.2 1.0 Une and Application 16.1 9 l 16.1.2.1 1.1 Definitions 16.1 9 r 16.1.2.2 1.2 Logical Connector 6 16.1 16 ;
16.1.23 13 Completion Times 16.1 18 16.1.2.4 1.4 Frequency 16.1 22 16.1.2.5 1.5 Legal Considerations 16.1 25 i t 6 ( l r N. 16.1-1i Amendment 9 l-
1 l I i 1 ABWR mumi- ! Standard Plant anu ! 16 TECHNICAL interim policy teatement sets out specific criteria for SPECIFICATIONS the content of technical specifications. The interim (a)- r Technical Specifications are explicit restrictions policy statemeut specifically recognizes that: on the operation of a commercial nuclear power *The purpose of Technical Specifications is to plant. They are designed to preserve the validity of impose conditions or limitations upon reactor i the plant safety analysis by ensuring that the plant is operation necessary to obviate the possibility of l operated within the required conditions bounded by an abnormal situation or event gising rise to an analysis, and with the operable equipment that is immediate threat to the public health and safety assumed to mitigate the consequences of an accident, by establishing those conditions of operation Technical specifications preserve the primary which cannot be changed without prior success path relied upon to detect and respond to Commission approval and by identifying those accidents. They also complement the concept of features which are of controlling importance to defense in depth, safety? Section 182a of the Atomic Energy Act of 1954, This set of proposed Technical Specifications as amended (the Act),47 U.S.C. 2011, at 2232, establish these conditions and limitations for the ; provides the legislative framework within which ABWR. This set of technical specifications is technical specifications are required Section 182a intended to be used as a guide in the development of of the Act requires in pcrtinent part; plant specific sets of technical specifications for nlants whose license annlientions reference ths211E3
*In connection witi; applications for licenses to standard clant.
operate production or utilization facilities, the applicant shall state such technical specifications, including information on the amount, kind, and source of special nuclear material required, the place of use, the specific h V characteristics of the facility, and such other information as the Commission may, by rule or regulation, deem necessary in order to enable it to find that the utilization or production of special nuclear material will... provide adequate protection to the health and safety of the public. Such technical specifications shall be part of anylicense issued? The regulatory framework implementing Section 182a of the Act is the NRC's regulation Title 10 Code of Federal Regulations Part 50, Section 50.36,' Technical Specifications". This regulation provides in part that each operating license:
*will include technical specifications...(to) be derived from the analysis and evaluation included in the safety analysis report, and amendments thereto...and may also include such additional technical specifications as the Commission finds appropriate. '
The Commission has issued an interim
' Proposed Policy Statement on Technical Specification Improvernents for Nuclear Power c Reactors,' 52 FR 3788, February 6,1987. This Amendment 9 16.11
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2u61our Reasularel Plant Rev.A , 16.1 CONTT.NTE,USE AND APPLICATION f i t (v"~~') 36.6.1 Tele of Centents i l- t t l' l' l TetLE OF CONTENT $ ,
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2LlaM............................... i ! Y2 6 It ' DF CONTENTS . . . . . . . . . . . . . . . . . . . . . . . . . . 11 L .,1 0F TABLt$ ........................... v1 LIST OF FIGURES . . . . . . . . . . . . . . . . . . . . . . . . . . . vil > LIST OF [FFICTIVE PAGI$ . . . . . . . . . . . . . . . . . . . . . . . v111 , v
- 1. U$( AND APPLICATIDN .
1.1 DEFINITIONS .......................... 11 ! ACTIDN$ . . . . . . . . . . . . . . . . . . . . . . . . . . 11 AVERAGE PLANAR LINEAR HEAT GtNtRAT]DN RATE ........ 11 ' COR[ AL1tRAT10N . . . . . . . . . . . . . . . . . . . . . . 11 [COREOPERATINGLIMITSREPORT) .............. 12 i m $t (0UIVALENT 3 13) . . . . . . . . . . . . . . . . . . . 12 LEAKAGE . . . . . . . . . . . . . . . . . . . . . . . . . . 13 LINEAR HEAT GENERATION RATE . . . . . . . . . . . . . . . . 13 HINIMUM CRITICAL POWER RA110 ............... 14 MODE ........................... 14 : OPERABLE . OPERABILITY .........,........ 14 PHY$1C$ TESTS . . . . . . . . . . . . . . . . . . . . . . . 16 RAT ED THE RMAL POWER . . . . . . . . . . . . . . . . . . . . 16 LMUTbOWN MARGIN . . . . . . . . . . . . . . . . . . . . . . 16 *
$1AGGERED it$1 8A$15 ................... 17 THE RMAL POWt R . . . . . . . . . . . . . . . . . . . . . . . 17 ;
3.2 LDGICAL CONNECIDR$ . . . . . . . . . . . . . . . . . . . . . . . 18 1.3 COMPL ET I DN T IME S . . . . . . . . . . . . . . . . . . . . . . . . 1 10 1.4 FREQUENCY . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 14 1.5 LEGAL CDN$1 DERAT 10NS . . . . . . . . . . . . . . . . . . . . . 1 17 I ADWR 11 $/31/89 Aamedemat 9 1&l 2 l a
-, r 31A610 HAP Send =rd Plant nev ^ !
l 9:! r l TABLE OF CONTENTS l (continued) EARt
- 2. $AF[TY LIMJ15 2.1 Safety Limits . . . . . . . . . . . . . . . . . . . . . . 21 2j infety tirit Violation ................. 21 ,
LIMITING CONDITIONS FOR OttMTION AND $URVilLLANCE REQUIREMENTS ' 3. 3.0 APPLICABILITY Limiting Conditions for Operation .............. 31 Surveillance Requirements .................. 34 ; 3.1 REACTIVITY CONTROL SYST(MS LJ 3HUTDollN MARCIN . . . . . . . . . . . . . . . . . . 3.1 1 ;
'J '
- entrot Rod OPERABillTY . . . . . . . . . . . . . . 3.1 5 ;
J i ent ro' Rod irram Times . . . . . . . . . . . . . . 3.1 10 , L 1:entro' tod Scram Accumulators . . . . . . . . . . . 3.1 14 L :entro'l lod Drive C ouclino . . . . . . . . . . .. . . 3.1 16 W 3.1.7 tod Pattern control . . . . . . . . . . . . . . . . Standby Liquid Control System . . . . . . . . . . . 3.1 18 3.1 20 3.2 POWtR DISTRIBUTION LIMIT $ 3.2.1 AVERAGE PLANAR LIN[AR HEAT GEN [ RATION RATE . . . . . 3.2 1 3.2.2 MINIMVM CRIT] CAL POW [R RATIO . . . . . . . . . . . 3.2 2 3.2.3 LINEAR HEAT GENERATION RATE , . . . . . . . . . . 3.2 3 3.3 INSTRUMINTATION I [Later) 3.4 REACTOR COOLANT SYSTEM W Retirculation Punes Doeratino . . . . . . . . . . . 3.4 1 W 3.4.3 Safety /Reitef Valves . . . . . . . . . . . . . . . . Operational Leakage . . . , . . . . . . . . . . . . 3.4 4 3.4 6 l 3.4.4 Specific Activity . . . . . . . . . . . . . . . . . 3.4 8 l W ?esidual Heat Removal - Shutdown . . . . . . . . . . 3.4 10 W 3.4.7 seactor Ceolant System Pressure /Temocrature limits . Reactor 5 team D9me Presrure . . . . . . . . . . . . 3.4 12 3.4 15 3.5 (CCS L}j [CCS - Ooeratina ....... .. ....... 3.5 1 3.6.2 (CC$ . $hutdown . . . . . . . . . . . ....... 3.6 7 ABWR 1i1 5/31/89 O Amendment 9 16.1 3
r I hkk 3146100AP [
.ta="=rd Plant Rev ^
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^
TA4LE OF CONitNTS (continued) I Itat
- 3. LIMITING CONDITIONS FOR OPERAT10N AND $DbflLLANCE Rf 0VlREMENTS *
(continued) 3.6 CONTA!NM[N1 $Y$1[M5 3.6.1 Primary Containmert $ystems 3.6.1.1 Primary Containment . . . . . . . . . . . . 3.61 3.6.1.2 Containment Air Locks . . . . . . . . . . . 3.6 3 3.6.1.3 Containment Pressure ........... 3.6 7 ! 3.6.1.4 Containment Average Air Temperature . . . . 3.6 8 ' 3.6.1.6 PrimaryContainment[andPressure) ! Isolation Yalves ............. 3.6 9 3.6.1.6 Wetwell to Drywell Vacuum Breakers .... 3.6 16 3.6.2 Suppression Pool
-[( 4.6.f.1 3.6.2.2 Egeression Pool Avernoe Tameerature Suppression Pool Water Level .......
... 3.6 19 3.6 22 4 3.6.f.3 ?etidual Heat Removal tuoorettien Pool
.coling . . . . . . . . . . . . . . . . . . 3.6 24 3.6.f.a ketidual Heat _ Removal Wetwe11 Sorav . . . . 3.6 27 3.6.3 Hydrogen Control 3.6.3.1 Containment Hydrogen Recombiner Systems . . 3.6 29 3.6.3.2 Primary Centainment Oxygen Concentration .............. 3.6 30 3.6.4 Secondary Containment Systems i 3.6.4.1 secondary Containment . . . . . . . . . . . 3.6 31 3.6.4.2 secondary Containment Isolation Valves .. 3.6 34 3.6.a.3 Standby Gas Treatment $vitem ....... 3.6 37 3.7 PLANT SYSTEMS W Reactor Buildino Cooline Water /Peactor Buildino service Wat:r system Operatino .......... 3.7 1 W Reactor Buildino foolino Water /Reatter Buildino krvice Water System - Shut down . . . . . . . . . . . 3.7 3 3.7.3 Control Room HVAC Emergency Recirculation S 3.7 8 3.7.4 Main Condenser Of fgas . . . . . . . . . . .ystem ..... . . 3.7 14 I
i 1-ABWR iv 5/31/89
' [
Amendment 9 16.1-4 l
MM 3M610MP ! Standard Plant Rev A I 9: i f TABLE OF COWitW15 (continued) e IASt
- 3. LIMITING CONDITIONS FOR D&f kATION AND SURVfill ANCE PfDUIREMLMI}
(continued) 3.8 ILtCTRICAL POW [R SY$1tMS W 3.8.2 A.C. Sources A.C. Sources Ober.AMag ....... .......
$hutdown ....... .......
3.8 1 3.8 20 3.8.3 Diesel fuel Oil . . . . . . . . . . . . . . . . . . . 3.8 22 [LL J.ources - Deeratino ...... ........ 3.8 26 D C . Sources . $buidown .............. 3.8 31
.8.6 Battery tiectrolyte ................ 3.8 33 A.C. Sower )istribution Systems Doeratina . . . . . 3.8 37 D.C. > owe r )istribution $vsters Oceratino . . . . . 3.8 40 A.C. and D.C Power Distribution $vstems Shutdown . 3.8 42 7
3.9 REFUELING OP[RATIDNS < 3.9.1 Refueling [quipment Interlocks . . . . . . . . . . . 3.9 1 3.W.2 Refuel Position One. Rod.Out Interlock .......3.92 3.9.3 Control Rod Position . . . . . . . . . . . . . . . 3.94 3.9.4 Control Rod Position Indication . . . . . . . . . . 3.9 5 W 3.9.6 Control Rod OPERf.BillTY - Refuelino . . . . . . . . Water Level Reactor Pressure vessel , . . . . . . 3.9 7 3.9 8 3.9.7 Water Level Spent Fuel Storage Pools . . . . . . . 3.9 9 3.9.8 Residual Heat Removal + High Water Level . . . . . . 3.910 3.9.9 Residual Heat Removal Low Water Level . . . . . . 3.9 12 3.10 $F[CIAL OPERATIONS 3.10.1 Inservice leak and Hvdrestatic Testino Doeration . . 3.10 1 3.10.2 Reactor Made Switch Interlock lesting . . . . . . . 3.10 3 3.10.3 Contro' Rod Withdrawal - Hot $hetdown , , . . . . . 3.10 $
}& Contro' tod Withdrawal . co ld Shutdown . . . . , . 3.10 8 3.10.5 Contro' Tod Drive Removal - Refuelina . . . . . . . 3.10 11 3.10.6 Multip'e Control Rod Withdrawa) Refueling . . . . 3.10 14 W Control Rod Testino Doeratino . . . . . . . . . . 3.10 16 ,
ABWR v 5/31/89 Amendment 9 16.1 5
3.%61DQAP
.StanAnrel Plant u.i /
kw Lllt 0F TABLt$ Table Wo- 11112 fARR I.1 1 MODI $ . . . . . . . . . . . . . . . . . . . . . . . . l=$ 3.1 13 3.1.3.1 1 Control RDd $ Cram Times .......... . . . 3.8.1 1 Diesel Generator 1951 Schedult . . . . . . . . . . . 3.8 19 3.8.6 1 Battery $urveillance Requirements . . . . . . . . . 3.8 36 ( ABWR vi $/31/E9 Amendment 9 16.1 6
l i MN 2M6100AP Rea " =rni Pimni - u.A O l Ll&T OF F14URtl 1 l i Floure No. M g. l 3.1.8 1 Sodium Pentaborate Solution J Temperature Versus Concentration Requirements . 3.1 23 l 1 1 I I 1 I I I eil 1 l l l l i i I l , 1 ABWR vii 5/31/89 O Amendment 9 16,1 7 l l 1 .. l l .._ _ _ _ _ . _ , - - . . _ _ _ -
. . . . _ . . _. . . _ . . _ _ _ _ . _ . _ _ _ . _ . . _ _ . _ _ . . _ _ ~.___ . _ . . . .
f i M 21A6100AP ,
.Samadard Plant Rev A l
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Ll5T OF EFFECTIVE PA&t$ l ((1 AMPLE) tage Revition Egig 1 . . 1i . . 1ii . . iy . . v . . vi . . i Eggg Anendment kgig e
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32 . - 13 . . 14 - . O 15 16 l l l 1 I l ABWR vitt 5/31/89 Amendment 9 16.1 4
MM 21A6100AP Remmalmed Plant m,v. A 36.1 2 1A Use and Appheeses nEtJ.: 1.1 naamasema g l l i Definitions I 1.1 ; 1.1 DUINITIONS l
...................................N0Tt.......................................
The defined tems of this section appear in capitalized type and are applicable throughout these Technical $pecifications and their Bases. - Igg befinition ACTIONS ACTION $ shall be that part of a Technical Specification which prescribes Required Actions to be taken under designated Conditions within specified Completion Times.
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AV[ RAGE PLANAR LINtAR The AVIRAGE PLANAR LINEAR HEAT HEAT GEN (RAll0N RAl[ GENERATION RATE (APLHGR) shall be applicable to a specific planar height and is equal to the sum of the heat ' seneration rate per unit length of fuel rod (LINEAR HEAT GENERATION RAT [) for all the fuel rods in the specified " bundle at the specified height divided by the number of fuel rods in the fuel bundle at that height. CORE AlitRATION A CORE ALTERATION shall be the movement or sanipulation of any fuel, , sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. Movement of SRNMs. LPRMs. TIPS or special movable detec. tors (including replacement) is not considered a CORE ALTERATION. $vspen. ston of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe, conservative position. I
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(continued) ABWR 11 $/31/89 A=*ad=*at 9 16.1 9
MM 23A6100AP hadard Plant hvA i i Definitions 1.1 1.1 DIFINIT10NS febetinued) Igm befinition [COREOPERATINGLIMITSREPORT) The CORE OPERATING LIMITS REPORT (COL ) is a rsload cycle specific document. Its supplements and revi. stons, that provides core operating limits for the current operating reload cycle. These cycle specific core operating limits shall be deter-eined for each ante with reload cy[cle Specification 5.x).inPlant accord-operation within these operating ' Itaits is addressed in indiviewel specifications. DOSE EDUlVAL[NT 1 131 005E EQUIVALENT 1131 shall be that concentration of 1 131 (microcuries/ gram) which alone would produce the same thyroid dose as the quantity and isotopic airture of 1 131 I.132, 1 133, 1 134 and 1 135 actually present. The thyroid dose conversion O' factors used for this calculation shall be those listed in Table !!! of T10 14844. ' Calculation of Olstance Factors for Power and Test Reactor Sites'. (continued) ABWR l2 5/31/89 Amendment 9 16.1 10
. _ - . . . . - . _ . - ~ . . . .
i MM z w io m , Standard Plant Rev.A i O' l Definitions 3.1 1.1 DUINIT10NS frontinued) 5 1RIE DefirMtion i LEAKAGE LEAKAGE shall be: I A. Identified Leakage: !
- 3. Leakage into collection systems, such as pump seal or valve packing leaks. that is captured and conducted to a sump or collecting tank, or
- 2. Leakage into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the
, operation of leakage t detection systems, or not to be Pressure Boundary Leakage. B. Pressure Boundary Leakage: i Leakage through a non isolable fault in a Ranctor Coolant System component body, pipe wall, or vessel wall. C. Unidentifisd Leakage: All leakage which is not Identified Leakage. D. Total Leakage: Sum of the Identified and 4 Unidentified Leakage. LIN[AR HEAT GENERATION RATE The LIN[AR HEAT G(NERATION RATE (LHGR) shall be the heat generation per unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length. (continued) ABWR 33 5/31/89 Amendment 9 16.111
MM geandard Plant 2M6100AP w^ l' ( . Definitions 1.1 1.1 DEFINITIONS (continued) Igg Definition MINIMUM CRITICAL POWER RATIO The MINIMJM CRll' CAL POWER RATIO (MCPR) shall t/tt:w smallest Critical which exists in the Power core. TheRatie SPR $ICfi.$'411 be the ratto of that power in the assembly which is calculated by appittation of the appropriate correlation (s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power. MODE A MODE shall correspond to any one inclusive combinetton of mode switch position and everage reactor coolant temperature specified in Table 1.1 1 with fuel in the reactor vessel. , OPERABLE - OPERABILITY A system, subsystem, train, component or device shall be OPERABL(, or have OPERABILITY, when it is capable of performing its specified function (s), O end when all necessary attendant instrumentation, controls, electrical power sources, cooling or seal water, lubrication or other aux 111ery equipment that are required for the system, subsystem, train, component er device to perform its function (s) are also capable of performing their related support function (s). (continued) ABWR l4 $/31/89 O Amendment 9 16.1. 2
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MN 21A6100AP q RinnAmsvi Plant new A ',',.i Definitions 3.1 , 1.1 DEFINITIONS frontinued! Table 1.1 1 HQDil REACTOR MODE SWITCH AVERAGE REACTOR MODE I ') TITLE COOLANT TEMPERATURE, 'T POSITION 1 Power Operation Run Any temperature 2 Startup Startup/ Hot Standby Any temperature 3 Hot Shutdown shutdown > 200 'T 4 . Cold shutdown Shutdown s 200 'I
-5 Refueling Shutdown or Refuel Any temperature
-(a) In MODES 1 through 4 fuel is in the reactor vessel with the reactor -
vessel head closure bolts fully tensioned. In MODE 5 fuel is in the reactor vessel with the reactor vessel head closure bolts less than fully tensioned or with the head removed. l-l' (continued) 1 ABWR l5 $/31/89 O Amendment 9 16.1-13
( A
-AB M 23raoose
. Standard Plant - nv^
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- /] I N)
Definitions 1.1 1.1 DEFINITIONS fcontinued) 31tg ' Definition PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation, and which are: A. described in Chapter 14 of the
$$AR.
B. authorized under the provisions of 10 CFR 50.59, or C. otherwise approved by the Commission. RATED THERMAL POWER RATED THERMAL POWER (RTP) shall be a total reactor core heat transfer rate to the reactor coolant of 3926 Mwt.
$HUTDOWN MARGIN SHUTDOWN MARGIN shall be the amount of reactivity by whuh the reactor is i subcritical or wnuld be subcritical L assuming all control rods are fully I
inserted except for the control rod m it of highest reactivity worth which is assumed to be fully withdrawn and the reactor is zenon free and the moderator temperature is 68'F. With a contrcl rod not capable of being fully inserted, the reactivity worth of this control rod must also be accounted for in the determination of. SHUTDOWN MARGIN. I. i Note? A control rod Dair consists of l two control rods which are connected to the same. shared
- Hydraulic Control Unit (HCU) scram accumulator. All control
(- rods share an accumulator with l -- l' one other rod exceot for the center control rod whitGas M own accumulator. (continued) ABWR l-6 5/31/89 O U i Amendment 9 16.114
t 23A6100AP : Sensadarel Planit a,v. A O' Definitions 3.1 , 1.1 DEFINITIONS feontinued) h Definition STAGGERED TEST BASIS A STAGGERED TEST BAS 15 shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the specified surveillance frequency such that all systems, subsystems, channels or other designated components are tested during n surveillance frequency intervals, where n is the total number of systems, subsystems, channels or other designated components in the associated Function. THERMAL POWER THElbiAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. O i ABWR l7 5/31/89 O I Amendment 9 16.115
23A6100AP
' Reasulard P! ant n, A 16.1J.2 1.2 Eadeel Connectare p
Logical Connectors 1.2 1.2 Loc 1 tat cosettioRs PURPOSE Logical connectors are used in Technical Specifications to i discriminate between, and yet connect, discrete Conditions. Required Actions Completion Times, Surveillances and
* ' Frequencies. The only logical connectors which appear in technical specifications are M and QB. The physical arrangement of these connectors constitute logical conventions with specific meaning.
The intent of this section is to provide specific examples of logical connectors and explain the intended meaning. 3 [XAMPL[$ Example 1 demonstrates that for Condition A, both Required Actions must be completed. In this case, the logical connector M is left justified and shows that both Required Actions A.1 and A.2 are required to be performed. [xample 1. ( l CONDITION REQUIRED ACTION l A A.) l E v A.2 l l (continued) A8WR 18 5/31/89 O V Amendment 9 16.1 16
-4 MQ 23A6 Standapf Plant O
Logical Connectors 1.2 , 1.2 t0G1 tat CONNICTORS fenntinued) [KAMPLES In [mample 2. Required Action A.) Required Action A.! and (continued) Required Action A.3 are required to be performed. For Required Action A.2, the indented position of the logical connector 3 indicates that A.2.1 and A.F.! are alternate and equal choices, only one of which has to be performed. , i [xample 2. CONDITION R[ QUIRED ACTION f A. A,1 f l= AG A.2.1 E A.2.2 89 l A.3 O i l l l f ABWR 19 5/31/89 O Amendment 9 16.117
MM 21M100AP Remndard Plant ha 16.1.2.3 1.3 CT 7x 11e es Completion ilmes 1.3 1.3 CGMPLETION TIMf1 PURPDSE When a tietting Condition for Operation is not met, the individual lechnical Specifications identify Required Actions which must be completed within specified times. An understanding of the correct interpretation of these specified Completion Times is necessary for compliance with the requirements of the Technical Specifications. This section discusses the proper use ar.j interpretation cf Completion Times. COMPLET10N The Completion Time is the amount of time allowed to perform a TIME Required Action and is referenced to the time it is discovered that an ACTION $' Condition has been entered, unless otherwise specified. Required Actions must be completed within the spectfted Completion Times. When the ACTIONS' Condition changes, new Required Action (s) and Completion Time (s) are established from the time of discovery that the new Condition is entered. The Required Actions for the Condition which no longer exists are not required to be completed. When a Condition has associated Reedired Actions which result in a shutdown, the total time to reach a MODE or other specified condition is independent of what MODE ine unit is in when the Condition is entered. This is acceptable because a penalty should not be incurred if a Condition is entered while. the unit is in a lower MODE of operation where the consequences of a postulated event are significantly reduced. If a lower MODE of operation is reached in less time than allowed, the total allowable time to reach MODE 4 or other applicable MODE, is not reduced. For example, if the kequired Action is to be in MDDE 3 in 12 hours and to be in MODE 4 in 36 hours, and the Condition is entered in MODE 3, the time allowed to reach MODE 4 is the next 36 hours because the total time to reach MODE 4 ts not reduced from the allowable limit of 36 hours. Another example is if the Required Action is to be in MODE 3 in 12 hours and to be in MODE 4 in 36 hours, and MODE 3 is reached in 8 hours, the time allowed to reach MODE 4 is the next 28 hours because the total time to reach MODE 4 is not reduced from the allowable limit of 36 hours. (continued) ABWR l 10 5/31/89 O Amendment 9 16.118
I MM 2M6100AP Standard Plant bA O Completion Tisies J 1.3 1.3 COMPLET10N TIMES frontinued) COMPLETION When a Condition is entered and its Required Action is not TIME completed within its specified Completion Time, a new Condition is established. This new Condition may have its own Required Action (s)andCompletionTime(s). For example, if one component is inoptrable and is not returned to OPERABLE status within the Completion time, the new Condition may be:
' Required Action and associated Completion Time not met.*
A new Required Action would then be given for this Condition, with a new Completion Time. Some Required Actions require completion based on an initial 5 discovery time instead of time of entry into the Condition. Typically, this is the same time. However, for Example 1.3.1, in the event the Condition of a single inoperable subsystem degrades into two inoperable subsystems, Required Action B.2 requires restoration of the first inoperable subsystem in seven days from discovery of its inoperability, in some cases, this may be a shorter period of time than the 72 hours allowed in Required Action B.I. Furthemore, thould one of the two subsystems be restored to OPERABLE status and the single inoperable subsystem Condition re entered, the Completion Time for this example is based on the time of initial discovery of the remaining inoperable subsystem and not from the time of re-entry into the Condition. Example 1.3.1 CONDITION REQUIRED ACTION .COMPLET10N TIME A. One subsystem A.) Restore subsystem 7 days from inoperable, to OPERABLE discovery of status, inoperable i' subsystem l B. Two B.3 Restore at least 72 hours subsystems one subsystem to inoperable. OPERABLE st.tus. 8@ B.2 Restore first 7 days from , inoperable discovery of 1 subsystem to initial OPERABLE status. inoperable subsystem ; i (continued) i ABWR 1-11 5/31/89 9 l Amendment 9 16.119 l l , l l __ .__ _ _ _ _ . _ _ _ _ _ _ _ . . _ _ _ . ._ _. _
gg 21A6100AP
' Standard Plant Rn A .(
Completion Times
, 1.3 1.3 COMPLf71DN TIMts fenntinuedi COMPLETION Some Conditions allow a variable number of inoperable TIME components. These Conditions specify this allowance as 'one or more' or
- 5 z'. Thtse Conditions may be entered each time the number of inoperable components change, as long as the Condition still applies. Each entry has its own independent Required Action and Completion Time. This multiple entry of one Condition is not an exception to the requirement that only one Condition within an LCOs' ACTION may be entered at any time (LCO 3.0.2).
Some Conditions allow multiple reasons for inoperability. These Conditions specify this allowance as ' inoperable for one or more reasons". They may be entered each time the reasons for increrability change, as long as the Condition still applies. Each reason for inoperability has its own independent Required Action and Completion Time. This multiple entry of one Condition is not an exception te the requirement that only one Condition within an LCOs' ACTION may be entered at any time (LCO 3.0.2). One additional exception to the Coupletion Time being referenced to the entry into its Condition is noted by the use O of 'thereafter* following the Completion Time. This requires the_ initiation of a periodic re performance of the Required I i i) b Action after the successful completion of its initial performance. 1 1 In some cases 'Immediately' and 'As Soon As practicable' are used as Completion Times. *1mmediately" is generally used when a Condition cannot be permitted to continue to exist and essentially little or no time is required to complete the 1 Required Action. Examples of when "lamediately* is used include placing the Reactor Mode Switch in the Shutdown position, and declaring a system inoperable. This is because placing a switch in a specified position requires only the time necessary to perform that action and declaring a system inoperable essentially takes no time as it is an administrative action intended to start an ACT!DNS Completion Time clock. 'As Soon As Practicable
- is generally used when a Condition cannot be permitted to continue to exist any longer than absolutely necessary given the prevailing plant conditions and some period of time is required to complete the Required Action. This is because some judgement is required to determine the quickest and safest method to complete the Required Action (e.g.,
suspending operations with the potential for draining the reactor vessel) or because the Condition is such that the only ACTION that can be taken is to continue to attempt to comply with the Required Actions (e.g., restoring both RHR shutdown cooling loops to OPERABLE status). 'As Soon As Practicable" is (continued) ABWR 1 12 5/31/89 O V Amendment 9 20
MM 23A6100AP
.Simadard Plant hv4
~ O Completion Times 1.3 , 1.3 COMPLET10N TIMES fenntinued) COMPLET10N intended to ensure that appropriate attiMs are taken in TIME reasonable time, but is not intended to require extraordinary (continued) actions be taken which do not provide any real safety benefit. a i O o i 4 I J-ABWR l 13 5/31/89 4 1 0 Amendment 9 16.121 i 4
' gg 234620orr ge.ndard Plant ""^ i 16.1.2A 1A Frogmency '[)
l Frequency 4 1.4 J 1.4 FRfoufNCY PURPOSE Each Surveillance Requirement (SR) has a specified Frequency in which the Surveillance must be met to support meeting the associated Limiting Condition for Operation (LCO). An understanding of the correct interpretation of this Frequency is necessary for compliance with the requirements of the Technical Specifications. This section discusses the prcper use and interpretation of the Surveillance Frequency. FREQUENCY In examples provided and discussed below, the Applicability of the LCO is given as MODES 1, 2, and 3. EXAMPLE Example SR 1.4.1 contains the type of SR most often encountered SR 1.4.1 throughout the Technical Specifications, it specifies 49 interval (31 days) during which the associated Surveillance Requirements must be performed at least one time. Performance of the Surveillance initiates the subsequent interval. Each performance must meet the requirements of SR 3.0.2. The measurement of this interval continues at all times even when O, the Surveillance Requirement is not required by SR 3.0.1. In cases where the interval as specified by SR 3.0.2 is exceeded while not in a MODE or other specified condition for which performance of the Surveillance Requirement is required, the Surveillance Requirement must be performed prior to entry into i the MODE or other specified condition. Failure to do so would result in a " failed' surveillance and an invalid MODE change per LCO 3.0.4. Sometimes special conditions dictate when a Surveillance is to be met. These conditions apply to the Surveillance or to the Frequency or both, and are not 'other specified condition (s) in i-the Applicability" described in LCO 3.0.1, LCO 3.0.4, SR 3.0.1 and SR 3.0.4. They are however, 'otherwise stated" conditions (_ allowed by SR 3.0.1. Furthermore, these conditions may be stated as clarifying notes or as part of the Surveillance Requirement itself. Examples SR 1.4.2, SR 1.4.3, and SR 1.4.4 discuss these special conditions. SR 1.4.1 Perform a [ CHANNEL FUNCTIONAL 31 days TEST). l lL (continued) ABWR l 14 5/31/89 O
/uncedment 9 6522
I
-M M 23A6100AP - Standard Plant Rn A Frequency 1.4 1.4 FREQUENCY leantinued)
EXAMPLE Example SR 1.4.2 requires that this surveillance be performed SR 1.4.2 only at or above 25% of RTP, The phrase 'Only required..." (continued) means this Surveillance may be performed in any MODI or other specified condition where unit status would allow successful t 1 completion, but is not required to be performed unless greater than or equal to 25% of RTP. ~ ' The interval measurement for the Frequency of this Surveillance continues at all times as described in Example 1.4.1. However. if this Surveillance did not meet SR 3.0.2 while operation continued at less than 25% of RTP, it would not constitute a failure to meet the LCO. The Surveillance is not required below 25% of RTP, even though the LCO, per its Applicability, t may be required to be met. Prior to reaching 25% of kTP, if ! the Surveillance were not performed within the interval as i allowed by SR 3.0.2 it must still be performed prior to ' reaching 25% of RTP. If it is not performed prior to reaching 25% c,f RTP, the provisions of SR 3.0.3 would apply. 1
$R 1.4.2 .............N01E............
Only required with THERMAL POWER 2 25% of RTP. 4-Demonstrate the absolute 7 days . difference between the APRM channels and the calculated power is 5 2% of RTP. i EXAMPLE Example SR 1.4.3 allows the option of selecting one of two SR 1.4.3 Frequencies in which to satisfactorily perform the surveillance. The first Frequency (92 days) is like that in Example SR 1.4.1. The second option is an example of a Frequency where the measurweent of the 12 hour interval does . not continue at all times. The measurement begins only after exceeding 920 psig, i If reactor pressure is less than or equal to 920 psig at a time when the first Frequency cannot be met as required by SR 3.0.2, the second Frequency can be selected and will not be considered a failure to perform a Surveillance Requirement within the specified Frequency. Upon reaching 920 psig, 12 hours (plus i 25% per SR 3.0.2) would be allowed within which the i Surveillance must be completed. If not performed within this interval, it would then become a failure to perform a Surveillance Requirement within the specified frequency, and (continued)
- ABWR 1 15 5/31/89 i
G Arnendmen 9 16.1 23 4 4
1 ABWR 3 m io w
. Standard Plant Rev.A i 'A Frequency .
l 1.4 l l 1.4 FREQUENCY fcontinued) 1 EKAMPLE only then would MODE changes be restricted per SR 3.0.4 Once i 1.4.3 perfomed, both Frequencies ars met. Selection of either (continued) Frequency would again be allowra, except, that as long as J reactor steam done pressure is k 920 psig. the 92 day frequency must continue to be met to satisfy Example SR 1.4.3. The condition of the frequency (e.g., when reactor steam dome , pressure is 2 920 pste) may be expressed as a note or as prose ' asitisinExampleSIt1.4.3. SR 1.4.3 Verify RCIC flow greater than 92 days 800 gpm with steam supplied to the turbine from 920 to 1050 QB psig. 12 hours when reactor steam done pressure is 2 [ 920 ). psig EXAMPLE Example SR 1.4.4 is a Surveillance with a one time performance SR 1.4.4 Frequency followed by an Example SR 1.4.1 type Frequency. The logical connector 'AND* will require that both Frequencies be g s met. "Once' allows a single performance to satisfy the specified Frequency (assuming no cther Frequencies are connected by "AND*). 'Thereafter" is used to mean future performances will be established per SR 3.0.2, but only after a specified condition is first met (i.e., the 'once' performance in this example). SR 1.4.4 Verify APLHGRS are less than Once within or equal to the required 12 hours limits. after 2 25% of RTP. AliD 24 hours thereafter ABWR l 16 5/31/89 D\ V Amendment 9 16I #
1 gg 23A6100AP Remndard Plant Rn A 1. 101.2J 1JIAgalCassiderstlema O Legal Considerations L 1.5 1.5 Ltr.At CONSIMRATIONS l INTRODUCTION The Atomic Energy Act of 1954 requires that technical specifications be a part of operating 11cenes. As such, they are enforceable under federal statute as well as Title 10 of the Code of Federal Regulations. When an applicant receives a license from the Nuclear Regulatory Commission to operate a cosmercial nuclear power plant, the technical specifications are included as Appendix A to the license. Consequently, whenever a change is made to a plant's technical , l specifications, an amendment to the operating license is ' required. l There are, however, certain sections and additional items included with these Technical Specifications that are for information or convenience and are not sogally a part of the Technical Specifications or Operating License. This section identifies the legal parts (i.e. , the items that require a license amendsent to change) of these Technical Specifications, l- and those additional parts that can be changed without i i l requiring a license (or technical specification) amendment. l LEGAL PARTS- Title 10 Code of Federal Regulations, Part 50.36 delineates i those items which are to be included in technical specifications. These items to be included are: o Safety Limits o Limiting Safety System Settings o Limiting Conditions for Operation o Surveillance Requirements o Design Features, and o Administrative Controls In addition, the Use and Applications Division of these Technical Specifications (comprised of Definitions Logical Connectors, Completion Times, Frequency and Legal Considerations) is also a legal part of these Technical Specifications. Since these Technical Specifications are issutd :.s Appendix A to the Operating License, any change to the above defined sections constitutes a license amendment. As such, the requirements of 10 CFR 50.90, 50.91, and 50.92 apply. (continued) ABWR l-17 5/31/89 O Amendmer.: 9 IblO
MM 23A6100AP Standard Plant an A
,v I
Legal Considerations 1.5 L5 LFGAL CONSIDERATIONS fenntinued) m FRONT MATTER Front Matter is all the material in the front of the technical specifications used to identify and locate specific information. It includes: o Preface o Title Page o Table of Contents o List of Tables o List of Figures o List of Effective Pages None of this material is required by Title 10. Code of Federal Regulations Part 50.36, and the Front Matter does not include any requirements on the safe operation of the plant. Therefore, the Front Hatter is not a legal part of these Technical Specifications or Operating License. CROSS Cross References are included in the body of these Technical REFERENCES Specifications to assist the user in determining other applicable requirements for a common system or component. This section is not required by Title 10, Code of Federal
-Regulations Part 50.36, and is included with this set of Technical Specifications at the discretion of the Licensee. As
(' such, they are not a legal part of these Technical i Specifications or Operating License. BASES Title 10, Code of Federal Regulations, Part 50.36 includes the following statement, "A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall be included in the application, but shall not become part of the technical specifications.' Therefore, the bases are not a legal part of either of these Technical Specifications or the Operating License. The bases are prepared and controlled by the Licensee, and a change / revision to the bases shall not constitute a change / amendment to the License. l ABWR l 18 5/31/89 Amendment 9 16.1 26
~ MM . 23A6100AP Remsadard Plant Rev A 16.2 2.0 SAFETY LIMITS Safety Limits 2.0 2.0 SAFETY LIMITS 2.1 SAFETY LIMITS 2.1.1 With the reactor steam dome pressure < 785 psig or core flow
< 10% of rated core flow:
THERMAL POWER shall be s 25% of RATED fMERMAL POWER. 2.1.2 With the reactor steam dome pressure 1 785 psig and core flow 2 10% of rated core flow: MINIMUM CRITICAL POWER RATIO shall be 2 1.07 2.1.3 Reactor vessel water level shall be > the top of active irradiated fuel. 2.1.4 Reactor steam dome pressure shall be s 1325 psig. 2.2 SAFETY LIMIT VIOLATION With any Safety Limit not met the following actions shall be met: l' 2.2.1 Within one hour notify the NRC Operations Center in X accordance with 10Ci'R50.72. 2.2.2 Within two hours: A. Restore compliance with all Safety Limits, ard B. Insert all insertable control rods. W The fanoroerinte on site and off site mananement eersonnel and safety review arounsi shall be notified within 24 hours. W A Licensee Event Report shall be prepared oursuant to 10CFR50.73. The Licensee Event Report sha'il be submitted to the Commission and the Ianoropriate on-site and off-site mananement eersonnel and safety review arouesl within 30 days of the violation. 2.2.5 Critical operation of the unit shall not be resumed until authorized by the Commission. ABWR 21 5/31/89 (> Amendment 9 16.2 1
.; -- = > , - . e , ~ . . . . - a-., o - r -- a-e- e. . -- --- - - - , - - . . . - - - . - - - -- - -- - - - --a------ --- - - -e - y e ---
t- 5 1 9 f L t 4 5 I l O. v..
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Standard Plant kn A 16.3 3.0 APPLICABILITY O-l I LCO Applicability 3.0 3.0 APPLICABILITY LIMITING CONDITIONS FOR OPERATION fLCDs) LCO 3.0.1 Limiting Conditions fot Operation shall be met during the MOD (5 or other specified conditions in their Applicability except as provided in LC0 3.0.2. LCO 3.0.2 Upon discovery of a failure to meet a Limiting Condition for Operation, the associated ACT10NS shall be met. Unless otherwise stated, only one Condition in each LCO's ACTIONS may be entered at any time. If the Limiting Condition for Operation is restored prior to expiration of the specified Completion Time (s), completion of the Required Action is not required, unless otherwise stated. LCO 3.0.3 When a Limiting Condition for Operation is not met, and the associated ACTIONS are not met, or an associated ACTION is not provided, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall b& initiated within I hour to place the unit, as applicable, in;
- a. MODE 2 within 7 hours,
- b. MODE 3 within 13 hours, and
- c. MODE 4 within 37 hours.
Exceptions-tc these requirements are stated in the individual ' Specifications. 1 Where corrective measures are completed that permit operaticn l in accordance with the Limiting Condition for Operation or ACTIOMS, completion of the actions required by LCO 3.0.3 are l not required. LC0 3.0.3 is applicable in MODES 1, 2, and 3. 3 (continued) ABWR 31 5/31/89
%d Amendment 9 163-1
~ MM 23A6100AP ~ Standard Plant ' Rev A O
LCO Applicability < 3.0 LIMITING CONDIT10NS FOR OPERATION fLCOst (continued) LCO 3.0.4 When a Limiting Condition for Operation is not met, entry into a MODI or other specified condition in the Applicability shall not be made unless the associated ACTIONS for the MODE or other specified condition to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time. This shall not prevent changes in MODES or other specified conditions in the Applicability which are required to comply with ACTIONS. l l Exceptions to LCO 3.0.4 are stated in the individual l Specifications. These exceptions allow MODES or other l specified conditions in the Aplicability of the LCO to be 1 entered while meeting the associated ACTIONS which do not l permit continued operation in the MODE or other specified condition for an unlimited period of time. LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative (- control to perform required Surveillances to demonstrate its l OPERABILITY or the OPERABILITY of other equipment. These ' l Surveillances are only those which otherwise could not be performed without returning the equipment to service. - l LCO 3.0.6 When equipment is rendered inoperable for performance of its LCO's Surveillance Requirements to satisfy SR 3.0.2, the Required Actions may be delayed for up to 8 hours to permit completion of the Surveillance Requirements. LCD 3.0.7 When a support system is inoperable and an LCO for that support system is specified in the Technical Spectitcations, the supported system is not required to be declared inoperable solely due to support system inoperability. Only the support system LCO's ACTIONS are required to be entered. This is a clarification of the definition of OPERABILITY. When a support system is inoperable and there is not an LCO for that support system specified in the Technical Specifications, the impact of the inoperability or degradation of the support system function on the OPERABILITY of the supported system shall be evaluated. Upon determination that the supported system is inoperable, the ACTIONS of its LCO shall apply. (continued) ABWR 32 5/31/89 O Amendment 9 16.3-2
ABWR m6 imp S*m=dard Plant am x
, ,~\. j 'o )'
i ' LCO Applicability 3.0 LIMITING CONDITIONS FOR OPERATION f tt01) frontinued) LCO 3.0.8 Special Operations LCOs in Section 3.10 allow specified Technical Specification requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other Technical Specification requirements remain unchanged. Compliance with Special Operations LCOs is optional. When a Special Operation LCO is not desired to be met. entry into its MODE or other specified condition in the Applicability shall be made in accordance with the other appitcable Specifications. t d l i l l l L ABWR 33 5/31/89 O Amendaunt 9 16.3 3
23A6100AP Standard Plant Rev A 1 SR Applicability
, 3.0 3.0 APPLICABILITY SupVE1LLANCE REQUIREMENTS (SRs)
SR 3.0.1 Surveillance Requirements shall be met during the MODES or other specified conditions in the Applicability for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement. Satisfactory performance of the Surveillances within the specified Frequency is required to verify the LCO is being met. Surveillance Requirements do , not have to be met on inoperable equipment. SR 3.0.2 The specified Frequency for each Surveillance Requirement is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency as measured from the , previous performance, or as measured from the time a specified condition of the frequency is met, for Frequencies specified as 'once', the interval extension above does not apply. If a Required Action's Completion Time requires periodic performance of 'once per....' the Frequency extension above applies to the Completion Time intervals.
- Exceptions to these requirements are stated in the individual Specifications.
SR 3.C.3 When a Limiting Condition for Operation is not met due to failure to perform a Surveillance within the specified Frequency, the requirement to declare the equipment inoperable may be delayed for up to 24 hours from the time it is identified that the Surveillance has not been performed to permit the completion of the Surveillance. If the Surveillance is not performed within the 24 heur allowance, the Completion Time of the Required Actions begins inunediately upon expiration of the 24 hour allowance. When the Surveillance is performed within the 24 hour allowance and the Surveillance Requirements are not met, the Completion Time of the Required Actions begins immediately upon the failure of the Surveillance. (continued) ABWR 34 5/31/89 O Amendment 9 16,3 4
.l c
< JAB M usuoox, ,
i s ' StamI=rd Plant a, A j l L SR Applicability 3.0 3.0 APPLICABILITY SURVEILLANCE REQUIREMENTS fSR11 frontinued) . SR 3.0.4 Entry into a MODE or other specified condition in the
. Applicability shall not be made unless the Surveillance Requirements associated with a Limiting Condition for Operation have been met. This provision shall not prevent passage through, or to, MODES or other specified conditions as required to comply with the Required Actions. Exceptions to these requirements are stated in the individual Specifications.
SR 3.0.5 Inservice inspection and testing of ASME Code Class 1. 2 and 3 components shall be applicable as follows:
- a. Inservice inspection of components, and inservice testing '
of pumps and valves, shall be in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda (the Code) except where relief has been requested pursuant to 10 CFR 50.55s(g)(6)(1).
- b. Test Frequer.cies specified in Section XI of the Code shall be applicable as follows:
Code Terminoloov Frecuency Weekly 7 days Monthly 31 days Quarterly or every 3 months g2 days Semiannually or every 6 months 184 days Every g months 276 days Yearly or annually 366 days Biannual or every 2 years 732 days
- c. The provisions of SR 3.0.2 are applicable to the above Code required Frequencies,
- d. Performance of Code inservice inspection and testing shall be in addt? ton to the specified Surveillance Requirements,
- e. Nothing in the Code shall supersede the requirements of these Technical Specifications.
=
p i L L l l l ABWR 3-5 5/31/89 O 16.3-5 Amendment 9 I
23A6100AP . Standard Plant , Rev A i SECTION 16A g ij CONTENTS Section Tillt East 16.4.1 3.1.1 Shutdown Margin 16.4 1 i 16.4.2 3.1.2 Control Rod Onerabilltv 16.4 5 16.43 3.13 Control Rod Scram Times 16.4 10 3.1.4 Control Rod Scram Accumulators ' 16.4.4 16.4 14 16.4.5 3.1.5 Control Rod D-ive Countinn 16.4 16 16.4.6 3.1.6 Rod Pattern Control 16.4 18 16.4.7- 3.1.7 Standby Llauld Control Systent 16.4 20 O 16.4-ii Amendment 9
g_ q MM 2arsiocre Sta=dard Plant ._ Rn A 16,4 3.1 RFACTIVITY CONTROL SYSTEMS m 16.4.1 3.1.1 Shutdown Margin SHUTDOWN MARGIN 3.1.1 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN HARGIN LCO 3.1.1 SHUTDOWN MARGlN shall be: A.- 2 0.38% Ak/k, with the highest worth control rod ELir analytically determined, j QB B. t 0.28% Ak/k, with the highest worth control rod glir determined by test. APPLICABILITY: MODES 1, 2, 3, 4, and S. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SHUTDOWN MARGIN less A.] n Restore SHUTDOWN MARGIN 6 hours than specified in to required limits.
'l }' MODE 1 or 2.
U r 1 B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not met. C. SHUTDOWN MARGIN less C.1 Fully insert all I hour than specified in insertable control rods. MODE 3. (continued) ABWR 3.1-1 5/31/89 I Amendment 9 16.4 1
ABMit 22461oo^r
. Standard Plant hv^
O SHUTDOWN MARClh 3.1.1 ACTIONS fcontinued) CONDITION' REQUIRED ACTION COMPLETION TIME D. $HUTDOWN MARGIN less D.) Fully insert all I hour than specified in insertable control rods. MODE 4 M D2 Ensure Secondary As soon as . Containment is OPERABLE. practicable b.0 ! D.3 Ensure at least one SGTS As soon as subsystem is OPERABLE. practicable M D.4 Ensure at least one As soon as Secondary Containment practicable Isolation Valve and associated actuation instrumentation is OPERABLE in each associated penetration not isolated. (continued) l ABWR 3.1-2 5/31/89 O Amendment 9 16.4 2
-MM 23A6100AP Standard Plant hv ^
+
.m SH'JTDOWN MARGIN 3.1.1 ACTIONS fcontinued)
CONDITION REQUIRED ACTION COMPLETION TIMI E. SHUTDOWN MARGIN less E.1 Suspend CORE ALTERATIONS lamediately than specified in except for control rod MODE 5. insertion. E.2 Fully insert all As soon as insertable control rods practicable in core cells containing one or more fuel assemblies. M E.3 Ensure Secondary As sonn as Containmer.t is OPERABLE. practicable M
~
E.4 Ensure at least one SGTS As soon as
-{. subsystem is OPERABLE. practicable
!? @ E.5 Fnsure at least one- As soon as Secondary Containment practicable Isolation Valve and associated actuation l instrumentation is OPERABLE in each l- associated penetration
- not isolated.
l' l i t. J l-. l l l , 1 l ABWR 3.1 3 5/31/89 l O l Amendment 9 16A3
MM 23A6100AP i Senndard Plant Rn. A i SHUTDOWN MARGIN
, 3.1.1 SURVIIttANCE REQUIREMENTS SURVE!LLANCE FREQUENCY SR 3.1.1.1 Demonstrate SHUTDOWN MARGIN. .....N0TE..... i Only required after fuel movement or control rod replacement within the RPV.
Once within 4 hours after criticality SR~3.1.1.2 Demonstrate SHUTDOWN MARGIN of each fuel Prior to movement during fuel loading sequence, fuel movement O CROSS. REFERENCES TITLE NUMBER Reactor Protection System Non Coincident Mode ( } Secondary Containment Isolation Actuation Instrumentation [ ] Secondary Containment 3.6.4.1 secondary Containment Isolation Valves 3.6.4.2 Standby Gas Treatment System 3.6.4.3 Control Rod Withdrawal . Hot Shutdown 3.10.3 Control Rod Withdrawal Cold Shutdown 3.10.4 Control Rod Drive Removal Refueling 3.10.5 l ABWR 3.1 4 5/31/89 l 4 i Amendment 9 16.4-4 I I 1 l
- MM 23A6100AP Standard Plant h^
; 16A.2 3.1.2 Centrol Rod Operability I
l Control Rod OPERABILITY i 3.1.2 ' 3.1 REACTIVITY CONTROL SYSTEMS 3.1.2 Control Rod OPERABILITY e i l LCO 3.1.2 All control rods shall be OPERABLE. APPLICABILITY: MODES I and 2. l
.........................N0TE-........................ .
Conditions A through F may be concurrently applicable. I ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One withdrawn control M ..........N0TE.......... rod stuck. A stuck control rod may be bypassed in Elli as allowed by LCO 1R;1 Block Instrumentationi, if required, to allow con-O .$f ....... Restore stuck control I hour 1 rod to OPERABLE status. l B. Required Action and 8.1 Disare the associated I hour associated Completion control rod drive. Time of Condition A not met. E y ..........N0Ti.......... Not applicable when
> 10% of RTP.
Verify all inoperable I hour control rods not in compliance with SESS are separated by 2 two OPERABLE control rods. M (continued) ABWR 3.1 5 5/31/89 Amendment 9 16.4-5
h 23A6100AP 1 Standard Plard anv ^
)
O' i Control Rod OPERABILITY 3.1.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TiH B. (continued) 3.J ..........N0TE.......... Not aplicable when s the LPSP of the E11$.
-Perform SR 3.1.2.2 24 hours and SR 3.1.2.3 for each withdrawn OPERABLE jj control rod.
M l B.4 Demonstrate SHUTDOWN 72 hours MARGIN is within the limits of LCO 3.1.1. C. Required Actions and C.) Be in MODE 3. 12 hours associated Completion Times of Condition B - not met. l' D. More than one D.1 Disarm the associated I hour withdrawn control rod control rod drive, stuck. 8 D.2 Be in MODE 3. 12 hours (continued) 1 '~ , ABWR 3.1-6 5/31/89 O Amendment 9 16.4 6
_ . _ _ _ . _ = _ _
<1(
i MM 21A6100AI' RemmAmed Plant - Rev A - { l l Control Rod OPERABILITY
.i.:
ACTIONS feettinued) , CONDITION REQUIRED AC110N COMPLETION TIMC , [. Less than or equal to eight control rods L1 ..........$1[..........
- 1. Inoperable control ;
inoperable for reasons rods may be bypassed e i other than Conditions in RC&l$ as allowed A or D. byLC0[RodBlock i instrumentation , , required,toal) low ' insertion of inoperable control rod (s) and continued operation, r L Inonerable control rods with failed
- maior drives can oniv he fully inter.
tad by individual ALIAL
,e Fully insert inoperable 1 bour
, control rod (s). F e$ 4 E.2 Disarm the associated 2 hours > conttel rod drive (s). l 5 l L2 ..........N0TE.......... - Not applicable when
> 101L of RTP.
Verify all inoperable 3 hours control rods not in compliance with 3 1S are separated by 2 two OP[RABL[ control rods. (continued) l ABWR 3.1 7 $/31/89 A Amendment 9 16.4 7
MM 2M6100AP 'leadard Plant - Est.a O Control Rod OP! MBltli) 3.1.2 ACTIONS frantinued) CONDITION REQUIRED ACTION COMPl[110N TIM [ F. Reevired Actions and F.1 Se in MODE 3. It hours associated Completion lines of Condition [ not met. 08 Greater than eight inoperable control ro68. O ABWR 3.1 8 $/31/89 O Anundment 9 16 4-8
]
1 i 21A610QAP Remmulard Plant u._ x ; t i 1 L.v) Control Rod OP(RABILITY l 3.1.2 { suRytf tt AN!r treultretNis ! l
$URVIILLANCE FR!0V[NCY l
$R 3.1.2.1 Determine position of all control rods. 24 hours ER 3.1.2.2 Insert each fully withdrawn control rod 7 days when at least one Lin. greater than the LPSP of the ELL 11
$R 3.1.F.3 Insett each partially withdrawn control 31 days when ,
rod at least one Lin . greater than the LPSP of the ELLli
. ,ew '\ CRD55 PffERENCf$
t llitt NUMB (R i
$HUTDOWN MARGIN 3.1.1 D.od Pattern Control 3.1.6 Control Rod Block Instrumentation [ ]
l l' l l ; l l l l , 1 i AR'/R 3.1 9 5/31/89
%J )
Amendment 9 16 4-9
MM 21A6100AP menaard Plant art ^ 16.4.3 3.1.3 Coettel Rod Scram Times O Control Rod Scram lines 3.3.3 3.1 R(ACTIVITY CONTROL SYSTEMS 3.3.3 Central kod Stras Times LCO 3.1.3 All control rods shall hava s cram times less than or equal to the limits shown in Tabli J.1.31. APPLICABIL11Y: W) DIS I and 2. AC1 IONS CONDITION REQUIRED AC110N COMPLET10N TIMI A. One or more control A.) Declarecontrolrod(s) Ismediately rods ' slow
- with scram with scree times times greater than >[ ] seconds limits as shown in inoperable.
Table 3.1.3 1.
.. g A.! Verify 5 20% of the 12 hours OP(RABLE control rods tested in SR 3.1.3.2 are
' slow'.
blQ A.3 Verify 5[ ] OP[lABLE 12 hours control rods have scram times greater than limits. A!O A.4 Verify no more than two !! hours OPERABLE ' slow
- control rods occupy adjacent locations.
l B. Required Actions and B.1 Be in MODE 3. 12 hours associated Completion limes of Condition A not met. AC'.'R 3.1 10 5/31/89 Amendment 9 16/ 10
.i N 21A6100AP Se==dard Plant % l -3 (V) ]
Control Rod $ cram Times 3.1.3 ,
$URVilLLANtt RIOUIRtutNTS +
$URVEILLANCE FRt0VINCY I i
............................ g ............................
During single or control rod agir scram time tests, the CRD pumps shall be isolated from the associated scram accumulator.
$R 3.3.3.1 Measure all control rod scram times .....N0TC.....
with reactor steam dome pressure 2 950 Only required psig. after fuel movement ; within the ' RPV or after each reactor shutdown i
> 120 days. .
(% i Once prior to exceeding
\ 40% of RTP SR 3.1.3.2 Measu-e control rod scram times with 120 days of reactor steam dome pressure 2 950 psig cumulative for 2 10% of the control rods, operation in MODE 1
$R 3.1.3.3 Measure affected control rod scram times .....N0TE... .
at any reactor steam dome pressure. Only required when work on control rod or CRD system could affect scram times. Once prior to declaring contrtl rod (s) OP(RABLE (continued) ABWR 3.1 11 5/31/89 r ()% Amendenent 9 16.4-11
ggg gu6100AP i' Seeadaed Pimnt *^ G , I Control Rod Scram Times { 3.3.3 SURVf1LLAktt Rtou1RtMENTS frontinuedi SURVIILLANtt FRIQU(NCY
$R 3.3.3.4 .................N0TE.....................-
Performance of this $R satisfies $R 3.3.3.1 and SR 3.3.3.3 for affected control rods.
............................=..............
l Measure affected control rod scram tie.es at .....N0T[..... reactor steam dome pressure t 950 psig. Only required ! when work on , contral rod or . CRD system could affect t scram times. ' Once prior to esteeding 40% of RTP
- Cross REFER [Nft$
O + llTLE NUMB (R Control Rod OPERABILITY 3.1.2 ABWR 3.1 12 $/31/89 O Amendment 9 16012
~
N. b i-MN 23A6100AP
*==dard Plant u4 O
'J i
l t i Control Rod Scram Times 3.3.3
- TABLE 3.1.3 1 (Page 1 of !)
toetrol kod Strae Tht SCRAM TIME I $ "I ROD (seconds) ' P0$1110N PERCENT REACTOR $TCAM DOME PRE $$URE(b) IN5tRT10N O psig 950 psig 1050 psig ; 10 (c) [ ] [0.42) 40 (c) [ ] [1.00) 1 60 [ } [ ] [1,44) O (a) h ximum scram time from fully withdrawn position, based on de energiration of scram pilot valve solenoids as time zero. (b) For intermediate reactor steam dome pressures, the scram time criteria are determined by linear interpolation. IC) For reactor steam dome pressure < 950 psig, only the 60% insertion scram time limit applies. , t l l ABWR 3.1 13 5/31/89 l. 1 { Amendment 9 16.4-13
i M 21A6100AP Standard Plant m.v A i 16AA 3.1A Control Rod Scram Accumulators i Oi Control Rod Scram Accumulators 3.1.4 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Control kod Scram Accumulators j LCO 3.1.4 All control rod scram accumulators shall be OPERABLE. APPLICABILITY: MOD [5 1 and 2.
...................... g g........................
Conditions A and a may be concurrent 1r acclicable, ACTIONS CONDITION R[0VIRED ACTION COMPl[ TION TIME L one or more control A.) Restore inoperable I hour scram accumulator (s) control rod scram inonerable. accumulator to OPERABLE status. Q3 A.2 Declare the associated I hour control rod (s) inoperable. L More than one inco. B.) Place the Reactor Mode 2 bours erable control rod Switch in the Shutdown scram accumulatcr position, associated with cartially or ful' v withdrawn contro IR A l l i l ABWR 3.1 14 $/31/g9 O Amendment 9 16.4-14
~
MN 21A6100AP A*madard Plant Rev.A 3 O Control Rod Scram Accumulators 3.1.4 1URVffttANCE RfoUIRIMENT$ SURV[lLLANCE FR[0V!NCY SR 3.1.4.1 Verify control rod scram accumulator 7 days pressureisa!1850)psig. CRD15.ptFERfNCf5 TITLE NUMB (R Control Rod OP[RABILITY 3.1.2 Control Rod Scram Times 3.1.3 l l l l i l l l
- ABWR 3.1 15 5/31/89 Amendment 9 16415
l l MN twtooAP wA e i RemmA=ed Plant i i l 16.4J 3.1J Centrol Rod Drive Coupling , O\! i CRD Coupling 3.1.b 3.1 RtACTIVITY CONTRDL SY51[M5 3.1.5 contre 1 had Drive toumline LCD 3.1.5 All control rods sh 11 be ciupled to their drive. ; 1 APPLICABILITY: MDDIS I and 2. ACTIONS COND1110N R[ QUIRED AC110N COMPLE110N 11MI A. One or more control A.1 Retouple control rod. 2 hours ! l rod (s) not coupled to its drive. QB A.2 Declare the control rod 2 hours I inoperable, i 1 l l-l- 0l j i 1
)
l I ! l l, 1 ABWR 3.1 16 5/31/89 O Amend sent 9 16.4-16
.. .~ - -
j MM 2M6100AP Reasulard Plant Rev A i 1
.:r'\ :
(,) 3 l
'. CRD Coupling 3.1.6
$UPvtitt ANtt Pt001RrMEN7s '
$URV[!LLANC[ FR[QUINCY
$R 3.1.6.1 Demor. strate each control rod does not go to once the first the overtravel position, time the control rod is withdrawn to
- Full Out*
after fuel mover **t within the RPV 8!iD Once prior to declaring control rod (s) , OP[RABLE when > work on control rod or CRD system ' could affect coupling.
-i]
CROSS REffRINCES TITLt NUMBER Control Rod OP[RABILITY 3.1.2 l l 1 l l l' l-l l l ABWR 3.1 17 5/31/89 [D v Amendmes: 9 6A17
l t 2M6100AP , R*==dard Plant a,v ^ 16.44 3.14 Rod Patterm Control O Rod Pattern Control 3.1.6 3.1 RIACTIVITY CONTROL SYSTEMS 3.1.6 Red Pattern Control LC0 3.1.6 OPERABLE control rods shall comply with the requirements of the cano Withdrawal secuente Restrictions (GWSR). APPLICABlll1Y: M00Cf I and 2 with THtRMAL POWER s 10% of RTP. ACTIONS CONDITION REQUIRED ACTION COMPLE110N 11M[ L Less than or equal to AJ ..........N01E..... - eight OPERABLE control Affected control rods rods not in compliance may be bypassed in RC&l5 with idG3 as allowed by LCO ' Control Rod B1gd; Jnstrumentation) , Move affected control 8 hours rod (s) to correct position. QB A.2 Declare affected control 8 hours rod (s) inoperable. L More than eight B.1 Suspend withdrawal of lamediately OPERABLE control rods control rods. not in compliance with fdd.B. A!iD B.2 Place the Reactor Mode I hour Switch in the Shutdown positicn. l l ABWR 3.1.!B 5/31/B9 Annendment 9 16.4-18
)
ABWR zusione Standani Plant a- A i A k i Rod Pattern Control i 3.1.6
$URVI1LLANCt RE0LilREMINT$
SURVIILLANCE FREQUINCY SR 3.1.6.1 Verify all OPERABLE control rods comply 24 hours with GWiB. CROSS RfftRtNCE$ , TITLE NUMB [R Control Rod OPERABILITY 3.1.2 Control Rod Block Instrumentation [ ] Control Rod Testing Operating 3.10.7 I l v ABWP. 3.1 19 5/31/89 t
\.
Amendment 9 16.4-19
s ! MM 2M6100AP Standard Plant wA 16.4.7 3.1.7 Standby Uguld Control System
$LCS 3.1.7 i
3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standhr Liauid control traten , LCO 3.1.7 The Standby Liquid Control System ($tt$) shall be OPERABLE. APPLICABILITY: MODES I and 2. At110NS CONDITION REQUIRED ACTION COMPLETION TIME A. One $lC$ subsystem A.) Restore inoperable 7 days from inoperable, subsystem to OPIRABLt discovery of status, inoperable subsystem , 8. Both $LC$ subsystems B.1 Restore at least one 8 hours inoperable. subsystem to OP[RABLE status. B.2 Restore the initial 7 days from inoperable subsystem discovery of to OPIRABLE status, initial inoperable subsystem C. Required Actions and 0.1 Se in MODE 3. It hours associated Coapletion Times of Condition A or B not met. ABWR 3.1 20 5/31/89 O Amendment 9 16.4-20
p ! l ABWR 2 m ia m ! Standard Plant *^ (.- I Stt$ 3.1.7 i suRvf fit &Nf f RfoUIREMENTS
$URV[lLLANC[ FR[0V[NCY SR 3.1. 7. I ' Verify available volume of sodium 24 hours pentaborate solution is t 6103 gallons, f
SR 3.1.7.2 Verify temperature of sodium pentaborate 24 hours ' solution is within limits of Figure 3.1.7 1.
$R 3.1.7.3 Verify temperature of pump suction piping 24 hours is within the limits of Figure 3.1.7 1.
a
$R 3.1.7.4 Demonstrate by chemical analysis the 31 days concentration of boron in solution is within the limits of Figure 3.1.7 1. &@
f Once within t 24 hours after water or boron added to solution AE Once within 24 hours efter solution
- temperature ,
is restored within the limits of . Figure 3.1.7 1 (
$R 3.1.7.5 Verify each valve in flow path not 31 days locked, sealed or secured in position.
is in its correct position. (continued) ABWR 3.1 21 5/31/89 O Amendment 9 16421
t M 21A610QAP i
.Se=d-rd Plant m.v. A I
T
$lC$
3.1.7 EURVE1LLANCE R1001kEMEh'T1 (continued)
$URV[lLLANC[ TRt0V[NCY ,
i
$k 3.1.7.6 Demonstrate each pump tevelops a flow rate 92 days t 50.0 gpe at a discharge pressure 2 1223 psig. .
ER 3.1.7.7 Demonstrate the flow path through one $lC$ 18 months on i subsystem from pump to reactor pressure a $1AGGtR[p vessel is available by initiating iht 1[$1 BA$l$ subsystee and pumping desineralized water to the reactor pressure vsstel. 3
$R 3.1.7.8 Demonstrate all heat traced piping between 18 months ,
storage tank and pump suction is unblocked. i M Once within 24 hours after solution temperature is restored within the limits of Figure 3.1.7 1 CROS$.
REFERENCES:
None 1 l l l l l ABWR 3.1 22 $/31/09 I O' Amend:nent 9 16.4-22
- - - ,- .- - - - , - - - . - , . - - , - v -- v-- --
f:'y ' 21A61 MAP Sea ' = d P lanit nev A r
'[
Stt$ 3.1.7 4 S I 14 l.i t - t.16 - gegegr3C ShAvitY . Se , 6 30'c 6AtonAtt4u .U
~
TenPBhatcht 'C g ... . .. d { g.... taas E
>><- E
. .. g in a
. 183- = 46 4Att 480 * ' ! ' ' ' *
- i 0 80 80 49 46 CoNCthTRATION til (WEIGHT PERetWT SODIUM PtwtA60aATE IN $0LUTIONI Figure 3.1.71 (Page 1 of 1)
Sodium Pentaborate solution Temperature Versus Concentration Requirements ABWR 3.1 23 5/31/89 O Amendment 9 16.4 23
$$ 23^61MAI' Standard Plant Rev.A
- SECTION 16.5 1 i
CONTENTS Section g 163.1 3.2.1 Average Planar linear Heat Generation Rate 163 1 163J 322 Mlaimam CrillM Power Ratio 163 2
)
164.3 3133Jaear Heat Generation Rate 163 3 I i l l lO i l O 163-ii l Amendment 9
}
MM 23A6100AP Standard Pisnt ma ^ '
- g. 16.5 3.2 POWER DISTRIBUTION LIMITS .
t
\ 16.5.1 3.3.1 Average Planar unear Heat C4eeration Rate i
t APLHGR 3.2.1 - 3.2 POWER Dl3TRIBUTl0N LIMITS 3.2.) AVERA&E PLANAR LINEAR NEAT RENERATION kATE LCO 3.2.) All AVERAGE PLANAR LIN[AR H[AT G[N[RAfl0N RAT [$ (APLHGRs) shall be less than or eoual to the limits SPecified in the [COR[ OPERATING LIMIT $ REPORT). APPLICABILITY: THERMAL POWER 2 !$% of RTP. AC110Ns CONDji10N Rt0VIR[D ACTION COMPL(TION TIME i A. Any APLG R greater A.] Restore APLHGR to 2 hours than the required less than or equal to lieits. the required limits.
- 8. Required Action and B.1 Reduce THERMAL POWER to 4 hours associated Completion < 25% of RTP,
'/' Time of Condition A not met.
SURVtittANCE REOUIREMENTS SURV[lLLANC[ [REQUtNCY
$R 3.2.1.] Verify all APLHGRs are less than or equal Once within to the required limits. 12 hours after 2 25% of RTP AtiD 24 hours thereafter l
CROS$.R[f[R[NC[$: None. ABWR 3.2 1 5/31/89 i v Anwndment 9 1631 I
~
l l Q 21A6100AP Rimsuled Pimrit Rev A 16J.2 3.22 Minimos Critical Power Ratio EPR , 3.2.2 , 3.2 POWER DISTRIBUTION LIMITS 3.2.2 alulital taff1fAL POWER RATIO LC0 3.2.2 The MINIMUM CRITICAL POWER RATIO (EPR) shall be greater than or ogs41 to the MCPR limit $Pecffled in the ' [ CORE OPERATING LIMITS REPORT). APPLICAtlLITY: THilMAL POWER 1 25% of RTP. Affl0N1 CONDITION R(0UIR[D ACTION COMPLifl0N TIMI l A. MCPR 1ess than the A.) Restore MCPR to greater 2 hours i required limit, than or t0ual to the required M elt. B. Required Action and B.1 Reduce THERMAL POWER to 4 hours , associated Completion < 25% of RTP. Time of Condition A l-i not met. Suavf1LLANEf RfDUIREMENTS
$URV[lLLANC[ FR[QUINCY
$R 3.2.2.1 Verify MCPR is greater than or equal Once within to the required limit. !! hours after t !$t of RTP AliD 24 hours thereafter r CROSS RLFfRENCES: None ABWR 3.2 2 5/31/89 O
Amendenset 9 16,$-2
i 23A610MP
.Sta==d=M Plaeit a,,_ x 103 3.23 IJacar Heat Generation Rate U,rx LHGR 3.2.3 32 POW [R DISTRIBUTION LIMITS 3.2.3 Llmram ut&T acarmafl0N kAff finalitable te non.At Fuel Dnivi LC0 3.2.3 The LINEAR HEAT $[NERAT10N RAT [ (LHGR) shall be less than or equal to the limits specified in the [ CORE OPERATING LIM 11$ REPORT).
APPLlfABILITY: THIRMAL POWER 2 25% of RTP. ACT1bNS CONDITION RIQUIRED ACTION COMPLETION TIMI A. Any LHGR greater than A.) Restore LNGR to less ! lours the required limits, than or equal to the required limits. B. Required Action and B.) Reduce THERMAL POWIR to 4 hours associated Completion < !$t of RTP. , p Time of Condition A not met, i l i f EURVE1LLANCf REDU1REMENTS
$URV[lLLANC[ FRIQU[NCY
$R 3.2.3.1 Verif:c all LHGRs are less than or equal to once within the r4 Quired limits. It hours after 3 25% of RTP A@
24 hours thereafter CRO$$
REFERENCES:
None ABWR 3.2 3 $/31/E9 Amendment 9 16.5-3
I 21A6100AP Rimsulard Plant nev a 16,6 3.3 INSTRUMENTATION ! [] (Later) f i l i i t i l I I
)
( ! N i i t l 1 l 1 I l i Amendment 9 1M 1 l
23A6100AP Standard Plant a,a : I SECTION 16.7 :
,m CONTENTS ! .(w) i Secilon M g ;
i 16.7.1 3.4.1 Recler '-*le Paimma Cu. 1.= 16.7 1 f 16.7.2 3.4.2 histv/E*ht Valves 16.7 4 16.7.3 3.4.3 Operational Imakage 1676 , 16.7.4 3.4.4 EneelRc Activitt 16.7 8 16.7.5 3.4.5 Residual llent kranoval. Shutdown 16.7 10 16.7.6 3.4.6 Remetor Coolant System fnssure/Temnerature Limits 16.7 12 , 16.7.7 3.4.7 Reactor Steam Dome Pressure 16.7 15 t I f 5 V t 16.7 li I Amendment 9 l
l ABWR nulmir ! Remsulmati Plant bA l 16.7 3.4 REACIOR COOLANT SYSTEM o I j fs._)\ 86.7.1 34.1 Radrumletion Ptimps Opernelag 1 l Recirculation Pumps Operatin l 3.4. 3.4 RIACTOR COOLANT SYSTEM LU hacirculation Paes haeratina l t Lt0 3.4.1 At least nine reatter internal cumos f tlPs) ahall be in neeration. APPtitABILITY: MODIS I and 2. ACTIONS CONDITION REQUIRfD ACTION COMPLE110N llMt
.........ggy..........
Provisions of tto 3.0.4
'are not ano11 cable.
L One of the reouired W Verify reatter cover is I hour RIPS not in coeration. < 9 n R1P.
.O
\ ......... g ..........
Provisions of tro 3.0J are not acclicable. L Two of the reouirt.d L1 Verify reatter cower is I hour RIPS not in coeration. < 90% R1P. L hree or four of the LJ Reduce reacMor cower 4 hours reoutred RIPS not in to < 2h RTP. oceration. 80 L L 1 testore at least seven 12 hours from tlPs to coeration. initial dis. covery of less QB less than anven RIPS in coeration. L L2 f.e in MODE 3 P4 hours from nitial dis. covery of less than seven RIPS in coer. ation. (continued) ABWR 3.4 1 5/31/89 i
-(
Asnendswet 9 16.7 1 ( .-
' MM 21A6100AP Danda.M Plant hv^ j 1
i Recirculation Pumps Operatin l i 3.4. ACTION $ (Continued) CONDITION REQUIRED ACTION COMPLETION TIMI 4 L Less than five RIPS L1 kaduce reactor onwer 4 hours f l in noeration, to < $1 RTP. All0 LL1 testore at least seven 12 hours from , tlPs to operation. initial dis. J tovery of less g than seven Rips in oogI. naa. ; L L2 Be in MODI 3. 24 hours from initial dis. covery of less
-than seven llPs in operation. i D
ABWR 3 4.g 5/31/89 Amendment 9 16.7 2
c
- f. NUI Sinad=ed Plant kev ^ '
)
/ 16.72 3.4J Safety / Relief Yalves V '
$/RVs 3.4.2 .
3.4 REACTOR COOLANT $Y$1[M 3.4.2 infatv/ Relief Valves - P LCO 3.4.2 The safety function of > 12 $afetv/ Relief Yalves ($/RVs) shall be OPERABtt, APPLICABILITY: MODES 1, 2, and 3.
&CJ.10NS COND1110N REQUIR[D ACTION COMPLETION TIM [
A. One or more of the A.] Be in MOD [ 3. 12 hours required $/RVs inoperable, &@ A.2 Be in MODE 4. 36 hours i
'w/
ABWR 3.4 4 5/31/89
. I Amendment 9 16.7 4
i MM 2M6100AP ma ^ Standard Plant l l O1' i i i $/RVs ] l 3.4.2 i RURVf1LLANCE kl001kfMINTS
$URV[lLLANC[ FR[QU[NCY l
l SR 3.4.2.1 Demonstrate the safety function lift According to i setpoints of the required $/RVs are as $R 3.0.5 l follows: OR Number of Setpoint 18 months _1/RVs fetin) - - 1 1160 + 11.5 2 1360 i 11.6 3 1170 I 11.7 3 1180 i 11.8 3 1190 i 11.9 , I 1
$R 3.4.2.2 Demonstrait each required $/RV opens when 18 months manually actuated. i 12 hours when reactor steam dome pressure is l 2I )Ps19 l
CRD15.REFERENCf5 TITLE NUMBER [CCS Operating 3.5.1 l l I I ( i j l I A8WR 3.4 5 5/31/89 we 9 16.7 5 i l l
ABWR 2=ime i standard Plant w^ J
,m .
( ) w/ ' i i i pecirculation Pumps Operating 3.4.1
$UPVIILLANtt RfDUIREMINTS
$URV[lLLANCE FRIOU(NCY 1R 3.4.1.1 Verify at least nine PIPS are in eneration. 24 hours CROS$.REFERENtt$
11TLE NUMBER ; Reactor Coolant System Pressure / Temperature Limits 3.4.6 ' v f~ i , l l ABWR 3.4 3 5/31/89 O Amendment 9 16.7 3 1
MM. 23A6100AP m* ^ Remndard Plant . c: 16.7J ' 3A.3 Operational tankage Operational Leakage 3.4.3 3.4 REACTOR COOLANT $Y$1gM 3.4.3 Omarational Lankage
-LC0 3.4.3 Reactor coolant system LEAKAGE shall be lia.ited to:
A. No Pressure Boundary Leakage, E. B. s I gpm total Ur.sdentified Leakage averaged over any 24 hour period, M C. s 25 gpm Total Leakage averaged over any 24 hour period. APPLICABillTY: MODES 1, 2, and 3.
......................N0Ti........................
Conditions A and B ney be concurrently applicable. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A .- Unidentified Leakage A.! Reduce leakage to within 4 hours
> 1 gp'a. the limits.
E Total Leakage
> 25 gpm.
E Unidentified Leakage
> 1 gpm and Total Leakage > 25 gpm.
(continued) ABWR 3.4-6 5/31/89 O Amendment 9 16.7-6
o
' MN 23A6100AP l Stand.ed Plant neo
?
_f -, 4 s Operational Leakage , 3.4.3 AET10NS frontinued) , CONDITION REQUIRED ACTION COMPLE110N TIME B. Required Actions and B.1 Be in MODE 3 12 hours associated Completion Times of Condition A &@ not met. B.2 Be in MODE 4. 36 hours DE Any Pressure BLundary Leakage. SURVEILLANCE REQUIREMENTS
$URVEILLANCE FREQUENCY
$R 3.4.3.1 Verify the reactor coolant system LEAKAGE 12 hours is less than or equal to the required limits.
CROS$
REFERENCES:
None ABWR 3.4-7 5/31/89 O Amendment 9 16.7-7 w-a r -w-w, -.--
MM _ 21A6100AP Siendard Plant Rev.A 16.7A 3.4A SpectSc Activity O Specific Activity 3.4.4 3.4 REACTOR COOLANT SYSTEM 3.4.4 Saecific Activity LCO 3.4.4 The specific activity of the primary coolant shall be s 0.2 microcuries per gram DOSE EQUIVALENT 1 131. APPLICABILITY: MODE 1. MODES 2 and 3 with any main steam line not isolated. ACT10NS CONDITION REQUIRED ACT]DN COMPLETION TIME A. Primary coolant A.] Perform an isotopic Once per specific activity analysis for lodine. 4 hours
> 0.2 but s 4.0 pCi per gram DOSE &!C EQUIVALENT I 131.
A.2 Restore specific 48 hours activity to within limits. B. Required Actions and B.1 Perform an 4sotopic Once per associated Completion analysis for lodine. 4 hours Times of Condition A not met. &@ QB B.2 Isolate all main steam 12 hours lines. Primary coolant l ' specific activity I
> 4.0 #Ci per gram DOSE EQUlVALENT I 131.
1~ l l l' l l l ABWR 3.4 8 5/31/89 Amendment 9 16.7-8
'AB M nasinorr Standard Plant aA '
rh ' I.j 4 Specific Activity 3.4.4
$URVfflLANCE REDUIRfMENTS
$URVElLLANCE FREQUENCY
$R 3.4.4.1 .
.. -- . NOTE. . .. .. .-
Only required in MODE 1. Demonstr6te specific activity of primary 31 days
, coolant is s 0.2 #Cl per gram DOSE EQUlVALENT 3 131.
CROSS
REFERENCES:
None O ABWR 3.4 9 5/31/89 Amendment 9 16.7-9
MM 23A6100AP , Standard Plant hv^
- 16.74 3.4J Residual Heat Removal Shutdown O
RHR - Shutdown 3.4.5 3.4 REACTOR COOLANT SYSTEM 3.4.5 tentdual unat - val - shutdar; t LCO 3.4.5 Two Residual Heat Removal (RHR) shutdown cooling subsystems shall be OPERABLE. APPLICABILITY: MODE 3 with reactor steam do.ae pressure < 135 psig. MODE 4 with heat losses to ambient not sufficient to maintain average reactor cr;olant temperature s 200*F. ACTIONS CONDITION REQUIRED ACTION COMPLET!ON TIME A. One of the required A.) Restore the required 2 hours from RHR shutdown cooling RHR subsystem to discovery of subsystems inoperable. OPERABLE status, inoperable subsystem B. Required Action and B.1 Provide an alternate As soon as associated Completion method capable of -practicable i Time of Condition A decay heat removal not met, for each required but inoperable subsystem. QB MD No RHR shutdown cooling subsystem L2 Restore to at least two As soon as l OPERABLE. OPERABLE RHR subsystems, practicable i ABWR 3.4 10 5/31/89 O Amendment 9 16.7 10
MM 23A6100AP
- Standard Plant nu ^
l,-~) ' \_/ RHR Shutdown 3.4.5
$URVfftLANCE RfDUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify for the required RHR shutdown- 31 days cooling subsystem (s) each manual, power operated, or automatic valve in the flow QB path, not locked, sealed or otherwise secured in position, is in the correct 12 hours when position or is capable of being reactor steam manually aligned in the correct position, dome pressure is < 135 psig CROSS-REFERENCES TITLE NUMBER ECCS - Operating 3.5.1 s ECCS Shutdown 3.5.2
\ Residual Heat Removal Suppression Pool Cooling 3.6.2.3 ABWR 3.4-11 5/31/89 O Amendnunt 9 16.7 11
j MM 21A61ooAP l Standard Plant nev ^ 16,74 3A4 Reactor Coolant System Pnssh.v/ Temperature Units O , l3 RCS Pressure / Temperature Limits 3.4.6 3.4 REACTOR COOLANT SYSTEM 3.4.5 anneter coolant sustan Pressure /Temaeratura Limits 3 LCO 3.4.6 The Reactor Coolant System (RCS) temperature and reactor , i vessel pressure shall-be maintained within the s - Pressure / Temperature (PT) Limits. , APPLICABILITY: At all times. ACTIONS CONDITION REQUIRED ACTION COMPLET10N TIME A. --...... NOTE - - A.) Restore RCS temperature 30 minutes Required Actions A.] and reactor vessel and A.2 must be pressure to within the completed whenever PT Limits. this Condition is entered. E A.2 Determine RCS is 72 hours l Operation outside the' acceptable for continued PT Limits, operation. _ B. Required Actions and B.) Be in MODE 3. 12 hours ,'_ associated Completion Times of Condition A E not met. B.2 Be in MODE 4. 36 hours I ~. ABWR 3.4 12 5/31/89 O Amendment 9 16.7 12
. = . . _ . . _ _ . . . . _ _ _
- s. . .
l
'MM 23A6100AP
~ Standard Plant m.v. A l
RCS Pressure / Temperature Limits i 3.4.6 l SURVEILLANCE REDUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1- - ~ . ~ . . - ~ . . NOT E - ~ ~ ~ - - - Oni required during system heatup, coo down and ir, service-leak and hydrostatic testing. Verify RCS pressure and temperature are 30 minutes within the PT Limits. l SR 3.4.6.2 Verify RCS pressure and temperature are Once within
- i. within the PT Limit Curve criticality 15 minutes l limit, prior to l initial ,
l :- control rod ' ) withdrawal I for the
- h. I purpose of achieving V criticality i !
i l-SR 3.4.6.3 . - . - ~ . . ~ ~ . N OT E . ~ ~ ~ ~ - ~ ~ . . ~ ~ hall. . ~ . Only required in MODES 1, 2, 3, and 4 only reautred ! with reactor steam dome pressure gtten less than ' 2 25 psig, five RIPS are
...................................... in oneration q l
l Verify the difference between the bottom Once within { head coolant temperature and the reactor 15 minutes , pressure vessel coolant temperature is orier to each J l s 145'F. startue of a l reactor internal Dumo > (continued) l ABWR 3.4-13 5/31/89 l b' l Amendment 9 16.7 13
33A6100AP Standard Plant Rev A h RCS Pressure / Temperature Limits 3.4.6 j SURVEILLANCE REDUIREMENTS feontinued)
$URVEILLANCE FREQUENCY .
$R 3.4.6.4 Verify the reactor vessel flenge and he6d 12 hours when-l flange temperature are t 70'F. in MODE 4 with I reactor coolant system temperature ,
s 100'F l AllD 30 minutes when in MODE 4 with reactor coolant system l temperature s 80'F AtlD 30 minutes when ' tensioning the l reactor vessel head bolting studs i CROSS-REFERENCES l TITLE NUMBER Recirculation Pumps Operating 3.4.1 Inservice Leak and Hydrostatic Testing Oper6 {on 3.10.1 ASWR 3.4 14 5/31/89 O Aw=nt 9 3g7 34
)
MM 23A6100AP Standard Plant am ^ 16.7.7 3.4.7 Itaaeter Steam Dome Pawssure l Reactor Steam Dome Pressure 3.4.7 l 3.4 REACTOR COOLANT SYSTEM 3.4.7 Ranctor Steam Dnas Pressure LCO 3.4.7 The reactor steam done pressure shall be s 1040 psig. APPLICABILITY: MODES I and 2, except during anticipated transients. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME t A. Reactor steam dome A.) Reduce reactor steam 15 minutes pressure dome pressure to
> 1040 psig, s 1040 psig.
B. Required Action and B.) Be in MODE 3. 12 hours associated Coopletion Time of Condition A not met. O i EURVIILLANCE RfDUIRIMENTS l SURVE!LLANCE FREQUENCY l SR 3.4.7.1 Verify reactor steam dome pressure is 12 hours s 1040 psig. CROSS.
REFERENCES:
None ABWR 3.4 15 5/31/89 A. Anwedment 9 167 15
{ 23A6100AP Sinndard Plant w.A SECTION 16.8
.D b.
CONTENTS Section Dile East 16.8.1 33.1 ECCS-Oneration 16 & 1 16A.2 3.5.2 ECCS-Shutdown 16 & 7 O l~
.~ 16.8-ii Amendment 9
'ABWR zwinxv - 4 Standard Plant Pn A l 16.8 3.5 EMERGENCY CORE COOLING SYSTEMS (O,)
1 162.1 - 3J.1 ECCS-Operation
.l ECCS - Operating 3.5.1 ;
3.5 EMERGENCY CORE COOLING SYSTEMS 3.5.1 ECCS - Oneratino LCO 3.5.1 All ECCS injection subsystems of Divisions 1. 2 and 3 shall be OPERABLE, AltD 8 ADS valves shall be OPERABLE. APPLICABILITY: MODE 1. MODES 2 and 3 excent ADS and RCIC are not reouired to be OPERABLE with reactor steam done crassure
< 50 osio for ADS and < 150 osia for RCIC.
ACTIONS CONDITION REQUIRED ACTION CONPLET10N TINE L one in.iection A.) Restore inoperable sub- 80 days from ,. subsystem inoperable. system (s) to OPERABLE discovery of i status, inneerable , subsystem (s) o L Any two indeetion B.1 Restore at least one 14 da.vs inoperable subsystem subsystems inoperable. j
- j. to OPERABLE status, i AtID B.2 Restore the initial 60 days from inoperable subsystem to gigeovery of OPERABLE status. . nitial namorable subsvstem (continued) l L
l 3.5 1 lO ABWR 5/31/89 l Amendment 9 16A1 l'
gg 21A6100AP + Sandard Plant Rev A 9: ECCS . Operating 3.5.1 ACTIONS (continutd) CONDITION REQUIRED ACTION COMPLETION TIME L Any three in.iection L1 Restore at least one 72 hours subsystems inocerable, inocerable subsystem to OPERABLE status. M . M Restore at least two 14 days from inonerable subsystems discovery of to OPERABLE status. 111QE1 inocerable g subsystem W Restore all inonerable . 60 days from subsystems to discovery of OPERABif status, initial inonerable subsystem
~D. Required Actions and D.1 8e in MODE 3. 12 hours associated Completion Times of Conditions A.
B, or C not met. E D.2 Se in MODE 4. 36 hours l L One or two ADS valygg l U ...... 3Q1[............. : inonerable. Provisions of LC0 3.0.4 l are not anoliegblg. ! i Restore inoperable Prior to I ADS valve (s) to 11grlun.from l OPERABLE status, any outace of
>30 days duration j U
Prior to l startuo from next refuelina j El&Et I i (continued) : i i A8WR 3.5 2 5/31/89 Amendment 9 16.8-2
.ABWR z wioore Standard Plant nev ^ / : .( -
[CCS - Operatin 3.5. ACTIONS (continued) CONDITION REQUIPED ACTION COMPLET50N TIME-L Three ADS valves L1 Restore at least h1 180ARIldg. one of the inanerable valves to OPERABtf 1111E1 L Nore than three G.1 Be in Mode 3. 12 hours ADS valves inonerable. MD QB G.2 Reduce reactor steam 36 hours done pressure to Required Action and s 50 psig, associated Completion Time of Condition F not met.
-p U
l-('
- ABWR 3.5 3 5/31/89 16 & 3 AWt 9 l
C - . , , __
ABM n^6ioo^P Standard Plant Rev A , O ECCS . Operating 3.5.1 SURVEfttANEf REQUIRINENTS SURVIlLLANCE FRIOUENCY t-l ' SR 3.5.1.1 Demonstrate for each ECCS injection 31 days subsystem the system piping is filled l with water from the pump discharge t valve to the isolation valve. SR 3.5.1.2 ..................N0TE.................... LPFL subsystems may be considered CPERABli during alignment to and operation in the [' RHR shutdown cooling mode when below l' 335 psig. if capable of being manually realigned and not otherwise inoperable. Verify for each ECCS injection subsystem 31 days - each manual, power operated or automatic . valve in the flow path not locked, sealed or otherwise secured in position is in its correct position. SR 3.5.1.3 Verify Atmospheric Control System 31 days supply pressure to ADS valves 2 161 psig. ! SR 3.5.1.4 Demonstrate the following ECCS pumps develop According to It the specified flow rate against a system SR 3.0.5 head corresponding to the specified reactor pressure: QB SYSTEH HEAD 92 days CORRESPONDING 10 - - REACTOR 11312! FLOW RATE PRE 11URE OF LPFL 2 4200 gpm 2 40 psig HPCF 2 800 gpm 2 1177 psig (continued) L i I . A f ABWR 3.54 5/31/89
=
l Amendment 9 16 M
I MM 23A6100AP Standard Plant- Rev A ! ii e ECCS - Operatin 1.5.
$URVEILLANCE REQUIREMENTS (continued)
~
$URVEILLANCE FREQUENCY
$R 3.5.1.5 Demonstrate, with reactor pressure 92 days s 1177 psig, the RCIC pump can develop a flowrote a 800 gpm QB against a system head corresponding to a reactor pressure a 1177 psig. 12 hours when reactor-steam dome pressure is t 920 psig.
SR 3.5.3.6 Demonstrate, with reactor pressure 18 months s 165 psig, the RCIC pump can develop a flow 2 800 gpm DB against a system head corresponding to a reactor pressure 2165 psig. 12 hours
-f when reactor steam dome pressure is 2 150 psig. ,
................n0TE,..................
5"'53' L
!!!..!"!!!!!!"."!'.6*!**!"d'd:......
Perfonn a system functional test 18 months for each ECCS injection subsystem including simulated automatic actuation of the system throughout its emergency operating sequence, to verify each automatic valve in the flow path actuates to its correct position.
.................n0TE....................
SR 3.5.1.8 Valve actuation may be excluded. Perform a system functional test for ADS, 18 months including simulated automatic actuation, throughout its emergency operating sequence. (continued) ABWR 3.5-5 5/31/89 Anwedment 9
A MM 21A6100AP
- Standard Plant Rev ^
O s ECCS Operatin . 3.5. SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.5.1.g Demonstrate each ADS valve opens when 18 months manually actuated at steam dome pressure 2[ ] sig. AB 12 hours when reactor steam dome pressure t[ ]psig. CROSS REFERENCES TITLE NUMBER - ECC$ Actuation Instrumentation [ ] Residual Heat Removal Shutdown 3.4.5 Reactor Building Cooling Water / Reactor Building Service 3.7.1 Water Operating Residual Heat Removal Suppression Pool Cooling 3.6.2.3 i l ABWR 3.5-6 5/31/89 Amendment 9 16.8 4
ABWR. n^62oo^r Standard Plant Rev A 163.2 3J.2 ECCS. Shutdown
\_):
ECCS Shutdown 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS 3.5.2 ECCS - Shutdown l LCO 3.5.2 Two ECCS injection subsystems shall be OPERABLE, APPLitABILITY: MODE 4 MODE 5 except with the spent fuel pool gate removed and water level 2 23' over the top of the RPV
. flange.
ACTIONS CONDITION REQUIRED ACTION COMPLET10N TIME l, A. One of the required A.) Restore the required 4 hours from i ECCS subsystems ECCS subsystems to discovery of , inoperable. OPERABLE status. inoperable ! subsystem B. Required Action B.1 Suspend operations As soon as and associated Com- with a potential for practicable pletion Time of draining the reactor Condition A not met. vessel i I C. Both of the required C.1 Suspend operations As soon as ECCS subsystems with a potential for practicable inoperable. draining the reactor vessel. i C.2 Restore at least one 4 hours ECCS subsystem to OPERABLE status. (continued) ABWR 3.5-7 5/31/89 Amendaunt 9 16,8-7
ABWR 23^6=^r Standard Plant Rev ^ F i O i ECCS Shutdown
- . 3.5.2 l.
! AC110NS (continued) l CONDITION REQUIRED ACTION COMPLET10N TIME I D. Required Action C.2 D.) Suspend operations with As soon as l and associated a potential for drain- practicable !> Completion Time not ing the reactor ves'.el, met. M D.2 Ensure Secondary As soon as Containment is practicable OPERABLE. M - D.3 Ensure at least one As soon as SGTS subsystem is practicable OPERABLE. M D.4 Ensure at least one As soon as Secondary Containment practicable-Isolation Valve and associated actuation instrumentation is OPERABLE in each associated penetration i not isolated, l -. r l l [ I t 1 ABWR 3.5 8 5/31/89 't s Amendment 9 16 & 8
ABWR zwim Standard Plant nev ^ 1 g ag i U i ECCS - Shutdown 3.5.2 SURVEILLANCE Rif,0fRIMENTS SURVilLLANCE FREQUENCY SR 3.5.2.1 Verify for each required LPFL 12 hours subsystem the suppression pool water level is 214.63 ft, f
. 1 SR 3.5.2.2 Verify for each required HPCF system the:
A. CSP water level 2 22.8 ft. 12 hours E B. Suppression pool water 12 hours
, , , level is a 14.63 ft.
O SR 3.5.2.3 Demonstrate for each required ECCS 31 days injection subsystem the system piping is filled with water from the pump discharge valve to the isolation valve. SR 3.5.2.4 -....-- -- NOTE .. - ... -- - LPFL subsystems may be considered OPERABLE during alignment to and operation in the RHR shutdown cooling mode if capable of being manually realigned and not otherwise inoperable. Verify for each required ECCS injection 31 days subsystem, each manual, power operated or automatic valve in the flow path not locked, sealed, or otherwise secured in position is in its correct position. (continued) A8WR 3.5 9 5/31/89 Aswadmea 9 16.8-9
n _;; ^ u u ; MM . 23A6100AP Standard Plant _ nev A l 0 L l^ p (CCS $hutdown l- 3.5.2 SURVEILLANCEREQUIREM[NTS(continued) SURV[lLLANCE FR[QUINCY L - ! $R 3.5.2.5 Demonstrate each required ECCS pump According develops the specified flow rates against to SR 3.0.5 a system head corresponding to_the specified reactor pressure: QB SYSTEM HEAD- 92 days CORRESPONDING TO - - REACTOR 313]ig FLOW RATE PRf11 LIRE OF LPFL t 4200 gpm 2 40 psig l HPCF t 800 gpm i 1177 psig l l SR 3.5.2.6 - . . . - - NOT E . - - - . - - - - - Vessel injection may be excluded. ,, Perform a system functional test for 18 months each required ECCS injection subsystem including simulated automatic actuation of the system throughout its emergency operating sequence, to verify each auto-natic valve in the flow path actuates to its correct position. l t 1 ~ {' l l' ! A6iiii 3.5 10 5/31/89 l Ameadamat 9 16&l0
m j 23A6100AP Standard Plant Rev A
,. y ECCS Shutdown 3.5.2 CROSS REFERENCES TITLE huMBER ECCS Actuation Instrumentation [ ]
ECCS - Operating 3.5.1 Residual Heat' Removal Shutdown 3.4.5 Secondary Containment 3.6.4.1 Residual Heat Removal High Water level 3.9.8 Residual Heat Removal Low Water Level 3.9.9 Reactor Building Cooling Water / Reactor Building Service 3.7.2 Water - Shutdown Standby Gas Treatment System 3.6.4.3 Secondary Containment Isolation Valves 3.6.4.2 A Secondary Containment Isolation Actuation Instrumentation [ ] l l l ABWR 3.5-11 5/31/89 Amendment 9 I I
1 23A6100AP - Standard Plant am a SECTION 16.9 O,, - CONTENTS - Section Title East 16.9.1 3.6.1 W==w C=*='- = : S;2-'=2 16.9 1 5 16.9.1.1 3.6.1.1 Primary Containment 16.9-1
-16.9.1.2 ,
3.6.1.2 Containment Air Locks 16.9-3 16.9.13 3.6.13 Pdrimary Containment Pressure 16.97 .
-16.9.1.4 3.6.1.4 Drywell Average Air Temperature 16.9 8 16.9.1.5 3.6.1.5 Primary Containment and Pressure -
Isolation Valves ' 16.9 9 16.9.1.6 3.6.1.6 Wetwell.to-Drywell Vacuum Breakers 16.9-16 16.9.2 3.6.2 Sunnression Pool 16.9 19 16.9.2.1 3.6.2.1 Suppression Pool Average Temperature 16.9-19 16.9.2.2 3.6.2.2 Suppression Pool Water Level 16.9 22 ' p-3.6.23 Residual Heat Removal Suppression Pool 16.9.2 3 Cooling 16.9 24 16.9.2.4 3.6.2.4 Residual Heat Removal Wetwell Spray 16.9-27 16.93 3.63 Hydronen Control 16.9 29 16.93.1 3.63.1 Hydrogen Recombiner System 16.9-29 16.9 3.2 3.63.2 Primary Containment Oxygen Concentration 16.9-30 16.9A 3.6.4 Secondary Containment Systems 16.9 31 16.9.4.1 3.6.4.1 Secondary Containment 16.9-31 16.9.4.2 3.6.4.2 Secondary Containment Isolation Valves 16.9-34 16.9.4 3 3.6.43 Standby Gas Treatment System 16.9-37
) ;
16.9.ii Amendment 9
. .. . _. - . . _ . - - _. ... - .~ .. -. . . - . . - . .-
' MM zwtooAP l Remndard Plant h^ *
. 16.9 3.6 CONTAINMENT SYSTEMS
~
16.9.1 34.1 PHaary Costal=-t Systeams 16.9.1.1 34.1.1 Primary Costalament i Primary Contalsment 3.6.1.1 3.6 CONTAINMENT SYSTEMS 3.6.1.1 Primary tantainment LCO 3.6.1.) The Primary Containment shall be OPERABLE. APPLICABILITY: MODES 1. 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLET10N TIME A. Primary Containment- A.1 Restore Primary 1 hour inoperable. Containment to OPERABLE status. B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A &lQ O not met. B.2 Be in MODE 4. 36 hours L ABWR 3.6-1 5/31/89 O h t9 16.9-1
ABM ax6ioore ' Standard Plant nev ^ O i Primary Containment 3.6.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY-
~
SR 3.6.1.1.1 Demonstrate Drywell to Wetwell 18 months differential pressure does not decrease - - at a rate > [ ]incheswatergauge AND per minute tested over a [ ] minute period at an initial differential pressure - - NOTE - of ( )psid. Only required
. efter two consecutive tests fail I until two l- consecutive tests pass.
9 months SR 3.6.1.1.2 Perfom required Type A leak rate ---- NOTE ---- testing in accordance with 10 CFR 50 Provisions of Appendix J and approved exemptions. SR 3.0.2 are i r.ot l applicable. l In accordance with 10 CFR 50 Appendix J and approved = camptions. SR 3.6.1.1.3 Perfom required Type B leak rate testing -.- . NOTE - - except for Containment Air Locks in Provisions of accordance with 10 CFR 50 Appendix J and SR 3.0.2 are approved exemptions, not applicable. In accordance with 10 CFR 50 Appendix J and approved exemptions. CROSS
REFERENCES:
None ABWR 3.6-2 5/31/89 Amundment 9 16.9-2
MM 21A6100AP ReamAmed Plant - w A-l
)
16.9.1.2 34.1.2 Costatesment Air lacks ;
/m !
%)
"k
\
l l i: Containment Air Locks L 3.6.1.2 3.6 CONTAINMENT SYSTEMS-
- 3.6.1.2 containment Air Lackt
- LCO 3.6.1.2 Each Containment Air Lock shall be OPERABLE.
APPLICABILITY: NODES 1, 2, and 3. Af710NS
, CONDITION REQUIRED ACTION CONPLEil0N TIME A. One Containment Air .............N0TE.............
1 Lock door inoperable. Provisions of LCO 3.0.4 are not applicable. . gg .............................. One Containment Air A.1 ..........N0TE.......... t Lock door and Access allowed under. associated interlock administrative control, mechanism inoperable. ................-. ..... Close the OPERABLE door Ismediately (. in the affected air lock. E l- A.2 ...--.....N0TE.......... Access allowed under administrative control, not to exceed one hour cumulative per year. Lock the OPERABLE air 24 hours lock door. E A.3 ..........N0TE.......... Access allowed under administrative control, not to exceed one hour cumulative per year. Verify the OPERABLE deer Once per in the affected air lock 31 days is locked closed. (continued) ABWR 3.6 3 5/31/89 O Amendment 9 16.9-3
i
+
g -M M 23A6100AP Standard Plant ._ nev ^ O Containment Air Locks l.- 3.6.1.2 ACTIONS fcontinued) CONDITION REQUIRED ACTION COMPLETION TIME , B, Containment Air Lock NOTES - - interlock mechanism Access allowed under inoperable in one or administrative control, both Containment Air --- Locks. B.1 Close one OPERABLE air Ismediately lock door. AtID l B.2 Lock the OPERABLE air 24 hours lock door. AtiD B.3 Verify an OPERABLE door Once per in each affected air 31 days lock is locked close:- C. One Containment Air C.1 Close one door in the adiately O Lock inoperable for affected air lock. reasons other than Condition A and B. AltD C.2 Restore inoperable air 24 hours lock to OPERABLE status. (continued) ABWR 3.6-4 5/31/89 9 Amendment 9 16.94
f ABWR zwi=^r Standard Plant . nev ^
./ \
Containment Air Locks 3.6.1.2 ACTIDNS frontinued)
- CONDITION REQUIRED ACTION COMPLETJON TIMI D. Condition A exists. .............N0TES.... ...-..-
for both Containment 1. Provisions of LCO 3.0.4 Air Locks. are not applicable.
- 2. Access allowed under administrative control.
D.) Close one OPERABLE door Innediately in both air locks. AD i D.2 Lock one OPERABLE door 24 hours closed in each air lock. A@ Restore one air lock 7 days D.3 ! '. (O to OPERABLE status.
-Q
- l. E. One Containment Air E.1 Perform Required in accordance Lock inoperable Actions of Contidions A with for reasons of and B concurrently. Conditions A Condition A and other and B .
Containment Air Lock inoperable for reasons of Condition B. i F. Required Actions and F.1 Be in MODE 3. 12 hours l associated Completion l Times of Condition A. &@ B, C, D or E not met. F.2 Be in MODE 4. 36 hours QS Both Containment Air Locks inoperable for reasons other than
- - Condition D or E.
ABWR 3.6 5 5/31/89 Amendment 9 36,9-5
MM 2146100AP { se ndard Plant k" ^ i i e! ! Containment Air locks I' 3.6.1.2 suRvrittamer arnorarettNis , l
$URVI!LLANC[ FR(QU(NCY SR 3. 6.1. 2. '. Verify Containment Air Lock seal air flask 7 days h pressure t [ ]psig. ;
i SR 3.6.1.F.! Demonstrate Containment Air Lock interlock .....N0TE..... mechanism is OPERABLE. Only required ' If not performed i within - previous ' 6 months Prior to entry ' into Primary - Containment . when Primary ! Containment is deinerted. i
$R 3.6.1.2.3 Demonstrate Containment Air Lock seal 18 months pneumatic system pressure does not decay at a rate > [ ]psigin( )hoursfrom *
[ }psig.
$R 3.6.1.2.4 Perfore requNd Type B containment Air .....N0TE.....
Lock leak rate tdsting in accordance Provisions of with 10 CTR 50 Appendix J and approved SR 3.0.2 are esemptions. The acceptance criteria for not air lock testing are: appitcable.
- a. Overall air lock leakage rate is 5[ ]scfhwhentestedatPa. In accordance with 10 CFR $0
- b. For each door, the seal leakage Appendix J and rate is s [ ] scfh when the gap approved between the door seals is pressurited exemptions to Pa. ,
CROSS.REFERENC($: None i ASWR 3.6 6 $/31/89 O
% ,9 16.9 4
I
=>=
M.m.M m a..a pi... m- x AL9.1J 34.1J Pi4enary t'he peeseen i
/m J l
i' Primary Conteinmer.t Pressure 3.6.1.3 3.6 CONTAllmINT $YSTEMS j j 3.6.1.3 Primars tantatament Pressure l LC0 3.6.1.3 The Primary Contstament pressure shall be s 0.76 pstg. l APPLICABILITY: 910D[$ l. 2. and 3. l AtfloN1 CONDIT10N RIOUIRED ACTION COM/Lifl0NT![ A. Primary Containment A.) Restore pressure to I hour pressure within limits.
> 0.75 psig.
- 8. Retutred Attien and B.) De in H00[ 3. It hours associated Completion flee of Condition A MD :
not met. ( l.2 De in MODE 4 36 hours i Rimuf fit ANCf kf builtfMINTS I
$URV[lLLANC[ FRE0V[NCY
$R 3.6.1.3.1 Vertfy Primary Containment pressure is it hours s 0.75 pstg.
I l-Cho$$.R[F[R[NCES: None l ABWR 3.6 7 l/31/89 O Amendmeat 9 16.9 7 I
fl k f" gg 2AA61I2AP Reamulard Phrit **v ^ 16.9.1.4 34.1.4 Deywell Average Air Tesaper;.ture 1 ( ! t i Drywel? Average Air Temperature l 3.6.1.4 : I i j 3.6 CONTAIN4[NT $Y$1(M$ 3.6.1.4 Elrmus11 Averama F r femmaratura LCO 3.6.3.4 Drywell average air temperatJre shall be s J35'F. l APPLICA81LITY: IIDD[$ 1. 2, and 3. t aft 10NS CONDITION Rt0VIRED ACTION COMPLIT!DN TIM [ , A. Drywell average air A.1 Restore average air 8 hours temperature temperature to within
> 136'F. Itetts.
B. Required Action and 8.1 le in 140D[ 3. !! hours associated Completion Time of Condition A AtlD ; not met.
- B.2 le in ItbDI 4. 36 hours !
suavtitiawI arnultrMENTS SURVIlLLANC[ FRIOU(WCY SR 3.6.1.4.1 Verify Drywell average air temperature is 24 hours s 135'F. 1 CROS$.REFERENC[$t hone l ABWR 3.6 8 5/31/89 O r l i a mi , 16.94
MN M100AP Rear *A= 4 Plant m- A l [ 16.9.lJ 34.1J Prtanary Contatement and Pesseur, leetation Valves I l Primary Containment and Pressure Isolation valves l 3.6.1.6 3.6 CONTAllMENT $Y$itMS 3.6.1.$ prinnew t.antai m met and Peasaura laalatian vaivat i LCO 3.6.1.$ The Primo < :ntainment and Pressure Isolation Valves i' shall be :ndBil. APPLICABILITY: M00t$ 1. 2 and 3 MODis 4 and 5 when associated actuation instrumentation is ,
)
recuired to be OP!RABit per LCO [ Primary Containment ActuationInstrumentation). l I .........................n0TE......................... Conditions A through 0 may be concurrently applicable. Affloks CON 0lT10N REQUIRED ACTION COMPitil0N TIM [ I Ae One or more required O Primary Containment Isolation Valves or Pressure Isolation A.) ..........H0Tt.......... hot appitcable to those penetrations that have only one isolation Valves inoperable, valve. t Verify at least one Ismediately isolation valve is ' OPERABLE in each effected open i penetration, htQ A.t.1 Restore the inoperable 8 hours valve (s) to OPERABLE ststus. 98 (continued) , ABWR 3.5 9 5/31/89 l I 1 ae:9 16.9-9 l
- u. . _._. _ _ _ _ . _ _ ._ _ -. _ _ . _ _ . _ _ -__ _ _ _ _ _ _ _ _ _ _ ___ _ _ ___ _ _ _ __ _.--- -
ABWR 2 mime ! Standard Plant R" ^ i i !
- i r
i Primary Containment and Pressure Isolation Valves t 3.6.1.6 l l ' AtT1hN1 framtinued) i ! CONDITION REQUIRED ACTION COMPLET10N 11Mt l A. < continued > A.t.t.i ......... 01r......... , Only applicable to
- Primary Containment ;
l Isolation Valves. t f Isolate each affected 0 hours penetration by use of , et least one closed - and deactivated automatic valve, closed manual valve or bited flange, t G A.t.t.t ........N0Tt..........
- 1. Only app 11 able to ;
Pressure Isolation Valves.
- 2. Check valves used to satisfy this >
Required Action i must have been demonstrated to i seet $R 3.6.1.5.10. l Isolate the high 8 hours pressure pot-tton of { the effected system l from the low pressure i portion by use of at I least one closed manual or deacttrated outomatic or check valve. E l A.2.2.3 Verify each affected Once per penetration is 31 days isolated. (continued) I l i l ABWR 3.6 10 6/31/89 i gg,9 16410
1 MM 21A6100AP Standard Plant Ro A j Primary Containmer.t and Pressure isolation Valves . 3.6.1.6 MTIONS frontinued) i CONDITION Rt0VIRfD AC110N COMPL[ TION llMI B. Required Actions and B.) te in MODI 3. It hours associated Completion lines of Condition A &@ not met in MODI 1, 2. or 3. B.! De in MODE 4. 36 hours C. Reevired Actions and C.) Suspend CORI AL1[ RATIONS. Ismediately associated Completion Times of Condition A hot met in MODI 4 or 5. LO l l l l l l l ABWR 3.6 11 $/31/89 O Amendment 9 16911
MM 21A6t0MP
. RennAmrd Plant Rev ^
O-Primary Containment and Pressute Isolation Valves , 3.6.1.5 6 IWtElllL&Kt aroulktMENTS SURv[lLLANC[ FR[GU[NCY SR 3.6.1.5.1 .................N0Tl................. Valves and blind flanges in high radiation areas may be verified by use of adelnistrative controls. + Verify all manual valves and blind flanges 31 days which are located outside the Primary Containment and required to be closed during accident conditions are closed.
$R 3.6.1.6.2 Verify continuity of the traversing 31 days In Core Probe System isolation valve explosive charge.
$R 3.6.1.5.3 Verify all manual valves and blind flanges .....N0T[ ....
which are located inside the Primary Only retvired Containment and required to be closed if not during accident conditions are closed. Performed in the previous 92 days. Prior to entering ' WC 2 or 3 from MOD [ 4 if Primary i Containment I was de inerted while in MOD [ 4 (continued) l l ABWR 3.6 12 $/31/89 9 Asseadment 9 16.9 12
1 MM 21A6100AP Rinnalard Plant hv^ Primary Containment and Pressure Isolation Valves 1 3.6.1.5 ! RURVffLLAktf ktfRfikIMINTS ftantinuedt l
$URV[lLLANC[ FR[QUINCY !
u Sn 3.6.1.5.4 ................. noir................. ; MSIVs may be excluded. '
- - t Demonstrate isolation time of each According to automatic or power operated Primary $R 3.0.5 Containment Isolation Valve is within ,
limits. QB & 92 days . SR 3.6.1.5.5 Demonstrate full closure isolation time of According to ', each Main Steam Line isolation Valve is SR 3.0.5
. from 3 to 5 seconds.
g i 92 days 5R 3.6.1.5.6 Demonstrate each automatic Primary 18 months Containment isolation Valve actuates to its + isolation position on a simulated " autoritic isolation signal. SR 3.6.1.5.7 Demonstrate each reactor instrumentation 18 months line excess flow check valve actuates on a slaulated instrument line break to restrict flow to s I sph. SR 3.6.1.5.8 Remove and test the explosive squib from 18 months the Traversing In Core Probe System on a isolation valves. STAGG[ RID - TEST BA515 (continued) A8tlR 3.6 13 $/31/89 gg,, 9 16.9-13 __. ~ . , - - - - - - . . . ~ - - - - - - -
MM 21A6100AP Standard Plant Rev A ) I O> Primary Containment and Pressure Isolation Valves 3.6.1.5
- WRVflit&NEf AfhulkfMENT1 fenntinued)
$URV!!LLANCE FREQUENCY ,
i
$R 3.6.1.5.9 Demonstrate leakage rate through each 18 months :'
Main steam Line Isolation Valve for all four main steam lines is s 20 scf per - i hourwhentestedat[ ]plig.
$R 3.6.1.b.10 .................n0Tt.................
Only required in Moot 1 or 2. Demonstrate the leakage rate for each 18 months Reactor Coolant System Pressure Isolation Valve is s 0.5 gpe per nominal inch of volvi site up to a maximum of $ gpe.
$R 3.6.1.5.11 Demonstrate the combined leakage rate of 10 months '
1 gpm times the total number of Primary Containment Isolation Valves in hydrostatically tested lines which penetrate the Primary Containment, is not exceeded when these isolation valves are testedat[ ]psig.
$R 3.6.1.6.12 Perfore required Type C leak rate testing .....W0TE.....
, in accordance with 10 CFR 50 Appendix J Provisions of l and approved exemptions. $R 3.0.2 are not applicable. In accordance with 10 CFR 50 Appendix J and approved exemptions l ABWR 3,6 14 $/31/89 1 I AmnAmnt 9 l6'9'l4 1 l l i
ABM zu6ioorr ' Standard Plant an A t Primary Containment and Pressure Isolation Valves , 3.6.1.5 l tk M 5 RfffRfWff$ TITLE NUMBIR j Primary Containment Isolation Actuation Instrumentation [ } i l O o ABWR 3.6 15 5/31/89 Ameadawat 9 16.9-15 b
I I MM MlW l Remndard Plant h^
]
16.9.1 4 34.14 Weewell te Drywell Vocauss Bevokers l Wetwe11.to Dryw 11 YB i 3.6.1.6 ; 3.6 CONTAlleltNT SY$1(MS '
\
3.6.1.6 Matue11.ta brumm11 vacuum trankars l LCO 3.6.1.6 Seven Wetwell to Drywell vacuum treakers sha11 be DPtRABLE. I l E l All Wetwell to Drywell Vacuum Breakers shall be closed. l APPLICABILITY: MODIS 1. 2. and 3. AtTIDNS COND1110N RIOUIRED AC110N CiplPL(TION llM[ A. One Wetwell to Drywell A.) Close the open vacuum 8 hours from Vetuum Breaker open, breaker, distcvery of open vacuum breaker
- 8. One required 8.1 Restore the inoperable 7 days Wetwell to Drywell vacuum breaker to from Vacuum Breaker OPERABLt status, discovery of t inoperable for reasons inoperable ether than Condition A. vacuum breaker (continued) i l
ASWR 3.6 16 5/31/89 l Amendment 9 16.9 16
MM 2M6100AP ? Standard Plant Rev A i Wetwell to Drywel) VB , 3.6.1.6 ACTIONS fecetinued) CONDITION RIOUIRtD ACT!DN COMPLET!ON TIMI C. One Wetwell to Drywell C.) Close the open vacuum 8 hours from Vacuum Breaker open, breaker, discovery of ; open vacuum i breaker i AliD AtiD
, C.2 Restore the inoperable 7 days One reeutred vacuum breaker to from Wetwell to Drywell OP[RABLE status, discovery of '
Vacuum 8 eaker inoperable inoperable for reasons vacuum breaker other than Condition A. D. Required Actions and D,1 Be in MODE 3. !! Sours
- associated Completion Times of Condition A. &!iD '
$ or C not met. . (O j D.2 Be in MOD [ 4 36 hours i
l { t ABWR 3.6 17 5/31/89 O gg,9 16.417 l
QQ 21A61f0AP i hadard Plant Rn A O. t Wetwell to Drywell VB 3.6.1.6 1DRVt!Lt&Et krotilttMtWT1 ,
$URVIILLANCE FRt0VENCY !
$R 3.6.1.6.1 Verify all vacuum breakers are closed. 14 days
$R 3.6.1.6.2 Perform a functional test of each required 31 days vacuum breaker.
$R 3.6.1.6.3 Demonstrate the full opening setpoint of 18 months each required vacuus breaker is s 0.5 psid.
CROS$ RIFERENCl$t None A8WR 3.6 18 $/31/89 9 16.9 18 _ - . . . .-. ._ . - . - - . - - . ~ _ - . , . . . - . . . _ . . -
MM 3M6100AP I Remndard Plant _ mov ^
' '"'""""' i (D 6.9.2.1 1 SU.1 Seppreselos Pool Average Tempeenture '
I
$uppression Pool Average fesperature f 3.6.2.1 ;
3.6 CONTAINM[NT $Y$1[M$ 3.6.2.1 ""- r tainn Peel Averate Tammerature RO 3.8.f.1 Suppression Pool average temperature shall be: L g 96*f when f rearter - - r in > lt of RTP1 and testing I which adds heat to the Suppression Pool is not being performed. , L g 105'T when freacter r in > 11 ef RTP1 and testing which adds heat to the Suppression Pool is being performed. i L g Il0'T when f reneter newer in < 11 ef RTP1 APPLICABILITY: MODIS 1. 2. and 3. ', AtTIONS ' I CONDITION R[0UlR[D AC110N COMPLITION TIMI > L suppression Pool A.) Verify everage Once per hour i avera e temperature temperature s 110'f.
> 95' but g Il0*f. W -
AND A.2 Restore average 24 hours Ikear"er newer is > 11 temperature to Af.,11?,J . g 96*f. AllD Not performing testing which adds heat to the Suppression Pool. . B. Recuired Actions and L1 Reduce power to 11,,,11 !! hours associated Completion gLRIf,1 lines of Condition A , not met. (continued) A9WR 3.6 19 $/31/89 9 4919
9 QQ tu610ur Remndard Plant hv^ ! i O; I i i [ Suppression Pool Average Temperature l 3.6.2.1 ; ACT10k1 frantinued)
- CMDITION R(0UIRED ACTION COMPLCTION llMt i
C. Suppression Pool C.1 Suspend all testing ismediately everagetemperature which adds heat to the
> 105 F. Suppres:1on Pool.
AllD Performing testing which adds heat to the Suppres:1on Pool. D. Suppression Pool D.) Place the Reactor Mode lamediately everagetemperature $ witch in the Shutdown t
> 110 F but position, s It0'F.
AllD
~ .
D.! Verify everage temperature s !!0'F. Once per 30 minutes Ot i E. Suppression Pool [.1 Depressurize the reactor 12 hours everagetemperature vessel to . t
> 120 f. < 700 psig.
AllD [.2 Se in MODE 4. 36 hours 6 P A8WR 3.6 20 5/31/89 Amendment 9 16.9 20
i i MM ganadard Plant 23A610MP Rev A $ o ; Suppression Pool Average Temperature 3.6.2.1 tuRviittANet ktnutktMINTS
$URV[lLLANC[ FRt0VtNCY f SR 3.6.2.1.1 Verify the Suppression Pool everage 24 hours temperature is within the applicable limits.
$ minutes f when
. Performing tests which ,
add heat to the Suppression Pool i , Cross.AtF[R[NC[$1 None o 9 Amendewat 9 16.9-21
MN # ne =dard Plant *^ ; IL9JJ 34JJ Suppressies Peel Water Level Suppression Pool Water Level ! 3.6.2.2 ! 3.6 CONTAINMINT SYST[MS 3.6.2.2 * - malan Paal trater Level LC0 3.6.2.2 Suppression Pool water level shall be maintained from ! 22.97' to 23.29'. APPLICABILITY: MODIS 1. 2. and 3.
. AffloN1 '
CONDITION kt0VIRt0 ACTION ; COMPLtil0N TIMI A. Suppression Pool water A.! Restore water level to 4 hours level < 22.97'. within ilmits. DB ' Suppression Pool water ' level > 23.It'. B. Required Action and B.) Se in MODE 3. 12 hours associated Completion Time of Condition A &!Q not met. B.! Be in MODI 4. 36 hours l I h b
- ., ,,,1, ,
g
MM zwlme Standard Plant Rev A e~
\ )
w/ Suppression Pool Water Level 3.6.t.2 1DRVIitt&WEI RfD'J1RIMINTS
$URV[l[LANC[ FRIQU(NCY
$R 3.6.2.2.1 Verify $uppression Pool water level is 24 hours from !!.97' to 23.29'.
Ek01$.RfftRENEt1 TITLE NUMB (R [CC$ Operating 3.6.1 O I l l l i i A8Wk 3.6 23 $/31/89 Amendment 9 16.9-23 l
l i MMme.a.ma pi..e Su6100AP l mi ;
$6.92J 3423 ResWest Heat Rauwel Se@ Peel W l
l ( I I ; l RHR luppression Pool Cooling ! ' 3.6.2.3 s 3.6 CONTAllMINT $YST[M5 I 3.6.t.3 hansdual annat ; al =~ maw pani t.maitna I f 1 i 4C0 3.6.2.3 The Residual Heat Removal (RHR) Suppression Pool Cooling l system sha'.1 1.4 GIRABLE. ! p APPLICABILITY: 9100($ l. 2. and 3.
- l ACT1013 ,
t CONDITION Rt00! RED AtTION CollPLETION 11M[ l A. One RHR Suppression A.! . . . . . . . . . 100T [ . . . . . . . . . . Pool Cooling subsystem ! inoperable, Provisions of LCO 3.0.. ! are not applicable. ' l' ........................ t ! Restore inoperable 30,,,ggy,3 from ; i subsystem to OPtRABLt tiscovery of l status. inoperable l Subsystem l L 330 RHR Suppression B.! Restore at least one 1,,6&u from Pool Cooling subsystems ; of the inoperable sub. discovery ' i inoperable, systems to OPtRABLE i of second i status, inoperable subsystem. B.! Restore the initial 3Ligy,,1from inoperable subsystem to ciscovery of OPERABLt status. initial inoperable subsystem l (continued) 1 i 4
- - 3..... ,,,1,B, e.
1 w a , - - , - . . - . - - . - , ...-m-- .<- - - ,--...-.-e. -
- , - .e--- . . . - - w-e-.- , - . . . . - - , - - - . - - . . , , . - - - . -
)
i MM 23A610aAP mndard Plant arv ^ , RHR Suppression Pool tooling 3.6.t.) 4 ACTIONS (Continued) CONDITION R(QVlRLD ACTION COMPittl0N 11MI L 111 i wee kHR tuntret. L} Rettore at least ene 8 hours ! ston knol toolino agb1111mm to DPikAhtf subtistemt innneratie, m. ' 8
- L2 ket". ore to at laatt twe 7 dart from
- DPlitAlit t subsvitent, discovery of i atLand inonerable
, subsystem i L} Restore the initial 30 dart from i able subsystem to discovery of ti status. Initial inanerable !
MM ' D. Required Actions and D.1 Be in MODI* 3. It hours associated Completion 11mes of Condition A, B ' or C not met. D.! Be in MODI 4 36 hours F l [ i ABWR 3.6 25 5/31/89 b v Assendewat 9 16.9 25
> MM me mdard Pir.nt el i
RHR suppression Pool Cooling 3.6.2.4 l RURVf1tt&Wff ktDulkfMINTE l
$URVIILLANCI IRl0V!NCY
$R 3.6.t.3.1 Verify each manual, automatic. er power 31 days operated valvc in the flow path not locked. :
sealed or otherwise st;ured in position, is in its correct position er can be aligned to its correct position.
$R 3.6.t.3.2 Demonstrate each RHR pump develops a flow 92 days rate t 4200 spe through the associated RNR heat eschenger while operating in QB the Suppression Pool tooling mode.
According to SR 3.0.6
- - t 5
tRoss RIFtRENtt$ 11TLt huMBIR . Residual Heat Removal $ystem . Shutdown 3.4.6 ECC$ . Opertting 3.5.1 Residual Heat Removal Wetwe)) $ pray 3.6.2.4 ASWR 3.5 26 $/31/89 O
, 16.9 26
., .. - . . .. .~ .. . . . . . - . . - _ _ - . . . . - . . - - -
MM 2M6100AP ReamA.ed Plant h^ ; BL9.1A 34.2A Residual Heat Rasawal Womett Spray t i RNR Wetwell Spray 3.6.2.4 , 3.6 CMTAIN4(NT $YST[M5 3.6.2.4 M1 ' nat A'-- ral untum11 Ameny f LCO 3.6.2.4 The Residual Heat Removal (RNR) Wetwell spray system shall i be OPIRABl[. , APPLICABILIVY: MODIS 1, 2, and 3.
. AfY1DN1 CONDITION REQUIRED ACT10N COMPl[T10N TIMI A. One RHR Wetwell $pra3 A.! ..........N0TE..........
subsystem inoperable. Provisions of LC0 3.0.4 are not applicable. ' Restore inoperable 7 days from
' subsystem to OPERABLt discovery of ,
status, inoperable subsystem
- 8. Both RHR Wetwell Spray 8.1 Restore at least one 8 hours subsystems inoperable. subsystem to OP[RABLt status.
AlQ 8.2 Restore the initial 7 days from inoperable subsystem to discovery of OPERABLE status, initial inoperable subsystes l C. Required Actions and C.) Se in MODE 3. It hours associated Completion Times of Condition A &lQ or 8 not met. C.2 Se in MODE 4. 36 hours , ABWR 3.6 27 5/31/89 l
\
l m .9 16.427
l i MM n46100AP 1 S!*mmAmed Plant a- A ! O l AHR Wete:11 $ pray : ( l 3.6.2.4 !' EURVf f t LANCf Lf 0DIRfttINTS SURV[lLLANCE FR[0U[NLY , t
$R 3.6.t.4.3 Verify each manual, automatic, or power 31 days !
i operated valve in the flew path is not ; 1 locked, sealed or otherwise secured in ; position. is in its correct position or can be aligned to its correct position. : i
. t 1
ER 3.8.2 a.f Demonstrate each RMR pump develops a flow 92 days rate 2 4200 Spa through the associated RHR heat exchtager while operating in the cambined *--- einian pan 1 f.nelinn/ Wetwell inrav Glade. ' l l I M t[FERfutit TITLE NUMBER Residual Heat Removal . $hutdown 3.4.5 ECCS Operating 3.5.1 Residual Heat Removal Suppression Pool Cooling 3.6.2.3 t i i 1 1 m g9 t&9 2 1
MN asA6100AP i R**dard Plant u.4 l 4 10J 34.3 Mydroges C6atrol 10.3.1 17.11 Hydrepe ReceaWeer System i Hydrogen Receabiner System 3.6.3.1 3.6 CONTAINM[NT SY$TIMS l 3.6.3.1 thdrmaan haesabiner antam - LCO 3.6.3.1 The Primary Containment Hydrogen Recombiner Systee shall be OPERABLt. APPLICABILITY: MDDt5 1 and 2.
. ACTfDks f
CONDITION R(0UIRED ACTION COMPLETION TIME 11Alifll F O ASWR 3.6 29 $/31/89 Ameedesat 9 16.9-29
_ . ~ _ _ _ - _ . . . _ _ t MM ReamAmed Plant 2M6100AP a- ^ I S&9.3J 343.3 Primary re e Osygem Cemeestration Primary Containment Osygen Concentration l 3.6.3.2 l 3.6 CONTAINM(NT $YSTEMS , 3.6.3.2 Primarr tantainment enraan tantantration i LCO 3.6.3.2 The Primary Centainment osygen concentration shall be s 45 by volume. i APPLICABIL11Y: NODI 1 with THERMAL POW [R t llt of RTP for > 72 hours. N0DI 1 when time remaining with THilulAL POWER > 16% of ATP
, Prior to the nort scheduled shutdown is > 72 hours. ,
ACTION 1 CONDITION REQUIRED AC110N COMPLETION 11Mt , f A. Primary Containment A.) Restore erygen is hours oxygen concentration concentration to
> 48 by volume. g 45 by volume.
B. Required Action and B.! Reduce THERMAL POW [R to 8 hours ! associated Completion 5 lit t,f RTP. l Time of Condition A . not met, e I- SURVF1LLAktI eraulttntuT1
$URV[lLLANC[ FR[OVENCY SR 3.6.3.2.1 Verify Primary Containment oxygen 7 days concentration is s 48 by volume.
CROSS.REFERENC[$: None ASWR 3.6 30 5/31/89 Anseement 9 16.9 30
. - - . , _ . _ . _ _ _ _ _ . . . _ . . . . _ ~ - _ , . _ _ . _ . . . _ . .
MM 2nA61onAP r Remndard Plaat ma A (' SUA SAA L ^^~,Cee % t SUA.1 34A.1 ~ _-i Centedement Secondary Containment f 3.6.4.1 ; I 3.6 CONTAllMINT SY$ftMS ' 3.6.4.1 Secondary Centainment
)
1 LC0 3.6.4.) The Secondary Containment shall be OP!4ABLE. APPLitA81LITY: 9100!$ 1, t, and 3. When handling trradiated fuel in the Secondary Containment, During CORI ALTERATION $. ' During operations with a potential for draining the reacter
, ves6el (0PORVs). ,
.........................m01t.........................
Conditions A 9 and C may be concurrently appilceble. EI.lalli._ - CONDITION RIQUIRED Atil0N COMPLCTION 11Mt A. Secondary Containment A.) Restore Secondary 4 hours i
, inoperable in Containment to OPERABLE
\ MODI 1. 2. or 3. status.
B. Required Action and 8.1 De in MODE ). It hours associated Completion I flee of Condition A MD
- not met.
B.! Se in MODI 4. 36 hours (continued) m ABWR 3.ti.31 5/31/89 Amendmeet 9 16331 l
MM 23A6100AP Stand ed Plarit Rev A O' Secondary Containment i 3.6.4.1 ACT10h1 fenntinued) CODIDITION R(QUIR[0 ACT10N COMPLETION TIM [ C. Containment Col ......... 100T[.......... ,
$ttender{a inoperab when Provisions of LC0 3.0.3 handling irradiated are not applicable.
fuel in the Secondary Containment. during , CORE ALTERATIONS, or $uspend handling of IIpediately
. during OPDRVs. irradiated fuel 1. the Secondary Containment.
M C.2 Suspend CORI ALT [ RATIONS. Ismediately E C.3 Suspend OPORVs. As soon as practicable O', l ABWR 3.6 32 $/31/89 O Amendment 9 16.9-32
I l MM 23A6100AP ! Standard Plant u^ r ! ( ! I l Secondary Conteinment 3.6.4.1 suRytittANtf RfDUlttM[WTS
$URV[lLLANCE IRIOUtNCY
$R 3.6.4.1.1 Verify Secondary Containment vacuum is 24 hours '
t 0.2$ inches of vacuum water gauge. SR 3.6.4.1.2 Verify all Reactor Butiding equipment 31 days hatches and blowout panels are closed i
. and sealed.
$R 3.6.4.1.3 Verify each Reactor Building access door 31 days is closed, except for routine entry and exit.
l [
$R 3.6.4.1.4 Demonstrate one $tandby Cas Treatment 18 months
- Subsystem will draw down the Secondary on a $TAGGERID Containment to ) 0.25 inches of TE$1 BA$1$
l r vacuumwatergaugeins[ ] seconds. L l
$R 3.6.4.1.$ Demonstrate one $tandby Cas Treatment 18 months Subsystem can maintain n [ inches on a $1AGGIRED of vacuum water gauge in the Secon]dary TE$1 BA$l$
Containment for one hour at a flow rate
< 10$0 cfe.
tRoss REFERENCES llTLE NUMBIR Secondary Containment Isolation Actuation Instrumentation [ ] Secondary Containment Isolation Yalves 3.6.4.2 Standby Cas Treatment System 3.6.4.3 l ABWR 3.6 33 $/31/89 O Amendment 9 16.9-33
- MM 31A6100AP StandardPlant au ^ - '
16.9AJ 34AJ 8eeendary Ceaselement Isolaties Velves i 1 Secondary Containment 1 solation Valves 3.6.4.2 3.6 CONTAINMENT SYSTEM 5 3.6.4.2 Ameandarr tantainmar.t lanlation Valves I LC0 3.6.4.2 The Secondary Contat:wnt Isolation Valves shall be OPERABLE. APPLICABILITY: NDDES 1, 2, and 3 When handling irradiated fuel ir,t h Secondary Containment, I During CORE ALTERATIONS, I
- During operations with a potential for draining the reactor vessel (0PDRVs).
l
.........................N0TE.........................
Conditions A, B and C may be concurrently applicable. ET10NS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 ..........N0TE.......... Secondary Containment Not applicable to those Isolation Valves penetrations that have inoperable, only one isolation valve. Verify at least one lamedistely isolation valve is OPERABLE in each affected open penetration. AllD A.t.) Restore the inoperable 8 hours valve (s) to OPERABLE status. l QB (continued) 1 6 ASWR 3.6 34 5/31/89 i t-Anwedment 9 16.9-34 l
. . - - , ,. .. - . - . . . - .--. . . , . . , - . - ~ . - - - . - - . . . - . - . -- -- - -
a
~M M 23A6100AP mndard Plant nev A j
-q.
f^')S Secondary Containment Isolation Valves 3.6.4.2 ] ACTIONS fcontinued) CONDITION- REQUlRED ACTION COMPLETION TIMI A. (continued) A.2.2.) Isolate each affected B hours penetration by use of at least one closed t' ' and deactivated automatic valve, closed manual valve 1 or blind flange. M A.2.2.2 Verify each affected Once per penetration is 31 days isolated. B. Required Actions and B.1 Be in MODE 3. It hours associated Completion Times of Condition A E not met in MODE 1. C' 2, or 3. B.2 Be in MODE 4 16 hours C. Required Actions and. f.1 - NOTE - - associated Completion Provisions of LCO 3.0.3 Times of Condition A are not applicable, not met when hand 11ng - - - - - 1rradiated fuel in the Secondary Suspend handlinD of Isuwdiately Containment, during irradiated fuel in the CORI ALTERATIONS, Secondary Containment. or during OPORVs. M C.2 Suspend CORE ALTERATIONS. .lsmediately M C.3 Suspend OPORVs. As soon as practicable A8WR 3.6 35 5/31/89 O Anwodment 9 IN
1 i Q 23A6100AP l Sundard Plant Rn A i l 9 Secondary Containment isolation Valves 3.6.4.2 RURVfittANCf kfDUIREMINTS SURV[lLLANCE FREQU[NCY 5R 3.6.4.2.1 Verify all Reactor Building blind flanges 31 days or rupture discs which are required to be closed during accident conditions are closed. SR 3.6.4.2.2 Demonstrate the isolation time of each According to Secondary Containment Isolation Valve SR 3.0.5 is within limits. E g2 days 'l i l l i
. SR 3.6.4.2.3 Demonstrate each Secondary Containment 18 ocnths i isolation Valve actuates to its isolaticn l position on a simulated automatic iso'estion signal. .i I l
l- -l i CROSS REFERENCES TITLE NUMBER - Secondary Containment Isolation Actuation Instrumentation [ ] l 1 ABE 3.5 36 5/31/89 Anwadawat 9 16.9 36 l
}
.3
. i M a pi..e M..u.
m 21A6100AP mi e 16.9A.3 34AJ Standby Gas Tmetuneet System
$tandby Gas Treatment System 3.6.4.3 >
3.6 CONTAINMENT SYSTEMS 3.6.4.3 *+ " v aan Tree w n system LCD 3.6.4.3- The Standby Gas Treatment System ($ CTS) shall be OPERABLE. + APPLICABILITY: N00E5 1, 2, and 3 When handling irradiated fuel in the Secondary Containment, During CORE ALTERAT10NS,
- During operations with a potential for draining the reactor vessel (OPORVs).
.........................N0TE.........................
Conditions A through E may be concurrently applicat.le. ACTIONS CONDITION REQUIRED ACTION COMPLEi10N TIME ' L One SGTS subsystem A.) Restore inoperable . 7 days from t
\ inoperable ist subsystem to OPERABLE discovery of reasons other than status. inoperable raadition C. subsystem L-l L Both SGTS subsystems B.1 Restore at least one inoperable in M00E 1, 4 hours subsystem to OPERABLE
- 2. or 3 for reasont status, other than condition C.
AND B.2 Restore the initial 7 days from inoperable subsystem discovery of to OPERABLE status, initial inoperable subsystem l l L The SCTS c - n filter L1 Reatore the e - n 4 hours train inonerable in fi' ter train to MODE 1. 2. or 3 OPI: Rant f status. (continued) l O ABWR 3.6-37 5/31/89
,9 16.9-37
-- ._._1---------------
gg i
- i. 21A6100AP
! l Riandard Plant ""^ 1 1 i L i t . . 9: l Standby Gas Treatment System 3.6.4.3 ACTIONS (cc,ntinued) COL'0!T10N REQUIRED ACTION COMPLETION TINE D. Required Actions and 0.1 Be in MODE 3. 12 hours associated Completion ) l l Times of Condition A B E' or C not met in MODE 1, 2. or 3. D.2 Be in MODE 4 36 hours L Required Action and E.1 -
-- . NOTE -
associated Completion Provisions of LCO 3.0.3 Time of Condition A are not applicable. i not met when handling - - - - - I 1 irradiated fuel in the Secondary Containment. Suspend handling of Ismediately during CORE ALTERAT10NS, irradiated fuel in the or during OPDRVs. Secondary Containment. 1 4 l E E \ Both SGTS subsystems E.2 Suspend CORE ALTERATIONS. Immediately inoperable when handling irradiated E i fuel in the Secondary l Containment, during E.3 Suspend OPDRVs. As soon as CORE ALTERATIONS, l practicable i or during OPDRVs. ' l The e - n filter train inonerable when handlino irradiated Juel in the Secondarv lenta' nannt durino 10Rt htTERA' IONS,
- or durino Ol0RVs. I 1
I l I 4 l ABWR 3.6-38 5/31/89 ! O 1 I gg 9 16.9 38 )
i i MM 2M6100AP 1 I se.ndard Plant hv^ f t 1 Standby Gas Treatment System ! 3.6.4.3 SURVEltLANCf RfDUIRIMINTS
$URVEILLANCE FREQUENCY SR 3.6.4.3.1 Demonstrate each $GTS subsystem operates 31 days e,.n_a with flow through the coeumon filter train 5"AGGEliD TEST-for 2 10 hours with heaters on. &W.
, SR 3.6.4.3.2 Demonstrate < 0.05% per.etration of the 18 months SGTS HEPA filters by a DOP test at a system flow rate of 1050 to 1200 cfa. M Once within 7 days after painting,
' fire or chemical release in filter service area
\
M Prior to declaring subsystem OPERABLE after each-complete or partial replacement of filter (continued) I l A8WR 3.6 39 5/31/89 Aswadment 9 IN
.g g asA6ioarr Standard Plant ""^
L l-9 i Standby Gas Treatment System 3.6.4.3
- suavtlLLANtf REQUIRIMENTS fenntinued) i-
, $URVtlLLANCE FREQUINCY-
$R 3.6.4.3.3 Demonstrate < 0.061 bypass leakage 18 months through the $GTS charcoal absorber section ,
by a halogenated hydrocarbon test at a M i system flew rate of 1050 to 1200 cfm. l' Once within 7 days ! , after painting, fire or chemical l- release in filter l. l service area
- 7. M ,
l-Prior to F declaring subsystem OPERABLE l- after each complete or ' partial replacement of adsorber bank. i (continued) 1 f .j 1 A8WR 3.6 40 5/31/89 1 t9 16.940
T MM 23A6100AP Standard Plant *^ h- Standby Gas Treatment System 3.6.4.3-t EURVEILLANCE RE0ulkEMENTS ftontinued) 1 SURVEILLANCE FRE0VENCY SR 3.6.4.3.4 .................N0TE................. Analysis to be completed within 31 days of sampling. Remove and perform a laboratory analysis 22,Qhours of a SGTS charcoP adsorber sample for of charcoal
. methyl todide pene. etion. adsorber operation M
18 months M Once within 7 days after p) 3 v painting, fire or chemical release in filter service area SR 3.6.4.3.5 Demonstrate < [ ] inches water gauge 18 months pressure drop across the combined SGTS HEPA filters and charcoal adsorber banks at a flow rate of 1050 to 1200 cfm. SR 3.6.4.3.6 Perform a system functional test of each 18 months SGTS subsystem which demonstrates filter train startup and isolation dampers opening upon receipt of a simulated automatic initiation signal. SR 3.o.4.3.7 Demonstrate each SGTS heater dissipates 18 months from( ) to [ ]kw. A8WR 3.6-41 5/31/89 A-d-st 9 16M1 I 1
l: _ .. 1 "MM i 23A6100AP 1, Riandard Plant u^ O Standby Gas Treatment System 3.6.4.3 CROS$. REFERENCES TITLE NUMBER
~
l' Secondary Containment Isolation Actuation Instrumentation [ ] l 1 l 1 i I I l l' ABWR 3.6 42 5/31/89 l Amendment 9 16.N2 O
23A6100AP - Standard Plant ' Rev A l SECTION 16.10 is CONTENTS l Section Title P_ age 16.10.1 3.7.1 Reactor Building Cool'un W=8-(RCW)/n etor Building Service Water (RSW) System-Operating 16.10 1 16.10.2 3.7.2 Reactor Bulldlug Cooling Water (RCW)/ Reactor , Building Service Water (RSW) System Shutdown 16.10 5 16.103 3.73 Control Room HVAC Emenyency Recirculation System 16.10-8 16.10A 7,7A Main Copdenser Oftmas 16.10-14 , i i w O 16.10-ii Amendment 9
. . _ , - . - - _ . . _ . . . ._ _. ~ . . _ _ . , . _ , . - . - - . .
gg zwimr Styndard Plant h^ fs t.4.10 3.7 PLANT SYSTEMS l 16.10.1 3.7.1 Reactor Bulldlag Cooling Water (RCW)/kenctor Building Service Water (RSW) Systema.Oneratina RCW/RSW System . Operating 3.7.1 3.7 PLANT SYSTEMS LM Reactor Bu11dino Coolinn Water IRCW1/ Reactor Buildino Service Water (RSW) System - Doeratino LCO 3.7.1 Tie Divisions I. if. and 111 Reactor Buildino Coolina Water i tCWl/ Reactor Bui'dino Service Water (R$WI systems shall be 0)ERABLE. APPLICABILITY: MODES 1, 2, and 3.
.......................ggJ1...........................
Conditions A throuch D may be concurrent 1v ano11 cable. I f ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME L .....jg}1.........- M Declare affected system Immediate1v fach of the three or conoonent inocerable. divisions may be in
, this Condition con-u '""*"P.::......... "
OneltCW oumo and/or one tSW oumo and/or-M ....-.... 3Q11.......... Provisions of LCO 3.0.4 *
- one tCW/RSW heat are not ano11 cable, j exchanner inonerable ..... ..................
in the same division, lest 3re the inonerable 60 days ICW/ISW componentis) from dis. to 0)ERABLE status, covery of inonerable conDonent. L One RTL/R5W Division M Declare affected system lamediately inonenble for or conoonent inonerable, IphsonL ether than Condition A, E M ltestore the inonerable 7 days from itCW/RSW Division to discovery of OPERABLE status, inonerable RCW/RSW
)1 vision (Continued)
ABWR 3.7-1 5/31/89 Amendment 9 IN
..'g g 23A6100AP Standard Plant *^ '
ei 1: RCW/R$W System - Operating 3.7.1 ACTIONS (continued) COMPLETION CONDITION REQUIRED ACTION TIME L Two RCW/R$W Divisions L1 Declare affected syst e lamandiat el y inonerable for connonent inonerable, reasons other tha, Condition A. QR l LL1 testore one inonerable 24 hours tCW/RSW Division to GPERABLE status. AND W Restore the ini tial 7 days from inonerable RCW/ tSW discovery of Division to 0FE LABLE initial i status. >nerable ' ICW/R$W '
)ivision.
D. Required Actions and D.1 Be in MODE 3. 12 hours O> associated Completion !- -Times of Condition A : AND l B or C not met. D.2 Be in MODE 4. 36 hours L l l. i i-l
- ABWR 3.7 2 5/31/89 Amendmcot 9 l .
t
?
MM 21A6100AP Standard Plant Rev ^ . ,y RCW/RSW System - Operatin 3.7. SURVEILLANCE pEDU1REMENTS 1 SURVEILLANCE FREQUENCY
~
SR 3.7.1.1 24 hours verify][ ultimate 2[ inches. heat sink) water level is ,
~
SR 3.7.1.2 Verify the [ ultimate heat sink) water 24 hours temperature is s 95'F
~
SR 3.7.1.3 Verify water level in the RSW pump well of 14 days the intake structure is 2 [ ] feet. 8!iQ 12 hours when pump well level is s [ } A - feet. V SR 3.7.1.4 Verify for each required RCW/RSW subsystem 31 days each manual, power operated or automatic valve in RCW/RSW flow paths servicing safety related systems or components not locked, sealed or otherwise secured in position is in its correct position. Sp 3.7.1.5 Demonstrate each fultimate heat sink 31 days active componenti operates for 2 15 minutes. SR 3.7.1.6 Perform a system functional test for 18 *mnths each required RCW/R$W subsystem including simulated automatic safety related functionino of the system. I ABWR 3.7 3 5/31/89 k
%J Amendment 9 16.10-3
h
. MM '
23A6100AP Siendard Pimnt Rev.A
.O I'
' RCW/RSW System + Operatin 3.7.
CROSS RUtRINCES TITLE NUMBER
.RHR Shutdown- 3.4.5 -
ECCS Operating 3.5.1 AC Sources Operating 3.8.1 i I i 1 01 l l I l l l 4 l l l l I ABWR 3.7-4 5/31/89 l l l Amendment 9 16.10 4 l
- ABWR z w ia m Standard Plant Rev.A q' 16.10.2 -3.7.2 Reactor Building Cooling Water (RCW)/ Reactor Buildlag Service Water (RSW) b/; Systein Sheldown RCW/R$W System Shutdown 3.7.2 3.7 PLANT $YSTEMS 3.7.2 Rameter Buildino toolina Water (RCW)/ Reactor Buildino service Water f R$W1 Svitem - Shutdown LCO 3.7.2 The Divisions 1. II, and 111 Reactor Building Cooling Water .
(RCW)/ Reactor Building Service Water (R$W) systems associated with systems and components required to be OPERABLE shall be OPERABLE. APPLICABILITY: MODES 4 and 5. When handling irradiated fuel in the secondary containment. ACTIONS COMPLET10N CONDITION REQUIRED ACTION TIME A. Required RCW/etSW A.) Declare affected system immediately Division inoperable, or component inoperable. (
%/
ABWR 3.7 5 5/18/89 O Amtndment 9 16.10 5
n LABWk 32^62ao^r ge.ndard Plant Rev ^ O RCW/R$W System . Shutdown l 3.7.2 SURVEltLANCE pf0VfREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify the [ ultimate heat sink) water level .te hours is 2 [ ] inches. SR 3.7.2.2 Verify the [ ultimate heat sink) water temperature is s 95'F 24 hours SR 3.7.2.3 Verify water level in the RSW pump well of '4 c;ys intake structure is 2 [ ] feet. AND 12 hours . when pump well level iss( ) feet. SR 3.7.2.4 Verify for each required RCW/RSW subsystem 31 days each manual, power operated or automatic valve in RCW/R$W flow paths servicing safety related systems or components not locked, sealed or otherwise secured in position is in its correct position. SR 3.7.2.5 Demonstrate each fultimate heat sink 31 days active comoonenti operates 215 minutes. SR 3.7.2.6 Perform a system functional test for 18 months each required RCW/RSW subsystem, including simulated automatic safety related functionino of the tvstem. t A8WR 3.7 6 5/18/89 Amendment 9 16.146
MM . 23A6100AP Standard Plant na A Q). RCW/R$W System . Shutdown 3.7.2 CRO$5 RFffPENCES TITLE NUMBER , RHR
- Shutdown 3.4.5 (CCS Shutdown 3.5.2 AC Sources . Shutdown 3.8,2 RHR . High Water level 3,9,8 RHR Low Water level 3.9,9 i:
i ABWR 3.7 7 5/18/89 l Amendment 9 16.147 l
, - -. ~.. - . . . .
1 l MM 21A6100AP Simadard Plant w^ 16.103 3.7J Control Room HVAC Emergency Recirculation System CR HVAC [R System 3.7.3 i
. 3.7 PLANT SYSTEMS 3.1.3 f. antral " WWat Emereener Racirculation Ersten LCO 3.7.3 The Control Room HVAC Emergency Recirculation (CRHER) system shall be OPERABLE. 1 APPLICABILITY: MODES 1, 2, 3, 4, and 5 When handling irradiated fuel in the secondary containment.
.........................N0Ti.........................
Conditions A, B and C may be concurrently applicable. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME I 1 A. One CRHER subsystem A.] Restore the inoperable 7 days from l inoperable. subsystem to OPERABLE discovery of ' status. inoperable ,."- subsystem i I 1 B. Required Action and B.! Be in MODE 3. 12 hours associated Completion ) j Time of Condition A &!Q not met in MODE 1, 2, or 3. B.2 Be in MODE 4 36 hours Both CRHER subsystems inoperable in MODE 1, 2, or 3. l l l (continued) l l l l ABWR 3.7.B 5/1B/89 O Amendment 9 16.1M
MM 23A6100AP Etandard Plant bA [) CR NVAC ER System 3.7.3 ACTIONS fcontinued) CONDITION REQUIRED ACT10N COMPLETION TIME C, Required Action and C.) Place OPERABLE subsystem Imediately associated Completion in the isolation mode of Time of Condition A operation. ' not met in MODE 4, 5, or when handling QB 1rradiated fuel in the secondary containment. C.2.1 Suspend CORE ALTERATIONS. Imediately QB E Both CRHER subsystems C.2.2 - - - NOTE - - inoperable in MODE 4 Provisions of LCO 3.0.3 5 or when handling are not applicable. irradiated fuel in the - - - secondary containment. Suspend handling of Imediately irradiated fuel in the secondary containment. O C.2.3 Suspend operations with As soon as a potential for draining practicable the reactor vessel. ABWR 3.7 g 5/18/89 O Amendment 9 16.10 9
I t ABM11 nasiao^r : St*=dard Plant hA CR HVAC [R $ystem 3.7.3
$URVElttANCE RfDUIREMENTS SURVEILLANCE FREQUENCY
$R 3.7.3.1 Demonstrate subsystem operates with flow 31 days through the HEPA filters and charcoal adsorbers for t 10 hours with heaters on.
SR 3.7 '.2 Demonstrate < 0.05% penetration of the 18 months HEPA filters by a DOP test at a system flowrateof[ ] to [ ]cfm. E Once within 7 days after painting, fire, or chemical release in filter service area E Prior to declaring subsystem OPERABLE after each complete or partial replacement of a filter (continued) ABWR 3.7 10 5/18/89 O Amendment 9 16.10 20
5 l ABWR zwi=^r Standard Plant Rev A b CR HVAC [R System
. 3.7.3
$UPVffttANCE RfDUIREMENTS feentinued)
SURV[lLLANtt FREQU(NCY
$R 3,7.3.3 Demonstrate < 0.05% bypass leakage 18 months through the adsorber section by a '
halogenated hydrocarbon test at a system E flow rate of [ ] to [ }cim. Once within 7 days after i painting, fire, or L l chemical release in filter service area M Prior to declaring subsystem O OPERABLE
'Q after each complete or partial replacement of an adsorber bank (continued) t ABWR 3.7 11 5/18/89 O
Amendment 9
1 I gg nAsicoAr Standard Plant ""^- i 1 e l CR HVAC ER System ll ) 1 3.7.3 l )
$URVflLLANCE Rf0UIREMENTS fcontinued)
SURVE!LLANCE FREQUENCY SR 3.7.3.4 ... ..
. * . N0TE......- - - -
Analysis must be completed within 31 days of sampling. i l Remove and perfors; a laboratory analysis 720 hours
= of a charcoal adsorber sample for methyl of charcoal iodide, adsorber
' operation 1 1, M 18 months i M Once within 7 days , after painting, fire, or chemical release in filter service area SR.3.7.3.5 Demonstrate s [ inches water gauge 18 months pressuredropacross]thecombinedHEPA filters and charcoal adsorber banks at a systemflowrateof[ ] to [ ]cfm. SR 3.7.3.6 Demonstrate heaters dis from 18 months [ ]to[ ]kw. (continued) 1 l. i A8WR 3.7-12 5/18/89 l: Amendment 9 16.1412
4 7 iA'B M us. loos, . Standartl Plant - m,s. x.
'1 CR HVAC ER System 3.7.3 SURVEILLANCE PEQUIREMENTS fcontinued)
$URVEILLANCE FREQUENCY SR 3.7.3.7 Demonstrate each sut' system automatically 18 months switches to the isoittion mode of '
ot'eration on receipt of an actuation signal. CROSS
REFERENCES:
None ABWR 3.7-13 5/18/89 O Amendment 9 16.10-13
-._ . _. ~ . _ .- . . . _ ~ . - _ . - .-. _ .. -- -
F 23A6100AP Standard Plant. Rev.A 16.10A 3.7A Mais Condenser Ofiges Main Condenser Offgas 3.7.4 3.7 PLANT SYSTEMS 3.7.4 nain e w anser effans LC0 3.7.4 The gross gama radioactivity rate of the noble gases measured at offgas recombiner effluent shall be s 390 millicuries /second, after 30 minutes decay. APPLICABILITY: MODE 1, MODES 2 and 3 with any main steam line not isolated. l -- ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Gross gamma A.1 Restore gross gama 72 hours radioactivity rate radioactivity rate to s i > 390 millicuries / within limits. l second after 30 minutes decay. B. Required Action and B.1 Isolate all main steam 12 hours associated Completion lines. Time of Condition A not met. i. l l ABWR 3.7 14 5/18/89 l l O Amendment 9 16.10 14
- MM 23A6100AP u
Standard Plant *A l D '
\
I 1 Main Condenser Offgas-3.7.4 SURVflLLANCf RfDUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Perform an isotopic analysis of a ..... NOTE ..- representative sample of gases taken at Only required the offgas recombiner effluent, after Steam Jet Air Ejector is in operation. 31 days MD l Once within 4 hours l after a t 50% increase in
.the nominal steady state
\
fission gas release after
//
( factoring out increases due to changes in THERMAL POWER level CROSS.
REFERENCES:
None l ABWR 3.7-15 5/18/89 O Amendment 9 16.10-15
l 23A6100AP ' '
- Standard Plant wx c SECTION 16.11 CONTENTS !
Sec11on Iltle East 1 i t 16.11.1 32.1 A.C. Sources Onerallna 16.11 1 16.11.2 3J.2 A.C. Sources Shutdown 16.11 20 16.11.3 3.83 Diesel Fuel Oil 16.11 22 16.11.4 ~ 3.8.4 D.C. Sources Operatinqt 16.11 26 16.11.5 3.8.5 D.C. Sources-Shutdown 16.11 31 $ 16.11.6 3.8.6 Batterv Electrolvte 16.11 33 16.11.7 3.8.7 A.C. Power Distribution Systems-Oneratinn 16.11 37 16.11.8 3.8.R D.C. Power Distribution Systems-Operating 16.11-40 16.11.9 3.8.9 A.C. and D.C. Power Distribution Systems-Shutdown 16.11 42 i O i l 1 l0 16.11 11 l Amendment 9
I' MM 21A6100AP
. Sundard Plant *^ j 1
16.11 3.8 ELECTRICAL POWER SYSTEMS l
~O 16.11.1 3.8.1 A.C. Sources Operating !
j 1 1 1 A.C. Sources - Operatin 3.8. 348 ELECTRICAL POWER SYSTEMS c 3.8.1 A.C. Sauccan - Baaratina
- LCO 3.8.1 The following A.C. Electrical Power Sources shall be OPERABLE:
A. Two physically independent circuits between the off site transmission network and the on site Class li distribution system, E L lhtgg independent diesel generators. APPLICABILITY: MODES 1, 2, and 3. ' ACTfDNS CONDITION REQUIRED ACTION COMPLETION TIME A. One of the required A.1 Perform SR 3.8.1.2 .I hour 3 circuits from off site for remaining re utred inoperable, circuits from of -site. [a. E Once per 8 hours thereafter M A.2 Restore required 14 days from inoperable circuit frca discovery of off-site to OPERABLE inoperable status, circuit from off site i B. One diesel generator B.1 Perform SR 3.8.1.2 for I hour
- i. inoperable, required circuits from j- off site. E Once per
- 8 hours thereafter M
(continued) l ABWR 3.8-1 5/31/89 O Amendment 9 16.11 1
- n ABWR zw =
Standard Plant m- A . 9' ! A.C. Sources . Operatin 3.8. AITIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME k B. (continued) B.2 Verify required redundant 2 hours systems, subsystems, trains, devices and components that depend-on the OPERABLE diesel generators as a source of emergency power are l; OPERABLE. A!Q B.3 ..........N0TES..........
- 1. Only required if diesel generator
! became inoperable for any causa other than preplanned maintenance or testing.
- 2. Not required if condition which caused the diesel l- generator to be declared inoperable does not impact OPERABILITY of OPERABLE diesel generators.
Perform SR 3.8.1.4 for 24 hours. OPERABLE diesel generators. AliD B.4 Restore inoperable 14 days from diesel generator to discovery of OPERABLE status. inoperable diesel generator (continued) ABWR 3.8 2 5/31/89 O Amendment 9 16.11 2
k M h. 23A6100AP SimmA.rd Plant an A I A.P $ources . Operatin 3.2. ACTION 1 frantinued) , CONDITION R(QUIRID 3CT10N COMPitT10N TIMC C. One of the required C.) Perfors $R 3.8.1 2 1 hour circuits from off. site for remaining re4uired inoperable, circuits from off.11te. M M Once per 8 hours One diesel generator I thereafter inoperable. # M C.2 Verify required redundant 2 hours systees, subsystees, trains, devices and components that depend on the OPERABL[ diesel generators as a source of emergency power are OPERABLE, C ., ...........i m ...........
- 1. Only required if diesel generator became inoperable for any cause other than preplanned maintenance or testing.
- 2. Not reevired if condition which caused the diesel generator to be declared inoperable does not impact OPERABILITY of OPERABLE diesel generators.
Perform $R 3.8.1.4 for 8 hours OPERABLE diesel generators. M (continued) ABWR 3.8 3 5/31/89 Amendmen' 9 16.11 3
l MM 2146t0QAP c %dard Plant h^ ! O A.C. Sources . Operating i 3.8.1 i ACTIONS feentinued) I CONDITION REQUIRED AC110N - COMPLETION TIME l i i C. (continued) C.4.1 Restore required 72 hours from J inoperable circuit from discovery of : off. site to OP[RABLt second inoper. i status. able component I l 98 i C 4.2 Restore thoperable 72 hours from i diesel generator to discovery of ' OP(RABL[ status, second inoper. able component j E < l C.$ Restore initis) 14 days from i inoperable component discovery of l (diesel generator or irottal i circuit from off. site) inoperable l to OPERABLE status, component , D. Two diesel generators inoperable, 0.1 Perform $R 3.8.1.2 for required circuits 1 hour G: - j from off site. E ; Once oer B hours there. . Afit.E E M Verify reouired r6dundant I hours systems. subsystems, trains. devices and conLon. ns that decend on the 0)!RABLE J1esel eenerator as a source ' of eneroenew onwer are
- OPERAkti.
E y .......... g .......... j L Only reouired if diesel eeneratorfs) became inonerable for any cause other g n oreolanned maintenance or testinc. (continued) ABWR 3.8 4 $/31/89 Amendaient 9 16.11 4
- . . . . - . . . - . . . - . . . ~ . . _ - - - - . - - . . . . .-
b 23A610a0 Rimadard Plant u.i . j s i
\v) '
i i i A.C. Sources . Operating i 3.8.1 l Afi10N1 frontinued) f CONDITION Rt0V!RtD ACTION C04PLtT10N TIM [ [
.... NOTES (cont.)........ !
D. (continued) L het renulmd if candition which ' raantad 19w p diggel manerathri tI 10 be declaree < nanarable , dota not ' anatt OP :RABI 1"Y of DP :RARL ' diesel " ethertinr. Perform SR 1.8.1.2 for 24 hours OPERAkt f dtate) generator. E M Pestore one of the 72 hours from O reoutred inanerable discovery of , ( diete) annarators to second inoper. OPERAnti status. able diesel M generator ' D.5 Restore the initial 14 days from inoperable diesel discovery of generator to OPERA 8Lt initial ; status. inoperable diesel generator L One of the reouired L1 Perform SR 3.8.1 J h circuits from off tite for remainino reouired inonerable, c1rtutt from off tite. M M Once oer 8 hours Two diesel annerators thereafter inanerable. 6 , (continued) ABWR 3.8 5 5/31/89 O Amendment 9 16.11 4
t I
. MM 23A6100AP l'
Simassed Piment nev ^ 9! ! I t A.C. Sources . Operatin l 3.8. ACTIONS feontinued) CONDITj0N REQUIRED ACTION COMPltTION TIME ! L ftentinued) L2 Verify renuired redundant 2 hours ' systems. subsystems, trains. devices and cannonents that denend t on the OPERABt[. diesel nonerator as a source of eneroency namer are OPERABtt. 8!sD l L.3 ..........g gs.......... L Only reautred if diesel oeneratorft) I became inocerable for any cause other i than oreolann&d maintenante or i testina, , L Not renuired if condition which caused the diesel , neneratores) to be ' deciared ' nacerable does lot ' monet OP LRAslL1"y of OP :RA3tt diesel generator. > Egrfom SR 3.B.1.4 for B hours OPERAEtt diesel Generator. AND W Restore reavired 12 hours ! inonerable circuit from off. site to OPERABtt status. Q3 (continued) ABWR 3.8 6 5/31/89 Amendment 9 16.11-6 {
t MM 2M610aAP
*"^ {
_handard Plant i
} ,m t (v) i l
A.C. Sources I Operating 3.8.1 j ACTIONS f ecati%ed) l CONDITION REDVIRED ACTION COMPLETION TIME i I L feentinued) W Restore one of the 12 hours reaut red 1--- rable ; diete) annarators to i OPtkAbtf stains. E Q Restore at least two of 72 hours from i the recuired inocerabi discovery.cf e - ----is (diesel aan.g tarond inocer. erators or circuit from able comoonent , off.siten to OPikABtf
- status.
$ 1 L i- Restore 'nitial 14 days from i nocerab' a comoonent discovery of
- fdiesel eenerator or MM circuit from off site) inocerable O
\
to OPikARfI status, taanonent [' i F. Both of the required F.! Restore one of the 72 heues circuits from off site required circuits from inoperable. off site to OPERABLE status. O r F.! Restore the initial 14 days from inoperable circuit from discovery of off site to OPERABLE initial i status, inoperable circuit from off site L Both of the recuired (1 Verify recuired redundant Lhgyn circuits from off site systems. subsystems, incoerable, trains. devices and connonents that decend
@ on the DPtkABLE diese) emnerators as a One diesel nenerator source of eneroenev incoerable, oower are OPERABLE.
M (continued) ' ABWR 3.8 7 5/31/B9 i hndant 9 16.11 7 i
. , . . . - , ~ . , , - , - - , . , , -
. . . - ~ -. -
I Ciasadarel Plant nev ^ : A.C. Sources . Operating ! 3.8.1 - ! ACT10N1 frontinuedl ' CONDITION REQUIRED ACTION COMPLET10N TIME l L ieoriinued1 (2 .......... g g}.......... ! L Oniv reautred if ! sietel menarator > Atame inanarable > Jor any cauan other
. "han ;7;;l-----
maintenante or ' 1811189 L het reoutred if i g wittien which . cauted the diesel
- annarator to be declared inocer4ble penerators. ;
Perform 1R 3.t.1.4 for h OPERARtf diesel nonerators. AtiD '. W Rettore one of the re. 24 hours nu< red ta - rable tir. cu' 11 fror off. site to OPLRAkt f status. . 98 i LL2 Restore the recuired 24 heurt , inocerable diesel Gen. erator to OPERABtf M. M L1 Restore at least two of 72 hours from the reoutred inoberable discoverv of ; connenents (diesel aan. second inocer. erator or circuits from able component off-sitel to OPERABtf
- status.
M (continued) ABWR 3.8 8 5/31/89 , Anwadment 9 16.11 4
MM 21A6100AP Remndard Plant nev ^ i (
^
i i e A.C. Sources . Operating ; 3.8.1 ! ACT1DNS f reetinued) i CONDIT!DN REOUIRED ACT10N COMPLETIDN TIM [ i g, frontinued) 1} Restore initial 14 dart fret incoerable e---'**nt discovery of idiesel nonerator er initial . circuit from off titel inont rable to OPEkAktf status, component ' i H. Three diesel centrators H1 Perfere SR 3.R.1.2 1 hggg inocerable. Jor reouired circuits i from off.111e. AND B2 Restore one of the re- 2 hours cuired inocerab' e diesel nonerators to OPERABtt 1141E1
&HD g
H} Rettore to at least L Lhpura from I two OPERARif diesel d' scovery of i penerators, ascend inerer. ' able diesel eenerater ' AND B3 Restere to three 14 davt from GPERABtf (intel noner. discovery of . 11kII. initial inocer. able diesel nenerator
- 1. Required Actions and 1.1 Be in MODE 3. 12 hours associated Completion Times of Condition A. &HD B C, D, E, F, G, or H not met. 1.2 Be in MODE 4. 36 hours ABWR 3.8 9 5/31/89 l
Amendment 9 16.11 9 >
MM NIN : Re=ndard Plant h^ ! i e!! 1 A.C. Sources . Operating 3.8.1 SURVt1LLANCE REDUIPEMENTS
$URV[lLLANCI FREQUENCY SR 3.9.i.1 Verify that the load sheddine and 12 hours tenuencino lonic f tiil1 auto test system is operating and is not indicating a faulted condition, t
$R 3.8.1.2 Verify correct breaker alignment and 7 days indicated power availability far each of the required circuits from off. site.
$R 3.8.1.3 Demonstrate thBt the response of the LS$L 31 days is within [the design criteria) by a manual ;
insertion of each of the following test signals: A. LOCA.
- 8. Bus undervoltage.
C. Bus undervoltage followed by LOCA. D. LOCA followed by bus undervoltage.
$R 3.8.1.a ..................N0TtS....................
- 1. Performance of $R 3.8.1.8 satisfies this surveillance.
- f. All diesel generator starts may be preceded by an engine prelube period, warmup procedures and grad.
ually loaded as recommended by the manufacturer. Demonstrate diesel generator achieves the According to following voltage and frequency: Tabic 3.8.1 1 A. Voltage of 6210 to 7590 volts. B. Frequency of 58.8 to 61.2 Hz. (continued) ABWR 3.8-10 5/31/89 O Amendment 9 16.11 10
y _ __ _ MIM l Standard Plant Rn. A n ( k ' A.C. Sources . Operating i 3.8.1 i
$URVfittANff Rf0DIRIMINTS fenetinued)
$URV(!LLANC[ FR[QU[NCY sR 3.8.1.s ................. 01rs....................
- 1. Performance of SR 3.8.1.9 satisfies this surveillance. -
1
- 2. All diesel generator starts may be '
preceded by an engine prelube period, warmup procedures and gradually loaded as recommended by the manufacturer.
- 3. Transtent loads outside of this band do not invalidate this test. '
Demonstrate diesel generators operate for According to 2 60 minutes at: Table 3.8.1 1
-- A. Lead of [ ] to }kwfor DieselGeneratorA.[
[ 8. Lead of [ ] to ]kwfor ', DieselGenerator8.[ C. Leadof[ ] to [ ]kwfor Diesel Generator C. SR 3.8.1.6 Verify diesel generators are aligned to According to provide standby power to their associated Table 3.8.1 1 - emergency buses. SR 3.8.1.7 Verify pressure in each reautred air According to F startreceiveris1[ ] psip. Table 3.8.1 1 (continued) l A8WR 3.8 11 i 5/31/E9 v g9 16.11 11
MM 21A6100AP l Standaed Plant ma ^ i i l O. I A.C. Sources . Operatin 3.8. EURVfftLANCE Rf001RIMENTS ftomiinued) l
$URV[lLLANC[ FR(QUINCY
' $R 3.8.1.8 ................N0T[....................... All diesel generator starts may be preceded by an engine prelube period. Demonstrate diesel generator starts from 184 days standby condition and achieves the following voltage and frequency in i 13 seconds: 1 A. Voltage of 6210 to 7690 volts.
- 8. Frequency of 58.8 to 61.2 Hz.
I 1 SR 3.8.1.9 ..............N0Tl$......................
- 3. All diesel generator starts may be
- _ preceded by an engine prelube period.
-)
i
- 2. Transient loads outside of this band I do not invalidate this test. I i
Demonstrate diesel generators are loaded 184 days 1- in 1 60 seconds after synchronization l i with the bus and operate for 1 60 minutes ) at: i l A. Loadof[ ] to [ ] kw for ) Diesel Generator A.
]
i- 8. Loadof[ ] to [ ] kw for l Diesel Generator 8. ; C. Loadof[ ] to [ ]kwfor Diesel Generator C. l l l I $R 3.8.1.10 Demonstrate manual transfer of unit power 18 months ! i supply from the normal circuit to alternate ' circuit for each of the required circuits from off site. 1 (continued) l l l ABWR 3.8-12 5/31/89 Amendment 9 16.11 12
i 21A610aAP Standard Plant an. x O A.C. Sources Operatin' > 3.8. .
}ki/f1LLANf f LfDUIRf a8fNT5 frontieued)
SURVIILLANCE FRt0VINCY SR 3.8.1.11 Demonstrate the automatic load secuence 18 months timer is OPIRABLE with the interval between each load block is [within i 10% ofitsdesigninterval.) 1R 3.8.1.12 Demonstrate diesel generator rejects the 10 months following lead while maintaining voltage of 3744 to 4576 volts and frequency of 58.8 to 61.2 Hz: : A. 2[ ]kwforDieselGeneratorA. .
- 8. 2[ ]kwforDieselGenerator8.
,, C. 2[ ]kwforDieselGeneratorC.
i SR 3.8.1.13 Demonstrate diesel generator does not 18 months tripandvoltageof<[ maintained during an3 following a lued
]voltsis rejection of [ ] to [ ] km.
(continued) . ABWR 3.8 13 5/31/89 Amendment 9 16.11 13
MM 2146100AP f
.Sfamulmed Plant hA '
l 9t! i l A.C. Sources Operating ; 3.8.1 ' SUPVffttANCf Rf0DIREMfh*TS feentinued)
$URV(ILLANC[ FR(QU(N0Y !
5R 3.8.1.14 .................N07E.............. ....... i All diesel generator starts may be preceded ' by an engine prelube period and warmup procedures recommended by the manufacturer. t
$1mulate a loss of off site pos t (LOSP) 18 months !
within 5 sinutes of shutting down the diesel after the diesel has operated 21 hour at1[ ] kw or the diesel is at nomal operating temperature, and demonstrate: A. Deenergitation of emergency buses, j B. Load shedding from emergency buses.
.. C. Diesel generator auto. starts, and:
- 1. Energires permanently connected loads in i 13 seconds.
- 2. Energites auto. connected shutdown loads through the load sequencing logic. >
- 3. Supplies pemanently and auto. connected loads for >~ 5 minutes.
- 4. Achieves and maintains steady state voltage of 6210 to 7590 volts.
- 5. Achieves and maintains steady state frequency of $8.8 to 61.2 Hz.
(continued) b Atwo 3.8 14 5/31/89 0. Ameedownt 9 16.11 14
.~ _ ._ _ , . . _ . _ . _ , . _ . . , . . . _ . - . _ . _ . . . . _ . . _ -___
I QQ 2M6100AP Standard Plant *^ L
-s g .
s_- t [ i A.C. Sources . Operating 3.8.1 10RVffitANff kf0VfREMINT1 fenetinuedi
$URV[lLLANCl FR(QU(NCY i
$R 3.8.1.lb ................N0TE*................ ....
All diesel pe.*erator starts may be preceded l' by an engine prelube period aad warmup procedures reconner.ded by the manufacturer. ; Demonstrate on an ECCS actuation test 18 months signal, without loss of off. site power, the diesel generator auto starts, and: A. Achieves and maintains voltage of 6210 to 7590 volts in < 13 seconds s'ter auto start and during , remainder of the test.
"' B. Achieves and maintains frequency of
[58.8)to[61.2)Hzin<[13) seconds after auto start an3 during \% remainder of the test. C. Operates for 1 5 minutes, t (continued) t ABWR 3.8 15 , $/31/89 k Anwndment 9 16.11 15
-. _. . .. . . . . ~ -- - - - - -
MM Riandergi Piarit 21M100AP nev ^ 9l , 1 A.C. Sources . Operating 3.8.1 EUPVf1LLAhti RIDUIREMENT$ frontinued) l
$URV[!LLANC[ FR(QU[NCY
$R 3.8.1.16 .................. 980Tl................ .
All diesel generator starts may be preceded i by an engine prelub* period and warmup ' procedures receamended by the manufacturer. l Demonstrete on a LOSP actuation test 18 months signal in conjunction with an ECC$ i actuation test signal: I l A. Deenergiration of emergency buses. 1
- 8. Load shedding from emergency buses.
)
C. Diesel generator auto starts, and:
- 1. (nergizes permanently connected l loads in < 13 seconds. l
- 2. Energizes auto connected emergency loads through load i sequencing logic, l
- 3. Supplies permanently and auto connected loads for > ~
5 minutes. j
- 4. Achieves and maintains steady state voltage of 6210 to 7590 volts.
- 5. Achieves and maintains steady state frequency of 58.8 to 61.2 Hz.
I
$R 3.8.1.17 Demonstrate automatic diesel generator 18 months trips are automatically bypassed upon an ECCS actuation test signal except:
A. Engine overspeed
- 8. Generator differential current
[C. Low lube oil pressure) (contenued) ASWR 3.8 16 5/31/89 Amendment 9 16.11 16 i 4
i ABM ussi arr Se==dard Plant bA
/m,.
1 I i A.C. Sourcet Operating : 3.8.1 l l
$URVflLLANff Rf00lkfMfWT1 fenntinuedi i I
$URV(!LLANtt FRIOUINCY ]
SR 3.8.1.18 ...............N0TE........................ 1 Transient loads outside of this band do not invalidate this test. Demonstrate the diesel generators operate 18 months < for g 24 hours: A. Loadedfrom[ ] to [ ]kw. ,
- 8. With voltage of 6210 to 76g0 volts in < 13 seconds after start signal an3 during remainder of the test. j t
C. With frecuency of 58.8 to 61.2 I H2 in < 13 seconds after start signal ' and during remainder of the test. t
$R 3.8.1.1g Verify auto.conaected loads s [ ] kw. 18 months
$R 3.8.1.20 Demonstrate each diesel generator: 18 months A. Manually synchront2es with off site power source, upon a sieuJ14ted restoration of eff. site power, while the diesel generator is loaded with e its emergency loads.
- 8. Transfers loads to the off site power source.
C. Returns to standby status.
$R 3.8.1.21 Demonstrate that with diesel generator 18 months operating in the test mode and connected to its bus, an (CCS actuation test signal overrides the test mode by returning the diesel generator to standby operation.
(continued) ABWR 3.8 17 $/31/89 s Amendment 9 16.11 17
?
MM 2W100AP hrd Plant an A i G:? P A.C. Sources Operatin ! 3.8. -
$URVfitLAN!f RfDUIRIMfh'TS feoetinued) i
$URV[lLLANC[ FR[OVENCY l l
SR 3.8.1.22 Inspect diesel generators in accordance 18 months with procedures peepared in conjunction with manufacturer's recomunendations. 5R 3.8.1.23 Demonstrate when started simultaneously. 10 years all three diesel penerators achieve the i following voltage and frequency in
< 13 seconds:
A. Voltage of 6210 to 7590 volts.
- 8. Frequency of 58.8 to 61.2 H2.
tho11.REFERIN!f5 TITLE NUM8[R A.C. Sources Shutdown 3.8.2 Diesel Fuel Oil 3.8.3 l l l l L l l l l i
)
I ABWR 3.8 18 S/31/89 O Anw Mmeat 9 16.11 18
- ~ ._ ~ - . _ - - _ - - . - - _ - . _ - _
i 21A6100AP l Reandmed Plant mm. A j l V A.C. Sources Operating 3.8.1 j Table 3.f.1 1 (Page 1 of 1) l Dietal f.anarater Test ichedule NUM8[R Of fAILUR[$ IN NUMBER OF FAILURE $ IN LAST 20 VAllD TESTS I') LAST 100 VALID TESTS I*I FR[0V[NCY j l 11 14 31 days 22 35 7 days (b) i
....................................NQTl$.....................................
(a) Criteria for determining number of valid failures and number of valid , tests shall be in accordance with Regulatory Position C.E.e of Regulatory , Guide 1.108. but determined on a per diesel generator basis. Failure , to satisfactorily complete SR 3.8.3.1 through SR 3.8.3.9 does not constitute a valid failure and does not require en increased frequency O of testing.
? :
(b) The associated Freovency shall be maintained until 7 consecutive r failure free tests have been performed and the number of valid failures in the last 20 valid tests has been reduced to one. For the purposes of determining the reevired Frequency, the previous test failure count may be reduced to zero if a complete diesel overhaul to like new condi tion is completed. This diesel overhaul, including apprcpriate post. , maintenance operation and testing, shall be specifically approved by the : manufacturer and acceptable diesel reliability must be demonstrated. The reliability criterion shall be the successful completion of 14 consecutive j tests. Tel. of these tests may be slow starts (in accordance with the t routine SR 3.8.1.4 and SR 3.8.1.5) and four tests shall be fast starts
- l. (in accordance with the 184 day testing requirement of SR 3.8.1.8 and
[ SR3.8.1.9). If this criterion is not satisfied during the first series ' i of tests, any elsernate criterton to be used to reset the valid failure
..... . . . ?. . .
A8WR 3.8 19 5/31/89 l O 16.11 19
- Amendment 9
'--_-.*_m_m_-_-__.___.----_,_,_____?__ _ , _ _ _ _ _ , , . _ _ _ , _ _ _ , _ , _ . , , _ _ ,, ,, .__. - , ya- _ , - . . , . . , - - -- w- *e-.-- W-
i MN 21A61ooAP i R*==d-rd Plant wA 16.11J 32.2 A.C. Sources.Sbutdown e A.C. Sources . Shutdown i 3.8.t c 3.8 EllCTRICAL F0m'IR $YSTEMS 3.8.2 A.C. Saueces - thutdawn LCO 3.8.2 The following A.C. Electrical Power Sources shall be 1 OPIRABLt: l A. One circuit between the off site transmission network and the on. site Class It distribution system, M , B. One diesel generator. APPLICABILITY: MODIS 4 and 5 When handitng irraciated fuel in the secondary containment. ACTIDNS , CONDITION Rt00! RID AC110N COMPLET10N 11MI A. Less than the required A.! Suspend CORE ALT [ RATIONS. Inmediately A.C. power sources
- OP[RABLE. M A.t .........noit...........
Provisions of LCO 3.0.3 are not applicable. 1 Suspend handling of lamediately ' I trradiated fuel in the secondary containment. ) M A.3 Suspend operations with As soon as ! a potential for draining practicable ' the reactor vessel. l M l A.4 Restore the A.C. power As soon as : sources to OPCRABLE practicable j ststus. ! I l I i l l ABWR 3.8 20 $/31/89 I Amendment 9 16.11 20 ) l
I 2M610MP Riandard Plant u.x , i V i i A.C. Sources $hutdown ! 3.8.2 ! t
$URVffttANEF RfDUIREMINTS '
$URVE!LLANCE FRIOutNCY i
$R 3.8.2.1 Perform $R 3.8.1.1 through $R 3.8.1.22 for According to the required equipment with exception of applicable $Rs
$R 3.8.1.5 and $R 3.8.1.9.
~
~ i i
I Q051.RfffRENCf5 TITLE NUMBIR + A.C. Sources Operating 3.8.1 i l
,n
'! x l t i l i ABWR 3.8.!! $/31/89 k Amendment 9 16.11 21
21A610QAP Sta=dard Plant Rn* A 16.113 3A.3 D6eest Feel Oil i ei 1 1 Diesel fuel 011 1 3.6.3 3.0 (LECTRICAL POWER $YSTEMS 3.8.3 Dinaal Fuel 011 {' ( LCO 3.8.3 The diesel fuel oil subsystem shall be OPERABLE for each I required diesel generator. I ! APPLICABILITY: When associated diesel generator is required te be OPERABLE. !
....................... 100T[..........................
Conditions A through F may be concurrently applicable. ! AtTIONS CONDlil0N Rt0VIRED At110N COMPLt110N 11Mt A. Fuel oil level low in A.! Restore fuel oil level I hour ! one or more day tanks. indayfueltank(s). i
- 8. Fuel oil transfer B.1 Restore fuel oil transfet 8 hours ,
capability inoperable capability to OPERABLE for one or more diesel status, generators. C. Fuel oil level low in C.! Restore fuel oli level 24 hours , one or more storage in storage tank (s), tanks. , D. Fuel oil properties, 0.1 Restore fuel oli 72 hours except for accelerated properties to within stability testing, de limits. not meet limits in one or more storage tanks. E. Fuel oil accelerated E.1 Restore fuel oil 30 days stability testing properties to with;n properties do not meet limits. limits in one or more storage tanks. (continued) ABWR 3.8 22 5/31/89 O Amendment 9 16.11 22
i i 21A6100AP SimsmAnrd Pimnt a, 4 :
?
f r% i Diesel Fuel Oil - 3.8.3 Att!DN1 frantinued) CONDITION RIQUIRID ACTION COMPLET!ON TIM [ F. Required Actions and F.1 Declare associated lesnediately associated Completion diesel generator (s) Tim s of Condition A, inoperable. B C D, or i not w t. E Diesel fuel oil subsystem thoprobie for reasons other than Condition A, B, C O, or (, t l ABWR 3,s.23 5/31/89 Annendment 9 16.11 23 l i
i MM 2M6100AP i Remndard Plant _ wA O Diesel Fuel Oil , 3.8.3 !
$UpVI1LLANff RfDUIREMENTS
$URVilLLANCE FRt0VENCY i ;
$R 3.8.3.1 Verify each fuel day tanc contains According to !
2I )gallonsoffuel. Table 3.8.1 1
$R 3.8.3.2 Verify the fuel storage tanks contains 31 days 2[ }gallonsoffuel.
SR 3.8.3.3 Demonstrate eeth required fuel transfer 31 days pump starts and transfers fuel from its story e tank to its day tank.
$R 3.8.3.4 Remove accumulated water in the fuel day Once within tanks. 24 hours after 2 1 hour of .
diesel , operation
$R 3.8.3.5 Demonstrate properties. other than impurity Prior to l I
level, of new fuel oli sample are within addition of , limits, new fuel oil l to storage tanks
$R 3.8.3.6 Demonstrate impurity level of fuel oil in Once within .
storage tanks is within limits. 7 days after addition of new fuel oil to storage tank i
$R 3.8.3.7 Demonstrate properties of fuel oil storage $2 days tank sample are within limits.
(continued) ABWR 3.8 24 5/31/29 ! O I Amendm:nt 9 16.11 24 1 l
1 MM 21A6100AP f Standmed Plant nev A - i f Diesel *sel Oil 3.8.3 itftVffttAN!f RfDUltfMINTS fecatieued) ,
$URV[lLLANC[ FR(QUINCY
$R 3.8.3.8 Remove accumulated water in the fuel oil 92 days .
Storage tanks. I i
$R 3.8.3.9 for the fuel oil system: 10 years ..
A. Drain each fuel oil storage tank. B. Remove the sediment from the storage tank. C. Clean the storage tank, t tkDSS.REFtRENCES TITLE HUMBER A.C. Sources Operating 3.8.1 A.C. Sources . Shutdown 3.8.2 l ABWR 3.8 25 5/21/89 Amendment 9 16.11 15
MM . 2M6100AP
.EtanAmed Plant Rn A 36.11A 3A4 D.C.Soortes Operating i
I D.C. Sources Operating ' 3.6.4 l 4 3.8 [LECTRICAL POWER SYSTEMS l
.1 3.8.4 D.C. Sources . Geeratine !
LC0 3.B.4 Divisions 1, 2, 3 and 4 of the D.C. Electrical Power I
$ources shall be OPERABLE.
APPLICABILITY: MODE $ 1, 2, and 3. gI10NS CONDITION REQUIRED ACTION COMPLETION TIME L Division 4 D.C. oower A.) Restore inoperable power 3D days source incoerable. source to OPERABLE i status. l l jL One of the required B.) Restore inoperable Z_ dan from glyb ion 1. 2. or 3 power source to discovery of - 0.C. power sources OPERABLE status. inoperable j inoperable, power source ; L Two of the reouired L1 Restore at least one of 24 hours , D.C. nower sources the inoperasle power f inoperable. sources to i)PERABLE statut. MD L2 Restore the initial 7 days from inonerable power tource discovery of to OPERABLE status. initial inoo. erable Divi. sinn 1. 2 or 3 cower source E 30 days from riiscovery of
..nitial unocerable Division 4 DoWar source (continued)
ABWR 3.8 76 5/31/89 O Amendment 9 16.11 26
ABWR w iour Sinndard Plant u.A 1 l U,m l D.C. Sources - Operating ; 3.8.4 l ACT1DNS febetinued) , l CONDITION RE0VIRED ACTION COMPLET10N TIM [ t t L Three of the reouired M Restore at least one of h D.C. nower sources the inanarable newer ' inocerable. sources to OPERARtf 1141E1 M M Restore at least two of 24 heues fre* the inonerable newer discovery of sources to OPERABLI second inon. 1,Q1,y1 erable oower ' source. 5 W Restore the initial 7 days from inonerable onwar source discovery of a to OPERABLE status. initial inon. etable Divi. [. sion 1. 2 or ( 3 nower source QB 3D days from discovery of initial inonerable Division 4 ' Dower suurte E. Required Actions and E.1 Be in MODE 3. 12 hour. associated Completion Times of Condition A, 3 B. C or D not met. E.2 Be in MODE 4. 36 hours l l ABWR 3.8 27 5/31/89 l O g Amendment 9 16.11 27
- - =_. - -- .- - - -
l 31A610QAP Simmstmeni Plaeit m,.. A . I 9: ! r t 0.C. Sources Operating 3.8.4 l suRytltt&K t krou1REMENis SURV!!LLANtt FRt0L2NCY .
- )
$R 3.8.4.1 Demonstrate the Battery Terminal Voltage 7 days 1129 volts on float charge. ;
i l r r SR 3.8.4.2 Verify no visible corrosion at terminals 92 days l or connectors. l AlO DB Once within L Demonstrate the connection resistance of 31 days ' terminal and connectors is after a s 150x10)6 char. battery discharge , below 110 ' volts blQ , i l Once within 31 days ' after a ' battery l overcharge above 150 volts l
$R 3.8.4.3 Verify the cells, cell plates and battery 18 months !
racks show no visual indication of physical I damage or abnormal de*erioration. SR 3.8.4.4 Verify the cell to cell and tereins) 18 months connections are clean, tight, free of corrosion and coated with anti corrosion i material. ' l (continued)
)
l l l A, ABWR 3.8 28 $/31/89 O Amendment 9 16.11 28
MM 23A6100AP Semadard Plant ma A . D.C. $0urces Operating 3.8.4
$URV[IllANf f RfDUIREMENTS f remtinued)
$URVilLLANtt FRt0VENCY
$R 3.8.4.5 Demonstrate the resistance of each cell and 18 months terminal connection is s 160 x 10'6 chss.
$R 3.8.4.6 Demonstrate each battery charger will 18 months supply the following emperage at a sinimum of 125 volts for at least 10 hours:
A. t 500 amperes for Divisions 1, 2, and 3.
- 8. t 200 amperes for Division 4.
$R 3.8.4.1 *... . ........N0TE.......................
Performance of $R 3.8.4.8 set'sfies this surveillance, j p) I V Demonstrate bettery capacit/ is adequate to 18 months supply and maintain in OPEFABLE status all actual or simulated emergency leads for the design duty cycle wheA the battery is subjected to a battery service test.
$R 3.8.4.8 Demonstrate the battery capacity is at least 60 s.anths (80%ofthemanufacturer'srating)when subjected to a performance discharge test. AliQ 18 months when battery shows degradation or has reached 85% of the expected life 1
ABWR 3.8 29 5/31/29 Amendment 9 16.11 29
4 MM 21A6100AP
. Standard Plant w^ ;
Ol! D.C. Sources.Operstitj l j 3.8. , i CRD$$ REFEPfNCES TITLE NUMB (R , l D.C. Sources . Shutdown 3.8.5 i l l Battery Electrolyte 3.8.6 ; i l l l O i l l > l l l l + I ABWR 3.8 30 5/31/29 6 , Amendment 9 16.11 30
p MN 23A6100AP y Senadard Plant Rev A
.f m- 16.11.5 3AJ D.C. Sources. Shutdown
{ J
'.L A. ..
D.C. Sources Shutdown 3.8.5 3.8 ELECTR' CAL POWER SYSTEMS 1 3.8.5 D. c. Sources Shutdown LCO 3.8.5 One Division of the Division 1. 2 or 3 D.C. Electrical Power Sources sn'all be OKRABLE. APPLICABILITY: MODES 4 and 5 When handling irradiated fuel in the secondary containment. ACTIONS CONDITION- REQUIRED ACTION COMPLETION TIME A. Less than the required A.1 Suspend CORE ALTERATIONS. Immediately D.C. power sources OPERABLt. E A.2 --........ NOTE- - --- - Provisions of LCO 3.0.3 are not applicable. Suspend handling of Innediataly
.( irradiated fuel in the secondary containment.
M A.3 Suspend operations with As soon an a potential for draining practicable the reactor vessel. M A.4 Restore at least one As soon as division (Divison 1, practicable 2 or 3) of the D.C. power system to OPERABLE status. ABWR 3.8 31 5/31/89 Amendment 9 16J1-31
i 21A6100AP - hadard Plant m3 D.C. Sources Shutdown 3.B.5
$URVE1LLANCE REQUIREMENTS i SURVEILLANCE FREQUENCY SR 3.8.5.1 - Perform SR 3.8.4.1 through $R 3,8.4.8 According to for the required equipment. applicable SRs 1
Cross REFERENCES TITLE NUMBER l D.C. Sources Dperatin9 3.8.4 Battery Electrolyte 3.8.6 ; 9 l l l l I l i 1 l l ABWR 3.8 32 5/31/89 l Amendmeat 9 16.11 M l l __ -,.m .
..m-.., e
i ABM 2miooAr i Seandard Plant Rn A 16.114 3.84 Bettery Dectrolyte {
- U .
i Battery Electrolyte 3.8.6 , 3.8 [LECTRICAL POWER SYSTEMS 3.8.6 pattery Electrolyte LCO 3.8.6 Battery electrolyte for the Divisions 1, 2, 3 and 4 batteries shall be within the limits of Table 3.8.671. APPLICABILITY: When associated D.C. power sources are required to be OPERABLE. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIMI A. One or more cells in A.1 Demonstrate the pilot I hour one or more batteries cells' electrolyte level not within limits, and float voltage meet Category C allowable values.
&!!D A.2 Demonstrate the 24 hours carameters in lable
.; 3.8.6 1 meet Category C
\ allowable values.
&!fD A.3 Restore the parameters 7 days to Category A and B liatts of Table 3.8.6 1.
- B. Requiled Actions and B.1 Declare associated D.C. Immediately associated Completion power source inoperable.
Times of Condition A not met. 08 Average electrolyte temperature of the pilot cells is s 60'F. ABWR 3.8 33 5/31/89 Amendment 9 16.11 33
u 1ABWR =me L: Standard Plant hvA , F Battery Electrolyte 3.8.6
$URVEILLANCf RfDUIREMENT$
i
$URVIILLANCE FR(QUENCY ,
! SR 3.8.6.1 Demonstrate the parameters in Table 7 days l 3.8.6 1 meet the Category A limits. SR 3.8.6.2 Demonstrate the parameters in Table 92 days 3.8.6 1 eneet the Category B limits. E Once within l 31 days after i a battery discharge below 110 volts M Once within' 31 days after a battery overcharge above 150 volts <
. SR 3.8.6.3 Verify the average electrolyte temperature 92 days I of the pilot cells > 60'F.
L m Once within 31 days after a battery l discharge below , 110 volts l-l; M 1 Once within 31 days after ( a battery overcharge l- ' above 150 volts i ABWR 3.8 34 5/31/89 O l_ Amendment 9 16.11-M
.~ . .. --. . . - _ - . . . . - .-
ABWR 2mimir
- Standard Plant Rev.A .p . , f
- 4 ,
Battery Electrolyte 3.8.6 , CROS$ REFERENCES TITLE NUMB [R ;
)
D.C. Sources Operating 3.8.4 D.C. Sources . Shutdown 3.8.5 i O 4 1 ABWR 3.8 35 5/31/89 Ameadmeat 9 16 11 35
MM . . 21A6100AP ' Sieadaed Pimnt _ Rev A O t k Battery Electrolyte 3.8.6 Table 3.8.61 (Page 1 of 1) Batterv Surveillance Raouirements CATEGORY A CATEGORY B CATEGORY C Parameter Limits for each Limits for each Allowable designated pilot connected cell value for each cell connected cell l l= l Electrolyte > Minimum level > Minimum level Above top of Level- indication mark, indication mark, plates, i and 5 1/4" above and 5 1/4' above and not I maximum level maximum level overflowing indication mark indication mark Float ~ Voltage (b). 2 2,13 volts > 2.07 volts 2 2.13 volts I Specific 1 1.195(C) Gravity I 'I -> 1.190 Not more than l 0.020 below the average of all connected cells M M Average of all Average of all connected cells connected cells
> 1.200 ICI 2 1.190 I
.......................N0TES..............................
(a) Corrected for electrolyte temperature and level. l (b) May be corrected for average electrolyte temperature. (c) Or battery charging current is < 2 amperes when on float charge. l l ABWR 3.8 36 5/31/89 9 Amendment 9 16.11 36 1
l 1 ABWR 2mino.w 1 standard Plant an ^ 16.11.7 3.8.7 A.C. Power Distribution Systems-Operating U l A.C. Power Dist. Systems Operating l 3.8.7. i 3.8 ELECTRICAL POWER SYSTEMS 3.8.7 A.C. Power Distribution systems - De;7 } y g Lt0 3.8.7 The Division 1. 2. and 3 A.L. x 23, !1stribution systems shall be enero12ed. APPLICABILITY: MODES 1, 2, and 3. AtT10NS CONDITION REQUIRED ACTION COMPLET10N TIME A. One of the required A.C. A.1 Declare affected lamediately power distribution equipment inoperable. Divisions deenergized. 0.8 A.2 Re. energize the 24 hours from A.C. power distribution discovery of Division, deenergized A.C. power p 4 distribution Division ,V L "wo of the recuired B.1 Declare affected lamediately i.C. nower distribution equipment inoperable.
)ivisions deeneraized.
0.B B.2.1 Reenergize at least one 8 hours from of the deenergized A.C. discovery of power distribution second Divisions. deenerg11td A.C. oower distribution division Ml0 B.2.2 Reenero12e the initial 24 hours from deenercized A.C. oower discovery of distribution division. initia' deener. aired A.C. nower distribution division, (continued) ABWR 3.B-37 5/31/89 N Amendment 9 16.11 37
i MM 23A6100AP Remndard Plant a- ^ O A.C. Pow r Dist. Systems Operating 3.B.7 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME L All three of the M Declare affected Immediate1v
- reouired A.C. nower mouinment inantrable.
distribution Divisions deenercized. 98 L2 Reeneraire at least one 2 hours A.C. nower distribution Division. D. Required Actions and D.1 Be in MODE 3. 12 hours associated Completion Times of Condition A. MD B or C not met. D.2 Be in MODE 4. 36 hours O 4 0
]
ABWR 3.8-38 5/31/89 Amendment 9 16.11 38
LABWR 2mloog . , Standard Plant Rev A ) f~s A.C. Power Dist. Systeus Operating 3.8.i SURVEILLANCE REQUIREMENTS , SURVEILLANCE FREQUENCY SR 3.8.7.1 Verify correct creaker alignment and 7 days voltage to A.C. Power distribution system. CROSS-REFERENCES + TITLE NUMBER , A.C. and D.C. Power Distribution Systems - Shutdown 3.8.9 l l l ABWR 3.8 39 5/31/89 O Amendment 9 16.11 39
* . - - _ _ _ -- --_- - . _ _ - a -- 9 +-w' * ?
MM 23A6100AP ~
- Standard Plant mA i
16.113 3AA D.C. Power Distribution Systems Operating i
'I D.C. Power Dist. Systems - Operating 3.8.8 3.8 [LECTRICAL POWER SYSTEMS 3.8.8 D.C. Power Distribution Sys+ -* - Doeratine LCD 3.8.8 The Division 1. 2. 3. and 4 D.C. nower distribution syst m shall be energized.
APPLICABILITY: MODES 1, 2, and 3. ACT1DNS CONDITION REQUIRED ACTION COMPLElI0N TIME
+
A. One of the required D.C. A.1' Declare affected lamediately power distribution equipment inoperable. Divisions deenergized.. E A.2 Re energize the 24 hours I D.C. power distribution Division. L Two of the reouired D C. El Declare affected immediately - Dower distribution enuineant incoerable, bivisions deenereired. Q.B LL1 Reeneroire at laatLADg 2 hours of the dennerai?.ed D.C. oower distribution Divisions. M LL2 Reeneraire t se intial 24 hours from deenerai2ed ).C . oower discovery of distribution division initial deenercized D.C. oomer distribution division. C. Required Actions and C.) Be in MODE 3. 12 hours associated Completion Times of Condition A MQ or B not met. C.2 Be in MODE 4. 36 hours (continued) ABWR 3.8 40 5/31/09 i Amendment 9 16.11-40 f
23A6100AP
- Standard Plant Rev A
-[ Ns. D.C. Power Dist. Systems - Operating 3.8.8 ACTIONS (continued) CONDITIDN REQUIRED ACTION COMPLETION TIME L Three or more of the D.1 Declare affected immediately required D.C. power equipment inoperable, distribution Divisions deenergized. D.B D.2.1 Be in MODE 3. 12 hours AND. D.2.2 8e in MODE 4. 36 hours SURVf1LLANCE REQUIREMENTS SURVE!LLANCE FREQUENCY U SR 3.8.8.1 Verify correct breaker alignment and 7 days voltage to D.C. power distribution system. CROSS-REFERENCES ' TITLE NUMBER A.C. and D.C. Power Distribution Systems Shutdown 3.8.9 ABWR 3.8 41 5/31/89 Amendment 9 16.11 41
ABWR . 2mlair Standard Plant a- ^
' 16.11.9 3A.9 A.C.and D.C. Power Distribution Systems. Shutdown O
I A.C. & D.C. Power Dist. Systes.is Shutdown 3.8.9 3.8 ELECTRICAL POWER SYSTEMS 3.8.9 A.C. and D.C. Peuer Distribution Systems . Shutdown LCD 3.8.9 One conclate division of A.C. and D.C. nower distribution systee shall be eneraired. APPLICABILITY: MODES 4 and 5 When handling irradiated fuel in the secondary containment. ACTIONS CONDIT]DN REQUIRED ACT10N COMPLETION TIME A. One or more of the A.! Suspend CORE ALTERATIONS. Immediately required power distribution buses E deenergized. A.2 ..........N0TE.......... Provisions of LCO 3.0.3 are not applicable. Suspend hardling of lunediately irradiated fuel in the secondary containment. E - A.3 Suspend operations with As soon as a potential for practicable draining the reactor vessel. E 1 A.4 Reenergize at least one As soon as ' complete Olvision of practicable j A.C. and 0.C. power distribution.
.r l
l ABWR 3.8-42 5/31/89 l Amendment 9 16.11-12 l
A15WR waar Standard Plant _. a.v ^ ' i
'k A.C. & D.C. Power Dist. Systems Shutdown 3.8.9
$URVEILLANCE REQUIREMENTS
$URVEILLANCE FREQUENCY SR 3.8.9.1 Perform SR 3.8.7.1 and SR 3.8.8.1 for According to the required equipment. SR 3.8.7.1 and SR 3.8.8.1 l
CROSS-REFERENCES , TITLE NUMBER A.C. Power Distribution Systems Operating 3.8.7
'~
D.C. Power Distribution Systems - Operating 3.8.8 O ABWR 3.8-43 5/31/89 Amendment 9 16.11-43
23A6100AP Standard Plant Rev A i SECTION 16.12
-,' O V CONTENTS Sec11on ' M g 16.12 1 3.9.1 Refueling Egult taent laterlocks 16.12 1 :
~ 16.12.2 3.9.2 Refuel Position One-Rod Out laterlock 16.12 2 16.12 3 3.93 Control Rod Position 16.12 4 16.12.4 3.9.4 Control Rod Position Indication 16.12 5 16.12.5 3.9.5 Control Rod OPERABILin' Refueline 16.12 7 16.12.6 3.9.6 Water Level Reactor Pressure Vessel 16.12 8
- 16.12.7 3.9.7 Water level.Snent Fuel Storane Pools 16.12 9 L 16.12.8 3.9.8 Residual Heat Removal.Hinh Water level 16.12 10 I-
! 16.12.9 3.9.9 Residual Heat Removal.1m Water level 16.12 12 l O
~t l
1 i O 16.12-ii Araendment 9
i
?
.ABWR zwi=^r Standard Plant e- A 16.12 3.9 REFUELING OPERATIONS 16.12.1 3.9.1 Refueling Equip;aent laterlocks Refueling Equipment Inte-locks 3.9.1 3.9 - REFUELING OPERATIONS 3.9.1 Refueline feuianent Interlocks LCO 3.9.1. The Refueling Equipment Interlocks shall be OPERABLE with the following inputs:
A. All rods in. B. Refuel platform position. C. Refuel platfore main hoist fuel loaded. APPLICABILITY: MODEassociated 5whenhandlingththeinterlocksfuel w assemblies using equipment ACTIONS CONDITION REQUIRED ACTION CONPLETION TIME
.- A. One or more of the A.! Suspend CORE ALTERAT!0NS lanediately required Refueling' with equipment associated Equipment Interlocks with the inoperable .(a -
t. Inoperable. interlock, i 1 SURVEILLANei MEDUIREMENTS _ SURVEILLANCE FREQUENCY L SR 3.9.1.1 Performa[CNANNELFUNCTIONALTEST). 7 days b l~ CROSS.REFE U NCES I TITLE NUMSER Control Rod Drive Removal Refueling 3.10.5 ABWR 3.9 1 5/31/89 O Amendment 9 16.12 1
ABWR - w>aae Standard Plant h^ . 1612.2 3.9.2 Refuel Position One Rod-Out laterlock i Refuel Position One Rod Dut Interlock : 3.9.2 , 3.9 REFUELING OPERATION $ 3.9.2 Refuel Position One Red out Interloc2 LC0 3.9.2 The Refuel position one rod out interlock shall be OPERABLE. APFLICABILITY: MODE 5 with the Reactor Mode Switch in the Refuel position and any control rod withdrawn. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME , A. Refuci position A.) Suspend control rod immediately one rod out interlock withdrawal. inoperable. AtiD
~~
A.2 Fully insert all As soon as insertable control rods practicable in core cells containing
- one or more fuel assemblies.
SURVElltANCE REDUIREMENTS SURVE!LLANCE FREQUENCY l u SR 3.9.2.1 Performa[CHANNELFUNCTIONALTEST). 7 days 03 1 hour when any control rod is withdrawn I ABWR S.9-2 5/31/89 4se Amendment 9 16.12-2
1 1
-ABWR 2muw -i Standard Plant am ^ J i
( A_/ Refuel Position One. Rod Out Interlock 3.9.I CROS$ REFfRENCES TITLE NUMBER Control Rod Withdrawal - Hot Shutdown 3.10.3 Control Rod Withdrawal Cold Shutdown 3.10.4 Control Rod Drive Removal Refueling 3.10.5 i i i 1 ABWR 3.9-3 5/31/89 O Amendment 9 16.12-3
MM 23A6100AP t fr !Illiandard Plant h^ 16.12.3 3.9.3 Control Rod Position ' Control Rod Pesition 3.9.3 3.9 REFUEL}NG OPERATIONS 3.9.3 control Rod Position LCO 3.9.3 All control rods shall be fully inserted.
-APPLICABILITYi When loading fuel assemblies into the core and not following an approved spiral reload sequence.
ACTIONS CONDITION REQUIRED ACTION COMPLET10N TIME A. All control rods not A.) Suspend loading fuel lamediately fully inserted. assemblies into the core. SURVEILLANCE REQUIREMENTS G. SURVEILLANCE FREQUENCY SR 3.9.3.1 Verify all control rods are fully inserted. 12 hours CROSS
REFERENCES:
None ABWR 3.9 4 $/31/89 Amendment 9 16.12-4
MM 2M6100AP ; Standard Plant ma ^ ' i
- . . . 16.12A 3.9A Control Rod Position ladication 2q ' y)-:
Control Rod Position Indication ' 3.9.4
- 3.9 REFUEL!NG OPERATIONS-3.9.4 control Red Position Indication LCO 3.9.4 All control rod Full In positior. indicators for control rods in core cells containing one or more fuel assemblies shall be OPERABLE.
APPLICABILITY: MODE 5. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1.1 Declere the Refueling Immediately position indicators Equipment Interlocks inoperable. inoperable. AD A.I.2 Declare the Refuel Innediately
.Ig .
i Position One Rod Out Interlock inoperable. d a - A.2.1 Fully insert the control As soon as rod (s) associated with practicable the increrable position indicator (s), dO A.2.2 Disarm the associated As soon as control rod drive (s). practicable I i 1' ABWR 3.9 5 5/31/89 A l u Amendment 9 16.12-5 l , _ _ _
M 23A6100AP i Standard Plant wA J V Control Rod Position Indication 3.9.4 - SURVEfttANCE REDUIREMENTS
$URVEILLANCE FREQUENCY
$R 3.9.4.1 Verify each required control rod 7 days Full in position indicator is OPERABLE.
CROSS REFERENCES Ti1LE NUMBER Refueling Equipment Interlocks 3.9.1 Refuel Position One. Rod.Out Interlock 3.9.2 Control Rod Withdrawal Hot Shutdown 3.10.3 Control Rod Withdrawal - Cold Shutdown 3.10.4 Control Rod Drive Removal - Refueling 3.10.5 ABWR 3.9 6 5/31/89 ) O l Amendment 9 16.12-6 f 1
ABWR zwaar w^ Standani Plant i
~
16.12J 3.9J Control Rod OPERABILIW Refueling : > v Control Rod OPERABILITY Refueling 3.9.5 ' 3.9 REFUELING OPERATIONS 3.9.6 control and OPERARILITY - Refueline LCO 3.9.5 Each control rod withdrawn from a core cell containing one or more fuel assemblies shall be OPERABLE. AFPLICABILITY: MODE 5. l ACTIONS CONDITION REQUIRED ACTION COMPLET10N TIME l A. One or more required A.1 Fully insert inoperable As soon as I control rods control rods, practicable 1 inoperable, i .. SURVEILLANCE REQUIREMENTS V SURVEILLANCE FREQUENCY SR 3.9.5.1 Insert eacn required withdrawn control rod 7 days at least one ilta. l- SR 3.9.5.2 Verify each required control rod scram 7 days
- accumulatorpressureis2[1350)rstg.
l 1 ;. !- CROSS. REFERENCES TilLE NUMBER Control Rod Withdrawal - Hot Shutdown 3.10.3 Control Rod Withdrawal - Cold Shutdown 3.10.4 Control Rod Drive Removal Refueling 3.10.5 ABWR 3.9 7 5/31/89 l=
%)
Amendment 9 16.12-7
-w .
MM zusio0AP - Standard Plant h^ 16.124 334 Water level Reactor Pressure Vessel ' - Water Level RPV
-3.9.6 3.9 REFUCLING OPERATIONS 3.9.6 Water Level . Reactor Pressure Vessel LCO 3.9.6 Water level shall be t 23' over the top of the Reactor Pressure Vessel (RPV) flange.
APPLICABILITY: When handling irradiated fuel assemblies over the RPV. When handling fuel assemblies over the RPV with irradiated fuel assemblies seated within the RPV. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Water level < 23' A.) Suspend handling fuel lamediately over the top of the assemblies over the RPV. RPV flange. SURVEILLANCE REQUIREMENTS SURVEILLANCC l FREQUENCi SR 3.9.6.1 Verify water level in the RPV is 24 hours 2 23' over the top of the RPV flange. CROSS-REFERENCES TITLE NUMBER ECCS - Shutdown 3.5.2 Residual Heat Removal - High Water Level 3.9.8 Residual Heat Removal Low Water Level 3.9.9 ABWR 3.9 8 5/20/89 l-I Amendment 9 16.12-8
.~ - -- . . . . - - . .
j MM 23A6100AP jf R*==dard Plant m- A I
- SL12.7 3.9.7 Weenr Essel Syset Feel Steeny Peels J
l Water Level . Fuel Pools + 3.9.7 -
)
3.9 REFUELING OPERATIONS 3.9.7 Mater Lawel - Ament Fuel Storane Peels LCO 3.9.7 Water level shall be 123 feet over the top of irradiated fuel assemblies seated in the Spent Fuel Storage Pools. APPLICABILITY: When irradiated fuel assemblies are stored in the Spent Fuel Storage Pools. aft 10NS CONDITION REQUIRED ACTION COMPLETION TIME A .' Water level < 23- A.) .... NCTE.- -- -- Ismediately l feet over the top of Provisions of LC0 3.0.3 trradiated fuel are not applicable, assemblies in the --.. --- . ........ .-
.. Spent Fuel Storage Pools. Suspend handling fuel i assemblies over the
() Spent Fuel Storage Pools. L l l- !
- SURVEILLANCE REQUIREMENTS l- i 1 --
SURVEILLANCE FREQUENCY l SR 3.9.7.1 Verify water level is a 23 feet over 7 days i the top of irradiated fuel assembites seated in the Spent Fuel Storage Pools. CROSS.
REFERENCES:
None ABWR 3.9 9 5/20/89 Assedment 9 16.12-9 l
.. . . - -. . - . . . . . ~ . . - . -
L MM . 23A6100AP Senndard Plant w^ ' L
'16.12A 3.9A Residual Heat Removal.High Water level O !
RHR - High Water Level 3.9.8 3.9 REFUEllNG OPERATIONS 3.9.8 Residual Heat " m al - Minh Hater Laval . LCO 3.9.8 One Residual Heat Removal (RHn) shutdown cooling subsystem shall be OPERABLE. APPLICABILITY: MODE 5 with water level 2 [ feet over the top of the RPV flange and heat losse] s to ambient not sufficient to
- j. maintain average reactor coolant temperature 5140'F.
ACTIONI l CONDITION REQUIRED ACTION COMPLET10N TIME l A. No RHR shutdown cooling A.I. Provide an alternate 8 hours subsystem OPERABLE. method capable of decay heat removal. I B. Required Action and B.1 Suspend operations that Immediately l' associated Completion could increase reactor -., Time of Condition A- decay heat load, not met. M . B.2 Verify Secondary As soon as Containment is OPERABLE. practicable M B.3 Ensure at les:t one SGTS As soon as subsystem is OPERABLE. practicable M B.4 Ensure at least one As soon as i Secondary Containment pr.acticable j Isolation Valve and associated actuation l instrumentation is OPERABLE in each associated penetration not isolated. M l (continued) l 1 ABWR 3.9 10 5/20/89 Amendment 9 16.12-10 i l 1 l-
. - . . . . . . - . . . . . ..-. - _ - - . . . . . _ . . _ . . . . , _ . . _ . ~ . - .
1 1 23A610MP Standard Plant ,_ Rev A 1 l
.< j
( ' 1 l [ RHR - High Water Level 3.9.8
, E T10NS fcontinued)
CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.5 Provide an alternate As soon as method capable of decay practicable heat removal. i t l. SUkVEILLANCE REDUIREMENTS SURVEILLANCE FREQUENCY L ' l SR 3.9.8.1 Verify for the required RHR shutdown 7 days cooling subsystem each manual, power operated, or automatic valve in the flow path not locked, sealed or otherwise secured in position, is in the correct position or is capable of being manually aligned in the correct position. CROSS-REFERENCES TITLE NUMBER Secondary Containment Isolation Actuation Instrumentation [ ] r ECCS - Shutdown 3.5.2 Secondary Containment 3.6.4.1 Secondary Containment Isolation Valves 3.6.4.2 Standby Gas Treatment System 3.6.4.3 ABWR 3.9 11 5/20/89 ( Amendment 9 16.12-11
i
-M M 2M6100AP '
Standard Plant * ^ - 16.12.9 3.9.9 Residual Heat Removal.14w Water level 4. L ; RHR - Low Weter Level 3.9.9 3.9 REFUELING OPERATIONS 3.9.9 Residual heat "- .si - Law Water Level LCO 3.9.9 Two Residual Heat Removal (RHR) shutdown cooling subsystems < shall be OPERABLE. APPLICABILITY: MODE 5 with water level < [ ] feet over the top of the RPV flange and heat losses to ambient not sufficient to maintain average reactor coolant temperature s 140*F. ACTIDNS~ CONDITION REQUIRED ACTION COMPLEil0N TIME A. One of the required A.) Provide an alternate 8 hours from ' RHR shutdown cooling method capable of decay discovery of subsystems inoperable, heat removal. inoperable subsystem E A.2 Raise water level to 8 hours from - 2[ feet over the discovery of topof]theRPVflange. inoperable subsystem B. Both of the required 8.1 ' Provide an alternate 2 hours RHR shutdown cooling method capable of decay subsystems inoperable, heat removal. (continued) ABWR 3.9 12 5/20/89 Amendment 9 16.12-12
1 ABWR mi. Senadard Plant- a ,, x A:. kj-. RHR - Low Water Level 3.9.9 ACTIONS fcontinued) CONDITION REQUIRED ACTION COMPLETION TIME . 1 C. ~ Required Actions and C.1 Sus nd operations that immediately associated Completion cou irecrease reactor l Times of Condition A decay heat load. I or B not met. M I
\
C.2 Ensure Secondary As soon as Containment is OPERABLE, practicable j M. C.3 Ensure at least one SGTS As soon as subsystem is OPERABLE. practicable M C.4 Ensure at least one As soon as-Secondary Containment practicable Isolation Valve and associated actuation i instrumentation is '
\ OPERABLE in each associated penetration not isolated.
L M l .. C.5 Provide at least one As soon as t alternate method capable practicable of decay heat removal. l-l ABWR 3.9-13 5/20/89 O A:nendment 9 16.12-13
.M M 21A6100AP 'f Standard Plant w^-
O RHR . Lcw Water Levt1 3.9.9
- SURVEILLANCE RfDUIREMENTS SURVElLLANCE FREQUENCY SR 3.9.9.1- Verify for the required RHR shutdown 7 days cooling subsystem each manual, power operated, or automatic valve in the flow ,
-path not locked, setled or otherwise secured in position, is in the correct position or is capable of being manually 6119 ned in the correct position.
C'A0$$-Rf f fftfNCES TITLE NUMBER
.5econdary Containment isolation Actuation Instrumentation [ ]
ECCS Shutdoom 3.5.2 L $1cond;ry Containerent 3.6.4.1 5econdary Containment Isolation Valves 3.6.4.2. - Standby Gas Treatment System 3.6.4.3 ABWR 3.9-14 5/20/89 O Amendment 9 16.12-14
l- i 23A6100AP Standard Plant Rn A n SECTION 16.13 I $h$ ~- CONTENTS Section Dilt 2 age
)
)
16.13.1 3.10.1 Innervlee Laak and Hydrostatic Testing Operation 16.131 :i 16.13.2 3.10.2 Reactor Mode Switch latedeck Testing 16.13 3 j 16.13.3 3.10.3 Control Rod Withdrawal-Hot Shutdown 16.13 5 16.13.4 3.10.4 Contrd Kod Withdrumul-Cold Shutdown 16.13 4 16.13.5 3.10.5 Control Rod Reinoval Ref'Ellag 16.13 11 16.13.6 3.10.6 Multinle Control Rod Withdrawal Refuelinn 16.13 14 !i\ 16.13.7 3.10.7 Control Rod Testinn-Oneratinn 16.13 16 ; h t V O [ I ( t i l l l O 16.13-11 Amendment 9
i MM i zu6100AP ! Sensulard Plant mev A 16.13 3.10 SPECIAL OPERATIONS () m. 16.13.1 3.10.1 Inservice Isak med Hydrostatic Testing Operellon i laservice Leak and Hydro lest 3.10.1 3.10 SPICIAL OPtRATION$ 3.10.1 innervice t aak and kvdrastatic festina emmention Lt0 3.10.1 The average reactor coolant temperature specified in Table 1.1 1 for MODE 4 operation may be changet' to 'Any temperature
- to allow performance of an inservice leak or hydrostatic test provided the following MODE 3 LCOs are met:
L LCO fiseaadary tantainment taniation Actuation lnstru. mentation. Functient for Law Reatter Water leveh and ' N10h kefuel flDDr kodiation). B. LCO 3.6.4.1. Secondary Containment. ' C. LC0 3.6.4.2, secondary Coniainment Isolation Valves. D. LCO 3.6.4.3, standby Cas Treatment System, i APPLICABILITY: MODI 4 with average reactor coolant temperature > 200'F. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more of the A.1 ..........N0T[.......... above required LCOs Required Actions to be not met. In MODE 4 include reducing average t'enctor coolant temperature to 5 200*F. i Enter the applicable immediately r ! Condition of the affected LCOs. QB A.2.1 Suspend activities that innediately could increase the average reactor coolant temperature or pressure, bND (continued) i l ABWR 3.10 1 5/31/89
?
v I Amendment 9 16.13 1
i MM 2%6100AP { Re.nue.rd Plant met ^ t i laservice leak and Hydro Test 3.10.1 ' ACTfDNS franH nued) CONDITIDN REQUIRED AtllDN COMPL(T!DN TIME A. (continued) A.2.2 Reduce average reactor 24 hours coolant temperature to , s 200'F. 1 E tuRytittANet arou1RtntNTS
$URV[]LLANC[ FRtQU(H0Y ,
SR 3.10.1.1 Perform the applicable SRs for the According to required MODI 3 LCOs. the applicable e SRs : t CROSS REFERENCES O TITLE NUMBER Secondary Containment Isolation Actuation Instrumentation ( ) , Secondary Containment 3.6.4.1 '- Secondary Containment isolation Valves 3.6.4.2 Standby Ces Treatment System 3.6.4.3 ' ABWR 3.10 2 5/31/89 . O Anwedment 9 IO2
-- ~ .. . - - - - .. .-
r MM 21M100AP Remndard Plant h^ 16.13.2 3.10.2 Itanctor Mode Switch laterlock Testing
/_1 [
V - Reactor Mode Switch Interlock Testing i 3.10.2 ! 3.30 $PICIAL OPERATIONS 3.10.2 * - ter " " seritch Interinek teattna LCO 3.10.2 The Reactor Mode Switch position specified in Table 1.11 for MODI 5 3. 4 and 5 operation may be changed to include the , Run Startup/ Hot Standby, and Refuel position to allow testing of instrumentation associated with the Reactor { Mode Switch interlock functions provided all control rods remain fully inserted and no other CORE ALT [ Rail 0N5 are in ' progress. APPLICABILITY: MODIS 3 and 4 with the Reactor Mode Switch in the Run. Startup/ Hot Standby, or Refuel position. ; MODE 5 with the Reactor Mode Switch in the Run or i Startup/ Hot Standby position. ACTIONS CONDITION R[ QUIRED ACTION COMPLITION 11Mt , n A. Reevirements of the A.) [ LCO not met. Suspend CORT ALT [ RATIONS except for control rod lamediately i ( insertion. i b.!iD A.2 Fully insert all I hour insertable control rods. O , A.3.1 Place the Reactor Mode i bour Switch in the Shutdown : position. DB A.3.2 ..........N01E.......... Not applicable in MODIS 3 and 4 Place the Reacto- Mode I hour Switch in the Refuel , l position. I l ABWR 3.10 3 5/31/09 O Amendment 9 16.13-3
i MM 23A5100AP i Reasulard Plant h^ el l Reactor Mode $ witch Interlock Testing 3.10.2
$UPvtittAgtt AtDutormtWTs
$URytlLLANCE FREQUINCY .
$R 3.10.2.1 Verify all control rods are fully 12 hours inserted. ;
$R 3.10.2.2 Vertfy no other CORE ALTIRATION$ are 24 hours in progress.
CROS$.R[F[R(NC[$1 Nose O . h i i ABWR 3.10 4 5/31/89 O Anwedment 9 16.tM
MM
. Sea 'eed Pimni 21A61oQAP m- A l 16.13J 3JtJ Centeel Red Withdrew 31. Hot Shutdown ..V
[._ I ! l i Control Rod Withdrawal - Hot Shutdown 3.10.3 1 3.10 $PECIAL OPIRATIONS { 3.10.3 control and Withdrawal het W - f t (to 3.10.3 The Reactor Mode $ witch position specified in Table 1.11 ' for MODI 3 operation may be changed to include the Refuel position to allow withdrawal of a single control rod or i control rod pair provided the following requirements are met: A. LCO 3.9.2. Refuel Position One. Rod.0ut Interlock. ! B. LCO 3.9.4. Control Rod Position Indication. . C. All other control rods are fully inserted. l L L LCO 'IReactor Pro eetion sys' ee Instrumentation, Fune< 1ons for: BRM Hioh F um. F ast Period, > and . NOP: APRM Hioh istaown Flur and INOPs t l Reacter Mode $ witch in Shutdown: and Manua' Scram. All for Nade 5 anolicability oniv. LCD 3.9.5, Control Rod OPERABILITY Refueling. I p ; E < L All other control rods in a five.by five array ' centered on gath control rod being withdrawn are disarmed- ; LCD 3.1.1, MODE 5, $HUTDOWN MARGlN, except the control rod for cair) to be withdrawn may be assumed to be the highest worth control rod glit. APPLICABILITY: POD [ 3 with the Retttor Mode $ witch in the Refuel position. l l t i i ABWR 3.10-5 5/31/89 Amendment 9 16.13 $
i MM 21A6100AP
. Rem-dard Plant %.A l
?
- ' i e! i Control Rod Withdrs=41. Hot Shutdown 3.10.3 ,
Affl0N$ CONDITION Rt0ylR[D ACTION COMPLIT!DN TIMt i A. One on more cf the A.) ..........N0TE.......... ; above requirements Recutred Actions to i not met. fully insert all ; insertable control rods include placing the Reactor Mode $ witch in the $hutdown i position, i Enter the applicable launediately Condition of the affecitd 8 LCOs. 93 A.t.1 Fully insert all I hour insertable control rods. A!iD A.2.2 Place the Reatter Mode
$ witch in the $hutdown positton.
I hour G ' ABWR 3.10 6 $/31/89 e Amendment 9 16.1 %
MM 21A610aAP l'
.Se==dard Plant mA
-(%./ ! - r Control Rod Withdrawal Hot $hutdown 3.10.3 ; supvritti,Ntt Rt0U1PtptN11
$URVEILLANC[ FR[QU[NCY
$R 3.10.3.1 Perfore the appittable $Rs for the According to required LCOs. the applicable SRs SR 3.10.3.P Verify all otner tot trol rods in a 24 hours r five by five array centered on g E h control ,
rod being withdrawn are disamed-SR 3.10.3.3 Verify all other control rods are fully 24 hours inserted.
=
r ( CROSS REFERENCES TITLE NUMBER SHUTDOWN MARGIN 3.1.1 Reactor Proter. tion System Instrumentation [ ] Refuel Position One Rod Out Interlock 3.9.2 ,
)
Control Rod Position Indication 3.9.4 Control Rod OPIRABillTY Refueling 3.9.5 ABWR 3.10 7 $/31/89 Ameadment 9 16.13 7
'!i MN MI#
!ktenulmed Plant ""^ l I
36.13.4 3J0.4 Centrol Rod Withdrawal Cold Shutdown O l Control Rod Withdrawa) . Cold Shutdown 3.10.4
)
3.10 $PICIAL OPERATIONS 3.10.4 central had Wit.hdrawal - Cnid $hutdown LC0 3.10.4 The Reactor Mode Switch position specified in Table 1.1 1 for MODE 4 speration may be changed to include the Refuel position to allow withdrawal of a single control rod or control rod pair or withdrawal and su:, sequent reuoval of the associated control rod drive (s) PNvided the following requirements are met: A. All other control rods are fully inserted. B. 1. LCO 3.9.2, Refuel Position One Rod Out Interlock. LCO 3.9.4, Control Rod Position Indication. E
- 2. A control rod withdrawal block is inserted.
L L LCO tRemeter Pro .ection system Instrumente, Functions for! SRift Hia ) Flux. Fast Period and i JNDP: APRM Hinh Setdown Flux and INOP Reactor Mode { Switth in Shutdown andInanual Scram. All for Mode 5 Ann' icability oniv.1 LCO 3.9.5, Control Rod OPERABillTY Refueling. E L All other control rods in a five by five array centered on ggh control rod being withdrawn are disarmed. LCO 3.1.1, MODE 5. SHUTDOWN MARGIN, except the control rod for mairl to be withdrawn may be assumed to be the highest worth control rod glit. APPLICABIL11Y: MDDE 4 with the Reactor Mode Switch in the Refuel position. ABWR 3.10 8 5/31/89 O Amer 4cne at 9 16.M i
,. . , , - - . _ . - __+ , _ _ _ . . _ _ ,
. -_ . . _ . . .. . - _ - . _ m. _
MM 2m6100AP Reandard Plant h^
.O I C/ l i
1 Control Rod Withdrawal . Cold shutdown 3.10.4 i ACTIONS CONDITION REQUIRED ACTION COMPLET10N TIMt i L One or more of the A.) ..........N0TE.......... ' aboverequirement(s) Required Actions to not met with the fully insert all affected control insertable control rods ! rod 111 insertable. include placing the Reactor Mode Switch in the Shutdown position. . [nter the esplicable lamediately Condition of the affected LCOs. E A.2.1 Fully insert all I hour insertable control rods, f 5 O A.2.2 Place the Reactor hode I hour Switch in the Shutdown position. L One or more of the L1 Suspend withdrawal of lanediately above requirements not the control r-1111 and met with the affected removal of a. 6 ciated control rodli) not control rod drive 1Ll. insertable. AE B.2.1 Fully insert all As soon as control rods, practicable e E B.2.2 Satisfy the requirements As soon as of the LCO. practicable l l l l AEWR 3.10 9 5/31/89 l Amendment 9 16.1M l l
MM 2nA6100AP had=rd Plant . . . w^ , O . Control Rod Withdrawal Cold Shutdown 3.10.4
$UPytittANet RtoufREMthis '
SURVilLLANCE FRt0VINCY
$R 3.10.4.1 Perfore the applicable $Rs for the According to required LCOs. the applicable
$Rs
$R 3.10.4.2 Verify all other control rods in a 24 hours five by five array centered on 3Agb control rod being withdrawn are disarmed.
$R .' 10.4.3 Verify all other control rods are fully 24 hours inserted.
$R 3.10.4.4 Verify a control rod withdrawal block is 24 hours inserted.
O CROSS-RiFERENCis ll1LE NUMBER
$HUTDOWN MARGIN 3.1.1 Reactor Protection System Instrumentation
[ ] Refuel Position One RK0ut Interlock 3.9.2 Control Rod Position indication 3.9.4 Control Rod OPERABIL11Y Refueling 3.9.5 6 ABWR 3.10-10 $/31/89 O Amendment 9 16.13 10
i MM 2 W 100AP had=rd Plant nev ^ 16.13J 3.10J Centrol Rod Drive Removel Refseling i k ' i Control Rod Drive Removal Refueling 3.10.5 . 3.1D $PECIAL OPERATIONS ' 3.10.5 fontrol had Drive ".n1. amfueline ' LCD 3.10.5 The requirenents of LCO f RPS Instrumentation 1. LCO 1RE1 ! non-coincident andel, LCO 3.9.1 LCO 3.9.2. LCO 3.9.4 and LCO 3.9.5 may be suspended dwing MODE 5 speration to allow i the removal of a single control rod drive or centrol rod I drive mair associated with control rod 111 withdrawn from ; core celliti containing one or more fueT assemblies provided the following re;uirements are met: l A. All other control rods are fully inserted. L All other control rods in a five b).five array centered , on the control rod 111 being removed are disarmed. ' C. A control rod withdrawal block is inserted. L LCO 3.1.1. MODE 5, SHUTDOWN MARCIN, except the single control rod for nairl to be withdrawn may be assumea to he the highest worth control rod 311t. !, E. No other CORE ALTERATIONS are in p# ogress. (/ APPLICABILITY: MODE 5 with LCO 3.9.5 not met. ACTIONS , CONDITION REQUIRED ACTION ICOMPLET!ON TIME A. One or more of the A.) Suspend removal of the immediately above requirements not control rod (s) and met, associated control rod drivemechanism(s). AND A.2.1 Fully insert all As soon as control rods, practicabla 03 A.2.2 Satisfy the requirements As soon as of the LCO. practicable ABWR 3.10 11 $/31/89 O Amendment 9 IEISII
MM 21A6100AP Secad-ed Phant . _ .
*^ ,
r Control Rod Drive Removal Retueling 3.10.5 !
$URVffttANtf REDUIREMENTS
$URVEILLANCE FREQUENCY
$R 3.10.5.1 Verify all other control iods are fully 24 hours inserted.
l Er 3.10.5.2 Verify all control rods in a five by five 24 hours array centered on a gh control rod being removed are disarmed.
$R 3.10.5.3 Verify a control rod withdrawal block is 24 hours inserted.
$R 3.10.5.4 Perform the applicable SRs for LC0 3.1.1. According to the applicable
$Rs
$R 3.10.5.5 Verify no other CORE ALTERATIONS are in progress.
24 hours O. l l l l ABWR 3.10 12 5/31/89 Amendment 9 16.t3-1.'
- ABWR .
zww.e ! Semndaed Plant _ hA l 1 Control Rcd Drive Removal Refueling 3.10.6 < 1 CRDS$ RtfERfNtf5 TITLE. NUMBtR SW1DOWN MARGIN 3.1.1 j Reactor Protection System Instrueentation [ } l Reactor Protection System Non Coincident Mode [ ] Refueling [wuipment Interlocks 3.9.1 Refuel Position One. Red Out Interlock 3.9.2 I Control Rod Position Indication 3.9.4 Control Rod OPERABILITY Refueline 3.9.5 i P t ABWR 3.10 13 5/31/89 0 _ Amc.Ameat 9 16.13-13
MN 23A6100AP Standard Plant _ m,,. A . i 16.134 3.104 Multiple Control Rod WitLdrawal. Refueling h Multiple Control Rod Withdrawa) Refueling 3.10.6 1 3.10 SPECIAL OPERATION $ " 1 3.10.6 hultin1a r. antral and withdrawal . Refueline I LCO 3.10.6 The ' Full In* Position indicators may be bypassed for any ' nuder of control rods during MODE 5 operation to allow i withdrawal of those control rods, removal of associated , control rod drives, t,r both provided the following requirements are met: ! A. The four fuel assemblies are removed from the core cells a:sociated with each control rod or control rod drive to be removed.
- 8. All other control rods in core cells containing one or '
more fuel assemblies are fully inserted. APPLICABILITY: M00t 5 with one or more " Full In* position indicators bypassed. ' ACTIONS CONDITION REQUIRED ACTION COMPLET!ON TIME A. Requirements of the A.I Suspend withdrawal of Isenediately LCO not met, control rods and l l removal of hssociated control rod drives. t i h!Q ! e ~ i l A.2.1 Fully insert all As soon as i ccetrol rods in core practicable cells containing one or more fuel assemblies. QB
- A.2.2 Satisfy the requirements As soon as of the LCO. practicable ABWR 3.10 14 5/31/89 O-Ameadmeat 9 16.1114 i
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I MM 2M6100AP Standard Plant hv ^ O O MultiM e Control Rod Withdrawa) Refueling , 3.10.6 1URVffttANCf Rf00;RE4fWT1 >
$URVt!LLANr.t - FR[OVENCY
$R 3.10.6.i Verify the four fuel assemblies are 24 hours ;
removed from core cells associated with a control rod or control rod drive to tt removed. )
$R 3.10.6.2 Verify all other control rods in core 24 hours cells containing one or more fuel assemblies are fully inserted. '
I CRD1$.RIFERENCES TITLE NUMBER -i Control Rod Position Indication 3.9.4 I L ABWR 3.10 15 5/31/89 O ,, Amendment 9 16.13 15
i 2146100AP ,, Re==ntard Plant u.2 ! 16.13.7 3.10.7 Cutrol Rod Testing Operating f i i Control Rod Testing Operating 3.10.7 3.10 $PiCIAL OPIRATION$ "
) '
3.10.7 canteel and fantina . amaratina , LCD 3.10.7 ( The requirements of LCO 3.1.6 may be suspended and control rods bypassed in the god tantrol and Infarantion System IE&ll.1 as allowed by $R [ ] during MODES 1 and 2 with THERMAL POWER less than or equal to the LPSP of the E&ll to slica performance of $HUTDOWN MARGIN ! demonstrations, c Mtrol red scram time testing, control rod ' friction testing, and the Startup Test Program provided confermance with the conttol roa seque".e for the specified l' test is verttled by a second licensed operator or other . Qualified member of the technical staff. APPL.!CABILITY: PODi$ 1 and 2 with LCO 3.1.6 not met. i ACTIONS ' CONDITION REQUIRED ACTION COMPLET10N TIME A. Requirements of the A.) Suspend performance of lunediately LCO not met. the test. 4 SURVt1LLANCE RIOUIREMENTS
$URV[lLLANCE FR[QU(NCY h
$R 3.10.7.1 Verify movement of control rods is in During control compliance with the approved control rod rod movement sequence for the specified test.
CROSS RfFERFNCES TITLE NUMBER Rod Pattern Control 3.1.6 Control Rod Clock Instrumentation [ ] ABWR 3.10 16 5/31/89 Amendment 9 16.13-16
i i l J l l O i i i i I f i APPENDIX 16A TECHNICAL SPECIFICATIONS O ! h i 9 I i r i s LO 1 l'
. . . . . - . - , _ . . - . , _ . . _ _ _ . , _ . . . , - - _ - - , . . - - . . , _ , - . . . ~ . . . , . _ . . . . . , _ . . . . - . . . . , , , - . . . - . , . .
23A6100AP
;. 1 ge."=rd Plant m ,v c APPENDIX 16A ,
O riBoe orcoureurs 16A BASES 16A.1 CONTENTS 16A.1 1 16A.2 R1A S47 TTY LIMITS 16A.21 1643 R3A APPLICABILITY 16A3 1 l I 16A3.1 Limiting Conditions for Operation 16A3 1 16A3.2 Survellance Requirements 16A3 7
]
16A.4 B3.1 REiCTnTIY CON 17tOL SYSTEMS 16A.41 16A.4.1 B3.1.1 Shutdown Margin 16A.41 16A.4.2 B3.1.2 Control Rod Operability 16A.4 7 16A.83 B3.13 Control Rod Scram Times 16A.414 i l 16A.4.4 B3.1.4 Control Rod Scrani Accc nulators 16A.4 20 ; e 16A.43 B3.1.5 Control Rod Drive Coupling 16A.4 24 l 16A.4.6 B3.1.6 Rod Pattern Control 16A.4 27 16A.4.7 B3.1.7 Standby Liquid Control System 16A.4 31 16A3 B3.2 POWER DISTRIBUTION LIMITS 16A.51 16A.5.1 B3.2.1 Average Plannar Linear Heat Generation Rate 16A.51 16A.S.2 B3.2.2 Minimum Critical Power Ratio 16AS 5 16A.53 B3.23 Linear Heat Generation Rate 16A.5 9 l 16A.6 R3.3 INSTRUMENTATION llater] . 16A.7 R3.4 REACTOR COOLANT SYSJEh! 16A.71 l 16A.7.1 B3.4.1 Recirculation Pumps Operating 16A.71 16A.7.2 B3 4 2 Safety / Relief Wlves 16A.7 5 16A.73 B3.43 Operatw..al Leakage 16A.7 9 16A.7.4 B3.4.4 Specific Activity 16A.713 16A-ii Amendment 9 l' l . - - -
MM 23A6100AP RemmAmed Plant m.v c APPENDIX 16A TABLE OF CONTENTS (Continued) g Section Ihlt East 16A.7.5 B3.4.5 Residual Heat Removal Shutdc e 16A.716 16A.7.6 B3.4.6 Reactor Coolant System Pressure Temperature j limits 16A.719 ), 16A.7.7 B3.4.7 Reactor Steam Dome Pressure 16A.7 26 34AJ R3.5 ECCS and RCIC 16A.81 16A.8.1 B3.5.1 ECCS-Operating 16A.81 1 16A.8.2 B3.5.2 ECCS Shutdown 16A.8-10 36A.9 B3.6 CONTAINMEhT SYSTEMS 16A.91 16A 9.1 B3.6.1 Primary Containment Systems 16A.9-1 16A.9.2 B3.6.2 Suppression Pool 16A t26 16A.93 B3.63 Hydrogen Control 16A.9-41 16A.9.4 B3.6.4 Secondary Containment Systems 16A.9-45 16A.10 B3.7 FLANT SYSTEMS 16A.101 l 16A.10.1 B3.7.1 Reactor Building Cooling Water / Reactor 1 Building Senice Water System Operating 16A.101 ! 16A.10.2 B3.7.2 Reactor Building Cooling Water / Reactor Building Senice Water System Shutdown 16A.10-8 ) 1 i 16A.103 B3.73 Control Room HVAC Emergency Recirculation System 16A.10-;1 16A.10.4 B3.7.4 Main Condenser Offgas 16A.10-16
)
i l 16A-iii Amendment 9
i
.Re==dard MN Plant 23A6100AP m- c APPENDIX 16A O T * * ' s o r c o n t e u rs (c ==tia d) :
Section M East 16A.11 R3J EIICTRICAL POWER SYSTEMS 16A.111 16A.11.1 B3.8.1 A.C. Sources-Operating 16A.111 i 16A.11.2 B3.8.2 A.C. Sources. Shutdown 16A.1121 ; 16A.113 B3.83 Diesel Fuel Oil 26A.1123 16A.11.4 83.8.4 D.C. Sources-Operating 16A.1128 16A.11.5 B3.8.5 D.C. Sources-Shutdown 16A.1134 16A.11.6 B3.8.6 Battery Electrolyte 16A.1136 16A.11.7 B3.8.7 A.C. Power Distribution Systems Operating 16A.1140 16A.11.8 B3.8.8 D.C. Power Distribution Systems-Operating 16A.11-44 16A.11.9 B3.8.9 A.C. and D.C. Power Distribution L Systems Shutdown 16A.11-45 16A.12 R3.9 REFFELING OPERATIONS 16A.121 16A 12.1 B3.9.1 Refueling Equipment interlocks 16A.121 16A.12.2 B3.9.2 Refuel Position Onc-Rod Out Interlock 16A.12-4 . 16A.12.? B3.9.3 Control Rod Position 16A.12 7 16A.12.4 B3.9.4 Control Rod Position Indication 16A.12-10 16A.12.5 B3.9.5 Control Rod OPERABILITY Refueling 16A.12-10 16A.12.6 B3.9.f Water Level Reactor Pressure Vessel 16A.12-15 16A.12.7 33.9.7 Water Level. Spent Fuel Storage Pools 16A.1217 16A.12.8 B3.9.8 Residual Heat Removal High Water Level 16A.12-19 16A.12.9 B3.9.9 Residual Her'. Removal Low Water Level 16A.12-22 16A.13 H3.10 SPECIAL OPERATIONS 16A.131 1 16A.13.1 B3.10.1 Inservice 1.cak and Hydrostatic 16A.13-1 Testing Operation O Amendment 9
.~ , . -. . . . _ . -
f-23A6f%AP h=dard Plant wc ! APPENDIX 16A ; TABLE OF CONTENTS (Continued) g Section Dilt East 16A.13.2 B3.10.2 Reactor Mode Switch Joterlock Testing 16A.13 4 16A.133 B3.103 Control Rod Withdrawel. Hot Shutdown 16A.13-7 i 16A.13.4 B3.10.4 Control Rod Withdrawel Cold Shutdown 16A.13-10 16A.13.5 B3.10.5 Control Rod Drive Removal Refueling 16A.1313 16A.13.6 B3.10.6 Multiple Control Rod Withdrawal Refueling 16A.13-16 16A.13.7 B3.10.7 Control Rod Testing Operating 16A.13-18 ! l O 1 i
~
1 l l 1 , 16A-v Amendment 9
ABM 2miam ! Standard Plant hA ! 16A.1 Contents, Use and Application i 7-w 16A.1.1 Table of Contents j r TABLE OF CONTENTS lAAt B 2. SAFETY LIMITS B 2.0 1 8 3. LIMITING EONDITIONS FOR OPERATION AND $URVf1LLANCE RfDUIREMENTS i
- 83. APPLICABILITY Limiting Ccnditions for Operation ...............B31 Surveillance Requirements ...................B37 8 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN ...................B3.11 B 3.1.2 Control Rod OPERABILITY ...............B3.17 8 3.3.3 Control Rod Scram Times . . . . . . . . . . . . . . . B 3.1 13 8 3.1.4 Control Rod Scram Accumulators . . . . . . . . . . . . B 3.1 19 B 3.1.5 Control Rod Drive Coupling and Seperstion Detection. . B 3.1 23 8 3.1.6 Rod Patterri Control . . . . . . . . . . . . . . . . . B 3.1 27 8 3.1.7 Standby Liquid Control System . . . . . . . . . . . . B 3.3 31 i B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE . . . . . B 3.2 1 B 3.2.2 MINIMUM CRITICAL POWER RATIO . . . . . . . . . . . . . P 3.2 5 8 3.2.3 LINEAR HEAT GENERATION RATE . . . . . . . . . . . . . B 3.2 9 8 3.3 INSTRUMENTATION
[Later) B 3.4 REACTOR COOLANT SYSTEM B 3.4.1 Recirculation Pumps Operating ............B3.41 B 3.4.2 Safety / Relief Valves . . . . . . . . . . . . . . . . . B 3.4 5 B 3.4.3 Operational Leakege .................B3.49 8 3.4.4 Specific Activity . . . . . . . . . . . . . . . . . . B 3.4 13 B 3.4.5 Residual Heat Removal - Shutdown . . . . . . . . . . . B 3.4 16 B 3.4.6 Reactor Coolant System Pressure / Temperature Litits . . B 3.419 B 3.4.7 Reactor Steam Dome Pressure , . . . . . . . . . . . . B 3. 4 2 6 , B 3.5 (CCS and RCIC B 3.5.1 ECCS Operating .................. B 3.51 B 3.5.2 ECCS Shutdown . . . . . . . . . . . . . . . . . . . B 3.5 10 ABWR i 5/31/89 Amendment 9 IRI'I i:
.. ~ . . .. _. _ _ __
t ABWR 3miour l Se==dard Plant m,v i 9 , I i l TABLE OF CONTENTS (continued) R&Qt B 3. LIMITING CONDITIONS FOR OPERATION AND SURVf1LLANCf RfoulktMENTS (continued) ! B 3.6 CONTAINMENT SYSTEMS B 3.6.1 Primary Containment Systems B 3.6.1.1 Primary Containment . . . . . . . . . . . . B 3.6 1 8 3.6.1.2 Containment Air Locks . . . . . . . . . . . B 3.6 5 B 3.6.1.3 Containment Pressure ........... B 3.6 11 B 3.6.1.4 Containment Average Air Temperature . . . . B 3.6 13 8 3.6.1.5 Primary Containment [and Pressure] , Isolation Valves ............. B 3.6 15 ' B 3.6.1.6 Wetwell to Drywell Vacuum Breakers. . . . . B 3.6 21 B 3.6.2 Suppression Pool i B 3.6.2.1 Suppression Pool Average Temperature ... B 3.6 26 8 3.6.2.2 Suppression Pool Water Level ....... B 3.6 31 B 3.6.2.3 Residual Heat Removal Suppression Pool Cooling . . . . . . . . . . . . . . . . . . B 3.6 34 8 3.6.2.4 Residual Heat Removal Wetwell Spray System .................. B 3.6 38 B 1.6.3 Hydrogen Control B 3.6.3.1 Containment Hydrogen Recombiner Systems . . B 3.6 41 B 3.6.3.2 Primary Containment Oxygen Concentration. . B 3.6 42
+
B 3.6.4 secondary Containment Systems B 3.6.4.1 Secondary Containment . . . . . . . . . . . B 3.6 45 B 3.6.4.2 Secondary Containment Isolation Valves .. B 3.6 51 B 3.6.4.3 Standby Gas Treatment System ....... B 3.6 55 , B 3.7 PLANT SYSTEMS B 3.7.1 Reactor Building Cooling Water / Reactor Building Service Water System - Operating .......... B 3.71 B 3.7.2 Reactor Building Cooling Water / Reactor Building Service Water System Shutdown . . . . . . . . . . . B 3.7 7 B 3.7.3 Control Room HVAC Emergeicy Recirculation S .. B 3.7 10 B 3.7.4 Main Condenser Offgas . . . . . . . . . . .ystem ..... B 3.7 15 i ABWR 11 $/11/89 , 9: Amendment 9 IM 12 ; 1 l l
R< 1 i i MNM 2M6100AF
.Stasulard Plant Rw A (A) 1 I
J l TA8LE OF CONTENTS (continued) l
)
hat l 8 3. LINITING CONDIT10NS FOR OPERATION AND SURVE1LLANCE REDUIREMENTS (continued) B 3.B ELECTRICAL POWER $YSTEMS I B 3.8.1 A.C. Sources Operatin9 ..............B3.81 B 3.8.2 A.C. Sources Shutdown ..............B3.821 8 3.8.3 Diesel Fuel Oil . . . . . . . . . . . . . . . . . . . B 3.8 23 ' B 3.8.4 0.C. Sources . Operating . . . . . . . . . . . . . . B 3.8 28 B 3.B.5 0.C. Sources Shutdown . . . . . . . . . . . . . . B 3.8 34 8 3.8.6 Battery Electrolyte . . . . . . . . . . . . . . . . B 3.8 36 B 3.B.7 A.C. Power Olstribution Systems - Operating . . . . B 3.8 40 8 3.8.8 0.C. Power Distribution Systems Operating . . . . B 3.B.45 B 3.B.9 A.C. and D.C. Power Distribution Systems Shutdown . B 3.8 49 B 3.9 REFUELING OPERAT10N$ B 3.9.1 Refueling Equipment Interlocks . . . . . . . . . . . B 3.9 1
~(
t B 3.9.2 B 3.9.3 Refuel Position One. Rod Out Interlock ....... Control Rod Position . . . . . . . . . . . . . . . . B 3.9 7 B 3.9-4
\ B 3.9.4 Control Rod Position Indication .......... B 3.9 10 B 3.9.5 Control Rod OPERABILITY Refueling ........ B 3.9 13 B 3.9.6 Water Level - Reactor Pressure Vessel ...... B 3.9 15 B 3.9.7 Water Level . [$ pent Fuel Storage Pools) . . . . . . B 3.9 17 8 3.9.8 Residual Heat Removal - High Water Level . . . . . , B 3.9 19 8 3.9.9 Residual Heat Removal Low Water Level ...... B 3.9 22 B 3.10 $P(CIAL OPERATIONS B 3.10.1 Inservice Leak and Hydrostatic Testing Operation . . B 3.10-1 i B 3.10.2 Reactor Mode Switch Interlock Testing ....... B 3.10 4 l B 3.10.3 Control Rod Withdrawal Hot Shutdown ....... B 3.10 7 l B 3.10.4 Control Rod Withdrawal Cold $hutdown . . . . . . . B 3.10 10
! B 3.10.5 Control Rod Drive Removal Refueling ....... B 3.10 13 8 3.10.6 Multiple Control Rod Withdrawal Refueling .... B 3.10 16 B 3.10.7 Control Rod Testing Operating .......... B 3.10 18 l l l ~ ABWR iii 5/31/89 Amendment 9 16A.13
__ __ _ . _ ____~_ _ __ _. i M 21A610MP Se==dard Plant ma a 16A.2 B2.0 Safety Limits { O V l l Safety Limits ! i B 2.0 B 2.0 SAFETY LIMITS RAsts l INTRODUCTION The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radio- , active materials to the environs. Safety Limits are estab. i' lished to protect the integrity of these barriers during j normal plant operettons and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. i-Because fuel damage is not directly observable, a stepback approach is used to estabitsh a safety Limit such that the. 3 Minimum Critical Power Ratio (MCPR) is not less than the limit specified in $pncification 2.1.2. MCPR grsater than : the specified limit represents a conservative margin relative
- to the conditions required to maintain fuel cladding integrity.
The fuel cladding is one of the physical barriers which separate the radioactive materials tros the environs. The integrity of this cladding barrier is related tc its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source (' is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thersal ' stresses which occur from reactor coeration significantly above design conditions. While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition bolling (i.e., MCPR of 1.0). These conditions represent a significant departure from the condition intended by design for planned operation. The MCPR fuel cladding integrity Safety Limit assures that during normal operation and during anticipated operational occurrences, at least 99.9% of the fuel rods in the core do not expertence transition boiling. FUEL CLADD]NG GE critical power correlations are applicable for all INTEGRllY critical power calculations at pressures at or above 785 psig (2.1.1) or core flows at or above 10% of rated flow. For operation , at low pressures and low flows another basis is used as follows: 7 (continued) f ABWR B 2.0 1 $/31/89 O Amendment 9 IM 2*I
ABWR :=mr : Standard Plant wi t G: !
$4fety Limits B 2.0 '
RAlts leontinuedt i FUEL CLADDING Since the pressure drop in the bypass regten is essentially ! INTE$RITY all elevation head, the core pressure drop at low power and i (2.1.1) flows will always be greater than 4.5 pst. Analyses show (continued) that with a bundle flow of 28 x 208 lbs/hr. bundle pressure drop is nearly ineependent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 pst drivino head will be greater than 28 m 108 lbs/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate t that the fuel assembly critical power at this flow is approximately 3.35 left. With the design peaking factors, this corresponds to a THEllMAL POWER of enore than 50% of RATED THEllMAL POWER. Thus, a THEllMAL POWER limit of fl% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative. MINIMUM The fuel cladding integrity Safety Limit is set such that no CRITitAL POWER fuel damage is calculated to occur if the limit is not ' i, RATIO violated. $1nce the parameters which result in fuel damage (2.1.2) are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate belling have been used to mark the beginning of the region where fuel damage could occur. Although it is - recognized that a departure from nucleate boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been
- adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transttion considering the power distribution within the core and all uncertainties.
The Safety Limit MCPR is determined using a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power. The probsbtlity of the occurrence of boiling transition is determined using the approved General Electric Critical Power correlations. Details of the fuel cladding integrity Safety Limit calculation are given in Reference 1, Reference 1 includes a tabulation of the uncertainties used in the determination of the Safety Limit MCPR and of the nominal values of the parameters used in the Safety Limit MCPR statistical analysis. (continued) ABWR B 2.0 2 5/31/89 N2 9 Amendment 9
. - _ .__ . . - ~
M 21A6100AP Standard Plant ui O Safety Limits f B 2.0 ' Mitt (continued) t REAC10R VE$SEL With fuel in the reactor vessel during periods when the i WATER LEVEL reactist is shutdown, consideration must be given to water , (2.1.3) level requirements due to the effect of decay heat. If the ' w.tter level should drop below the top of the active - irradiated fuel during this period, the ability to remove , decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad
- perforation in the event that the water level became less than two thirds of the core height. The Safety Limit has ,
been established at the top of the active irradiated fuel to provide a point which can be monitored and alto provide adequate margin for effective action. REACTOR The Safety Limit for the reactor steam done pressure has been
$1EAM DOME selected such that it is at a pressure below which it can be PRESSURE shown that the integrity of the system is not endangered.
(2.1.4) The reactor pressure vessel is designed to Section !!! of the ASME Boiler and Pressure Vessel Code (1971 Edition, including Addenda through Winter 1972), which permits a maximum t pressure transient of 110%, 1375 psig, of design pressure ' 1250 psig. The Safety Limit of 1325 psig, as measured by the reactor steam done pressure indicator, is equivalent to 1375 psig at the lowest elevstion of the reactor coolant system. The reactor coolant system is also designed to Section 111 of the ASME Code and is similarly protected by the 1325 psig ( Safety Limit. REFERENCES 1. ' General Electric Standard Application for Reactor Fuel,' NEDE 240ll P A, September 1988. < ABWR E 2.0 3 5/31/29 LO Amender.ent 9
+,_ -
ABM . 23A6100AP Standard Plant Rn A ,)
- SECTION 16AJ -
qy
' 'u CONTENTS :
4 Section Dk Eggt 16A.3.1 R3.0 Unmitime Candialems tw Omarmtlan 16A.31 ; 16A.3J R3.0 Survellance Requ!mments 16A.3 7 4 A w 16A= . i
' Amend;nent 9 1
# pg y., y - -...e, - - , ,
~'
hh .
' 21A6100AP
' Standard Plant - m.a .
.16A.3 B3.0 Applicability
,-R
) 16A.3.1 ilmittag Conditions for Operath.a
~( V .
LC0 Applicability 8 3.0 8 3.0 APPLICABILITY LIMITING CONDITIONS FOR OPERATION (LCOs) aASES The LCO 3.0's are general requirements which apply to Section 3 0 timiting Conditions for Operation ar.d Surveillance Requirements and apply at all times, unless otherwise stated. LCO 3.0.1 LC0 3.0.1 establishes the Applicability statement within each individual specification as the requirement for when which MODES or other specified conditions) conformance(i.e., to the in LCO is required. LCO 3.0.2 The Required Actions establish those remedial measures that must be taken within specified Completion Times when the- < requirements of an LCO are not met. The Completion Times of .> the Required Actions are applicable from the point in time it. ; is discovered that the Condition erists and is entered and the l- Applicability of the LCO is set. Only one Condition in each i LCO's ACTIONS may be entered at any time, unless otherwise A: stated (i.e., ' Conditions...may be concurrently applicable'). f
-]' If the Limiting Condition for Operation is met prior to expir-ation of the specified Completion Time (s , completion of the Required Action is not required, unless o)therwise stated (i.e.,
l
' Required Actions...sust be completed whenever this Condition is entered').
The Completion Times of the Required Actkas are also applicable when a system or component is vemoved from service intention. allw. The reasons for intentionally relying on the ACTIONS include, but are not limited to, performance of Surveillance Reouirements, preventive maintenance and correctivo maintenance. It is not intended that this intentional entry into ACTIONS be made for operational convenience. This intentional entry shall be under appropriate administrative control. This is to I!mit voluntary removal of redundant equireent from service in lieu if other alternatives that would nit result in redundant equipment being inoperable, thus iiniting the time both subsys-tems/ trains of a safety function are inoperable and limiting the time other conditions exist which result in LCO 3.0.3 being entered. L (continued) ABWR B31 5/31/89 Amendment 9 16A.3-1 s n -- - . - . , . - - -i-..-.-._.--_- a
ABWR .
==r St.mndard PlaM w E
~
LC0 Applicability B 3.0 RASES fcontinued) LCO 3.0.3 LC0 3.0.3 establishes the actions that must be implemented when an LCO is not met and:
- 1. An associated Required Action and Completion Time is '
g not met and no other Condition applies, or
- 2. The Condition is not specifically addressed by the associated ACTIONS.
l The purpose of this Specification is to delineate the time limits for placing the unit in a safe MODE or other specified l condition when operation cannot be maintained within the limits for safe operation defined by the LC0 and its ACTIONS. - It is , not intended to be used as an operational convenience to permit (routine) voluntary removal of redundant systems or components j from service in lieu of other alternatives that would not { result in redundant equipsut being inoperable. I one hour is allowed to prepare for an orderly shutdown before , i i initiating a change in plant operation. This time permits the ' operator to coordinate the reduction in electrical generation with the load dispatcher to ensure the stability and avail-ability of the electrical grid. The time limits specified to reach lower MODES of operatim permit the shutdown to proceed in a controlled and orderly manner that is well within the . capabilities of the facility assuming only the minir'Am required equipment is OPERABLE. This reduces the thermal stresses on components of the Reactor Coolant System and the potential for ij a plant upset that could challenge safety systems under condi-tions for which this Specification applies. The use and , interpretation of the specified times to complete the actions I of LCO 3.0.3 shall be consistent with the discussion of ' Section 1.3, Completion Times, a unit shutdown required per LCO 3.0.3 may be terminated and LCO 3.0.3 exited if any of the following occurs:
- 1) The LCO is now met.
- 2) A Condition exists for which the Required Actions have now been performed, i
- 3) ACTIONS exist which do not have expired Completion l Times. These Completion Times are applicable from ti.e point in time there was a failure to meet that LCO and not from the time LCO 3.0.3 is exited.
l L l (continued) 4' 4 ABWR 9 3-2 5/31/89 O; ; I Amendment 9 M2 l l
. _ _ __ . . _ _ . _ ~ _ _ _ . .
MM 23A6100AP Shandard Plant P LCO Applicability B 3.0 BASf5 (continued) LCO 3.0.3 The requirements of LC0 3.0.3 do not apply in MODES 4 and 5 I (continued) because the plant is already in the most restrictive Condition that LCO 3.0.3 would require the pl6nt to be placed in. The requirements of LCO 3.0.3 do not apply in other specified conditions in the Applicability, unless in MODES 1, 2 or 3, because the ACTIONS of individral specifications sufficiently define the remedial measures to be taken. The exceptions to LCO 3.0.3 are provided in instances where requiring a plant shutdown in accordance with LCO 3.0.3 would , not provide appropriate remedial measures for the associated
. condition of the unit. An example of this is in LCO 3.9.7, Water Level Spent fuel Storage Pool. LC0 3.9.7 has a..
Applicability of 'When irradiated fuel assemblies are stored in the Spent Fuel Storage Pool'. Therefore, this LCO can be applicable in any or all MODES. If the LCO and the Required Actions of LCO 3.9.7 are not met while in MODES 1, 2, or 3, there is no safety benefit to te gained by placing the plant in a shutdown condition. The Required Action of LC0 3.9.8 of
' Suspend further movement of fuel assembiles over the Spent Fuel Storage Pool" is the appropriate Required Action to attempt to complete in lieu of the ACTIONS of LCO 3.0.3.
L LCO 3.0.4 LCO 3.0.4 establishes limitations 33 MODES or other specified conditions in the Applicability changes when an LCO is not met. It precludes placing the unit in a different MODE or other specified condition when the requirements of an LCO in the MODE or other specified condition to be entered are not met and ~ continued noncompliance with these requirements would result in
~
a shutdown to comply with the Required Actions if a change in 1-MODES or other specified conditions was permitted. Compliance with Required Actions that permit continued operation of the unit for an unlimited period of time in the MODE or other specified condition provides an acceptable level of safety for continued operation without regard to the status of the plant before or after the MODE change. Therefore, in this case, entry into a MODE cr other specified condition in the [ Applicability may be made in accordance with the provisions of I the Required Actions. The provisions of this Specification should not, however, be interpreted as endorstrg the failure to exercise good practice in restoring systems or components to OPERABLE status before plant startup. The provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability which are l required to comply with ACTIONS.
~~
(continued) ABWR B33 5/31/89 l b
-\ j l
Amendment 9 6AS3
. - . . . - . . . -- _. ~ . - . --- __ - . . _ _ _ - . .
Mg~ Shadard Plant u ] i l 1 LCO Applicability B 3.0 ggts (continued) 1C0 3.0.5 LC0 3.0.5 establishes the allowance of restoring equipment to I service under administrative controls when it has been removed I from service or declared inoperable to comply with ACTIONS. I The purpose of this Specification is to provide an exceptica to j LC0 3.0.2 to allow the performance of Surveillance Requirements- i to 1) demonsttsie the OPERABILITY of the equipment oeing returned to service, or 2) demonstrate the OPERABILITY of other i equipment. I An example of demonstrating the OPERABIL)TY of the equipment beier returned to service is the re opening of a containment isolation valve that has beer closed to comply with Required-Actions, and must be re opened to perfore the Surveillance j Requirements. 1 An example of demonstrating the OPERABILITY of other equipment is the taking of an inoperable channel or trip system out of the tripped condition to prevent the trip function from occur-ring during the performance of a Surveillance Requirement on another channel in the other trip system. Another similar example of demonstrating the OPERABILITY of other equipment is the taking of an inoperable channel or trip system out of the tripped condition in permit the logic to function and indicate the appropriate response during the perfomance of a Survell-lance Requirement on another channel in the same trip system. LCO 3.0.6 LCO 3.0.6 establishes the al % ance of delaying Required . Actions for up *.o B hours to perfore Surveillance Requirements I when systems are rendered inoperable for the performance of the l Surveillance Requirements. The purpose of this Specification is te provide an exception to LCO 3.0.2 to allow performance of Surveillance Requirements required to demonstrate OPERABILITY, i (continued) ABWR B34 5/31/89 0
. Amendment 9 16A.34
4 TABWR m. Standard Plant w
\.
LCO Applicability i B 3.0 1 BASES feontinued) 1 LCO 3.0.7 LCO 3.0.7 establishes which ACTIONS are applicable when support , systems are inoperable, depending on whether or not they have < en LCO specified in the Technical Specifications. The sup- J ported system is not required to be declared iaooerable solely due to support system inoperability. Only the support system LCO's ACTIONS are required to be entered. This is a clarifica-tion of the definition of OPERABILITY but is necessary to establish the relationship between the support systems and the supported system in order to to preclude cascading to multiple supported system ACTIONS and to eliminate the confusion associ- , ated with entering multiple LCOs' ACTIONS. Examples of support ' systems with LCOs specified in the Technical Specifications include cooling / service water, diesel generatois and AC and DC distribution When a support system is inoperable and there is not an LCO for that support systen specified in the Technical Specifications, the licensee shall evaluate the impact of the inoperability or degradation of the support system function on the OPERABILIT) of the suppcrted system. This is b6cause the inoperability or I degradation of the support system function may or may not affect the OPERABILITY of the supported system, depending upon l the intended function of the supported system and the level of
.4 / support the support system proddes. Upon determination that l
the supported system is inoperable, the ACTIONS of its LCO (~)i - r shall apply. LCO 3.0.8 There are certain special tests and operations required to be performed at various times over the life of the plant. These special tests and operations are necessary to demonstrate select plant performar.ce charactcristics, to perform special maintenance activities and to perform special evolutions. Special Operations LCOs in Section 3.10 allow specified Technical Specifications requirements to be changed to permit performance of these special tests and operations which otherwise could not be perfomed if required to comply with the requirements of these Technical Specifications. Unless otherwise specified, all the other Technical Specification requirements remain unchanged. This will ensure all appropriate-requirements of the MODE or other specified condition not directly associated with or required to be changed to perform the special test or operation will re nin in effect. (continued) ABWR B 3.s 5/31/89 bv Amendment 9 16A.33
,, LABWR 23, ,
Standard Plant .-
- m. u i
e: I LC0 Applicability B 3.0 . B _ASES fcontinued) LCO 3.0.8 The Applicability of a Special Operations LCO represents a (continued) conditio9 not necessarily in compliance with the normal .1 requirements of the Technical Specifications. Compliance with 5pecial Operations LCOs is optional. A special operation may be performed either under the provistor.s of a.he appropriate Special Operatians LCO on the other appitcable Technical 1 Specification requirements. If it is desired to perfom the ) special operation undet the provisions of the $r 'Jpera- , tions LCO, the requirements of the So* tal Oper LCO shall be followed. When a Special Operat' LCO regt 9.e ) LCD to be met, only the requiremer' he LCO ... . .n 1 be met regardless of that LCO's & -1. $.. .;ould u requirements of the other LCO no' # 25 of the , Special Operations LC0 apr'", av he other LCO. ' However,' there are insta dei *
< rations LC0 i ACTION may direct the ot' 3s * . . The Survell- l .
lances of the other LCO . rew. . wiess ] -l otherwise specified P 'ial s:itions ; may exist such that t' -ilit, . is met and all that LCO's rer are r 0 . eencur-rent with the requirec ) Special (- . . 1 O i 3 1 I h (- i L I I L l l' \ l l l 1 1 l l I l ABWR f 1-6 5/31/89
. O Amendment 9 16A.3-6 2 1-l 1
MM moorp Standard Plant m 16A.3.2 Survellance Regciremients I
'\
SR Applicability B 3.0 i B J.0 APPLICABILITY SURVEILLANCE REQUIREMENTS (SEs) t 3MU The SR 3.Os are general requirements which apply to Section 3.0 Limiting Conditions for Operation and Surveillance Requirements and apply at all times, , unless otherwise stated. SR 3.0.1 SR 3.0.1 establishes the requirement that Surveillance Requiremerts must be performed during the MODES or other specified ;onditions in the Applicability of the LCO unless otherwise stated in an individual Surveillance Requirement. The purpose G this Specification is to ensure that Surveil-lance Requit. Nnts are performed to verify the OPERABILITY of systems and u nponents and that parameters are within specified limits. Systes and components are assumed to be OPERABLE when Surveillances have been met within the specified Frequency. However, nothing in this provision is to be construed as implying that systems or components cre OPERABLE when they are known to be inoperable although still meeting the Surveillance Requirements. Surveillance Requirements do not have to be performed when the unit is in a MODE or other specified condi-I tion for which the requirements of the associated LCO do not apply, unless otherwise specified. An example of this is a Surveillance Requirement whose LC0's Applicability is MODES 1, 2, and 5. and requires a CHANNEL FUNCTIONAL TEST t,nce during each MODE 4 of greater than 24 hcdra duration. Surveillance Requirements do not have to be performed on inoperable equipment because the ACTIONS define the remedial measures that apply. However, the Surveillance Requirements have to be met as required by SR 3.0.2 prior to restoring equipment to OPERABLE status. Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This includes meeting applicable Surveillance Requirement $ in accordance with SR 3.0.2. Post maintenance testing may not be
~,
possible in the current MODE or other specified conditions in the Applicability due to the necessary unit parameters having not been established. In these situations, the equipment may be considered OPERABLE provided testing has been satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performing its function. This will allow operation tc proceed to a MODE or other speci-fied condition where other necessaiy post maintenance tests can be completed. (continued) ABWR B 3-7 5/31/89 l
)
Arnendment 9 16A.3-7
J .
' 23A6100AP Standard Plant w.
- e SR Applicability B 3.0 l
MRS hontinunfi 8 h SR 3.0.1 Some examples of this process are: l'- (continued)
-- CR0 meintenance during refueling which requitas scram testing at > 950 psi. However, if other appropriate testing is satisfactorily completed and the scrar. time testing of SR 3.1.3.3 is satisfied, the control rod can be considered OPERABLE. This allows startup to proceed to reach 950 psi to perform other necessary testing.
- RCIC maintenance during shutdown which requires system functional tests at a specified pressure. Provided other appropriate testing is satisfactorily completed, startup can proceed with RCIC considered OPERABLE. This allows operation to reach the specified pressure to complete the necessary post maintenance testing.
SR 3.0.2 SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillai.ces and any Required Action with a Completion Time requiring the periodic performance of t, the ACTION on a 'once per...' interval. $R 3.0.2 permits an I extension of the interval associated with the stated Frequency [ to facilitate Surveillance scheduling and for consideration of 1, plant operating conditions that may not be suitable for l conducting the Surveillance; e.g., transient conditions or i other ongoing Surveillance or maintenance activities.
~< The 25% extension does not significantly degrade the reliability which results from performing the Surveillance at ;
its specified frequency. This is based on engineering i judgement and the recognition that the most probable result of I any particular Surveillance Requ'.nement being performed is the verification of conformance with the Surveillance Requiremeats. The exceptions to SR 3.0.2 are those Surveillances where the 1.25 times the interval specified in the Frequency does not j != apply. An example of where SR 3.0.2 does rot apply s SR ; 3.6.1.1.2 whose surveillance is ' Perform required T /pe A leak j L rate testing in accordance with 10 CFR 50 Appendix J and i approved exemptions
- and whose Frequency is "In accordance with t
10 CFR 50 Appendix J and approved exemptions." The l requiroments of regulations take precedence over the Technical i l Specifications. The Technical Specifications cannot in and of i themselves extend a test interval specified in the regulations. Therefore, SR 3.6.1.1.2 has a note in the Frequency stating r
" Provisions of SR 3.0.2 are not applicable."
l 1 I (continued) J ABWR B38 5/31/89 i. Amendment 9 16A.3-8 e) ' l 1 l 1
--. , , . - - , , ---.-n,.n- , . , , - - - - + - - - . ,
1 23A6100AP Standard Plant - _ m 1 v SR Applicability B 3.0 l PASESfcontinued) SR 3.0.3 SR 3.0.3 establishes the flexibility to defer declarine the j equipmeet inoperable when Surveillances have not been completed , within the specified Frequency. An allowed 24 hour deferral applies from the point in time that it is discovered that the . Surveillance has not been performed in accordance with SR 3.0.2 ; and not at the time that the specified Frequency was not met. This deferral provides an adequate time limit to complete Suiveillances that have boon missed. The purpose of this illowance is to permit the completion of a Surveillance before compliance with Required Actions or other resiodial measures would be required that may preclude completion of a Surveil-lance. The basis for this allowance includes consideration for plant condittons, adequate planning, availability of personnel, the time requfred to perform the Surveillance and the safety 1- significance of the delay in completing the requited Surveil-l arice. This provision also provides a time limit for the completion of Surveillance Requirements that become applicable as a consequence of MODE changes imposed by Required Actions ,' and for comil eting Surveillance h q;!rements that are applic. eble when an u ception to the requirements of SR 3.0.4 is allowed. If a Surveillarce is not completed within the 24 hour allowance, the equipment is considered inoperable and the -
- (fm) Completion limes of the Required Actions begin insediately upon
(,/ expiration of the 24 h:ur allowance, if a Surveillance is failed within the 24 hour allowance, the equipment is considered inoperable and the Completion Times of r' the Required Actions begin tamediately upon the failure of the i Surveillance. SR 3.0.4 SR 3.0.4 establishes the requirement that all applicable Surveillance Requirements must be met before et.try into a MODE or other specified condition in the Applicability. The purpose of this specification is to er.sure that system and con.3cnent OPEpABILITY requirements or parameter limits are met be' ore entry into a MODE or other specified conditict' in the l Applicability for which these systems and components ensure ! safe operation of the facility. This provision applies to changes in MODES or other specified conditions in the Appitcability associated with plant shutjown as well as startup. The provisions of SR 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability which are required to comply with ACTIONS.
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(continued) ABWR B 3-9 5/31/89 1 O A=nendment 9 16A3 9
23A6100AP
- Standard Plant m O
SR Applicability B 3.0 RAsts feontinued) SR 3.0.5 SR 3.0.5 establishes the requirement that inservice inspection
- of ASME Code Class 1, 2 and 3 comporents and inservice testing of ASME Code Class 1, 2 and 3 pumps and valves shall be tar.
formed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55a. These requirements apply except when relief has been requested pursuant to 10 CFR 50.55a (g)(6)(1). This Specification includes a clarification of the Frequencies for performing the inservice inspection and testing activities required by Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda. This clarification is provided to ensure consistency in Surveillance Frequencies throughout the Technical Specifications and to remove any ambiguities relative to the frequencies for performing the required inservice inspection and testing activities. Urder the terms of this
' Specification, the more rettrictive requirements of the Techni.
cal Specifications take precedence, over the ASME Boiler and Pressure vessel Code and applicable Addenda. The requirements of SR 3.0.4 to perfors Surveillance activities before entry into a MODE or other specified condition in the Applicability tekes precedence over the ASME Boiler and Pressure Vessel Code-provision wb!ch allows pumps and valves to be tested up to one week after return to normal operation. Also, the Techrocai Specificatica defiriition of OPERABILITY does not allow a grace period before a component that is not capable of performing its specified function is declared inoperable. The Technical Specifications take precedence over the ASME Boiler and Pres-sure Vessel Code provision which allows a valve to be incapable of performing its specified function for up to 24 hours before
, being declared inoperable.
t f ABWR B 3 10 5/31/89 i O Amendment 9 ' 16A.3-10 i
4 a M 23A6100AP Standard Plant hA SECTION 16A.4 O cours"rs . l SectiOR M EAgt 16A.4.1 B3.1.1 Shutdown Mar #m 16A.41 16A.4.2 53.1.2 Centrol Rod Operabillir 16A.4 7 16A.43 R3.13 Control Rod Scram 'Hanns 16A.414 16A.4.4 B3.1.4 Control Rod Scram Amumulators 16A.420
- 16A.4.5 B3.1.5 Control Rod Drive Compting and 16A.4 24 16A.4.6 B3J.6 Rod Pattern Control 16A.4 27 16A.4.7 B3.1.7 Standhv Linuld Control System thA.4 31 l
O . W 16A.4-ii t Amendment 9 11______ . _ _ , _ _ _ _ . _ _ _ _ _ _ _ . . . _ , . . . . . , _ . , _ . . _ . . .
LAB M . 3346oorr j h=Aard Plant Rm A l 16A.4 B3.1 ReactMty Control Systems 16A.4.1 R3.1.1. Shutdown Margin SHUTDOWN MARGIN . B 3.1.1 8 3.1 REACTIVITY CONTROL SYSTEMS B 3.].) IMUfDOWN MAR &lN Basts - , t BACKGROUND SHUTDOWN MARGIN is specified to ensure:
- a. The reactor can be made suberitical from all operating ,
conditions.
- b. The reactivity transients associated with postulated accident conditions are controllable within acceptable Itaits.
- c. The reactor will be maintained sufficiently suberitical te preclude inadvertent criticality in the shutdown condition.
General Design Criterion (GDC) 26 requires the reactivity control systema be capable of holding the core subcritical under cold conditions. ' A n t A ( V fHUTDOWN MARGIN is an explicit assumption in several of the ! APPLICABLE SAFETY evaluations in SSAR Chapter 15, Accident Analyses. SHUTDOWN , ANALYSES MRGIN is assumed as an initial condition for the Control Rod Removal Error During Refueling (Ref. 1) accidents. The t- analysis of these reactivity insertion events assumes the re veling interlocks are OPERABLE when the reactor is it the l REFUELING mcde of operattom. 'Thase interlocks preveat the L withdrawal of more than one mirol rod, or control rod pair, from the core during refueling. (Special consideratica and l. . requirements for multiple control rod withdrawal durin refuel. I ing are covered in Special Operations LC0 3.10.6, Mult 31e ' l Control Rod Withdrawal - Refueling.). The analysis assunes this condition is acceptable since the core will be shutdown with the highest worth control rod pair withdrawn, if adequate SHUIDCWN MARGIN his been demonstrated. Prevention or mitigation of reactivity insertion events is necessary to limit energy deposition in the fuel to prevent significant fuel damage which could result in endue release of radioactivity (see Bases for LCO 3.1.6). Adequate SHUTDOWN MARGIN ensures inadvertent criticalities and potential control rod withdrawals involving high worth control rods will not cause significant fuel damage. (continued) i ABWR B 3.1-1 5/31/89 Amendmeat 9 16A.41 i. l
- ABWR zwioare Standard P: ant - , hA SHUTDOWN MiRGIN B 3.1.1 BASES feontinued)
APPLl:ABLE SetVTDOWN MRGIN satisfies the requirements of Selection SAFETY Criterior. 2 of the NRC Interim Policy Statement on Technical ANALYSES Specification Improvements as documented in Reference 4. (continJed) LCO The spe:ified SHUIDOWN MA9 GIN accounts for the uncertainty in the demonstration of SHUTDOWN MRGIN thereby ensuring the reactor can be held subcritical during shutdown canditteis and during refueling with the highest orth control red pair withdrawn. Separate SHUTDOWN %RGIN limits are pra.ided for demonstrations where the highest worth control rod pstr is determined analytically or by measuremeit. This ir due to the reduced uncertainty in the SHUTDOWN MRGIN demonstration w;&n the highest worth control rod pair is determined by meas'na-ment. To assure adequate SHUTDOWN M RGIN, an additional design margin is included in the design process to account for uncertainties in the t;esign calculations (Ref. 2). APPLICABI'. ',TV During MODES I and 2. the SHUTDOWN MRGIN specification is applicable because suberiticality with the highest worth conti01 rod pair withdrawn is assumed in the analysis. Also, the capability to reach MODE 4 conditions from any initial stats is required by GDC 26. For operation in MODES 3 and 4, the SHUTDOWN MARGIN specification is required to ensure the reactor will be held subcritical with margin for a single stuck control rod; The SHUTDOWN MARGIN specification is applied to MODE 5 to prevent an open vessel, inadvertent criticality curing the withdrawal of a single control rod from a core cell containing or.e or more fuel assemblies or of a control rod pair
.from loaded core cells during scram time testing.
(continued) r ABWR B 3.1-2 5/31/89 O Amendment 9 16A.4 2
21A6100AP. Standard Plant _
- m. a ;
1 j l l
. ,K 1
i SHUTDOWN MP. GIN j B 3.1.1 j 1 Be$fS fcontinued)
% MNS U During M001 1 or 2 failure to meet the specified SHUTDOWN mar *IN may be caused by a control rod that cannot be inserted.
Because the reactor can still be shutdown assuming no failures of add aional control rods to insert, operation is allowed to continue for a short time to allow restoratten of SHUTDOWN MARGIN. M If the SHUTDOWN MARGIN cannot be restored promptly, the reactor must be in MODE 3 (i.e. insert all insertable control rods) to prevent the potential for furthtr reductions in available SHUTDOWN MARGIN (e.g. additional stuck control rods). L1 With SHUTDOWN hARCIN less than tpecified in MODE 3 the operator mast inser t all insertable control rods. This action results in the lea:t reactive conditiTn for the core. D.I. D.2. D.3. D.4
'.ith SHUTLOWN MARGIN less than specified in MODE 4, the V operator wst insert 7" insertable control rods. This action results .n the least re.M tive condition for the core, f.ctions are also taken to provide means for control of potential radioactive releases caused by an inadvertent reactivity excursion. This includes en:uring Secondary Contains.ent is OPERABLE (LC0 3.6.4.1), at least one Standby Gas Treatment System (SGTS) subsystem is OPERA 0LE (LC0 3.6.4.3) and at least one Secondary Containment Isolation Valve (LCO 3.6.4.2) a-d 1- associated actuation instrumentation is OPERABLE in each I
associated penetration not isolatet. Ensuring tha OPERABLE i status of the comoonents involves administrative checks, examining logs or ciner information, to determine if the components are out of service for maintenance or other reasons. It does not require performing surveillan:es needed to demon-strate the OPERABILITY cf the components. If however, any required component is inoperable, then it must be restored to OPERABLE status, in this case, surveillance requirements may need to be performed to restore the component to OPERABLE status. (continued) ABWR B 3.1-3 5/31/89 Amendment 9 16A.4 3 i
, ., ,-n -
,n_ .. _ , _ . - - . _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ . _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - - - _ _ _ _
j
- MM
' Standard Plant 2146100AP j Rw. A O
SHUTDOWN MARGIN B 3.1.1 4 BA5tS fcontinued) l ACTIONS E.1. E.2. E.3. t.4. [.5_ 1 (continued) ' With SHUTDOWN MARGIN less than a cified in MODE 5 the , opere or must suspeno CORE ALTERATIONS since these activities I could reduce SHUTDOWN MARGIN. This involves the movement of fuel in the core or the withdrawal of contrcl rods. The requirement that the activities be suspended immediately is not intende$ to prohibit the completion of an action that would improve or not affect SHUTDOWN MARGIN (e.g. complete the ir.sertion of a control rod containing or.e or more fue). All control l asserblies rods must in core cells be ir;erted to pl6ce the core in the least reactive condition Control rods in core cells with no fuel assemblies are not required to be e inserted since they have a negligible irpact on core reactiv. ity. Actions are aise taken to provide means for control of potential radioactive releases c:.used by an inadvertent reac-tivity excursion. This includes ensuring Secondary Containment is OPERABLE (LCO 3.6.a.1), at least one Standby Gas Treatment System ($GiS) subsystem is OPER*BLE and at least one Secondary Containment Isolation Valve (LCO 3.6.4.2) and associated actuation instrumentation is OPERABLE in each associated penetration not isolated is descr? bed in the Bases for Required Actions D.2 D.3 antt D.4. EgBQlttion Tinn > All Completicn Times are based on industry accepted practice ard engineering judgement considering the number of available . systems and the time required to reasonably ;owplete the L Required Actions. SURVEILLANCE SR 3.1.1.1 {
- REQUIREMENTS l- ;
' Adequate SHUTDOWN MARGIN sust be demonstrated for the entire ' l cycle length and must be performed before or during the first startup following CORE ALTERATIONS which ir.volve changes in the core reactivity. Since core reactivity will vary during the cycle as a function of fuel depletion and poison burnuo, the beginning of cycle comonstration must also account for changes i in core reactivity during the cycle. Therefore, the specified l SHUTDOVN MAkGIN srJst be increased by a factor, R, which is the L '~ difference between the calculated value of maximum core reactivity during the operating cycle and the calculated beginning of cycle core reactivity. If the value of R is negative (that is, beginning of cycle is the most reactive point in the cycle), no correction to the beginning of cycle j SHUTDOWN MARGIN is required (Ref. 3). (continued) ABWR B 3.1-4 5/31/89 O Amendment 9 I
~'n. OS nusaw - i
. Standarti Plant _ m 1
,,m .-
4 (, ' l 1 I SHUTDOWN MARGIN B 3.1.1 l BASES fcontinued) SUR,EILLANCE ' SR 3.1.1.1 (continued) REQUIREMENTS (cintinued) The SHUTDuWN MARGIN may be demonstrated during an in sequence control rod withdrawal ir. which the highest worth control rod ;
' pair is analytically determined or during local criticals, where the highest worth control rod pair is determined by testing. Local critical tests may also be performed, but tequire the withdrawal of out of sequence (i.e., adjacent) control rods. This testing would therefore require bypassiag of the rod pattern c)ntrol systems to allow the out-of* sequence withdrawal and additional requirements must be met (see LCO 3.10.7, Control Rod Testing Operatins).
Four hours after reaching criticality is provided to allow a reasonable time te perform the required calculations and have appropriate verification.
}R 3.1.1.2 During MODE 5 adequate SHUTDOWN MARGIN is also reautred. An evaluation of each fuel movement during fuel loading shall be performed to ensure adequate SHUTDOWN MARGIN is a intained during refueling. This ensures the intermediate loading
/' patterns are bounded by the safety analyses for the final core t loading pattern. For example, bounding analyses which N demonstrate adequate $HUTDOWN MARGIN for the most reactive configurations during the refueling may t,e performed to demonstrate acceptability of the entire fuel movement sequence.
Spiral offloao/ reload sequences inherently satisfy the surveillance requirement provided the fuel assemblies are reloaded in the same configuration analyzed for the new cycle. i Removing fuel from the core will always result in an increase in $HUTDOWN MARGIN. Surveillance Frecuencies In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the system / component status. (continued) l' l A0WR B 3.1 5 5/31/89 l. i Amendment 9 16A 4-5
. . . . .. ~. . ~. - - . . . . . . - . -.
h g3g /' Standani Plant m , 4 i ei SHU1DOWN MARGIN B 3.1.1 RASES feontinued) REFERENCES 1. ASWR $$AR, Section 15.4.1.1. l t. A>WR SSAR, Section 4.3.2.4.1,
- 3. NEDE 240ll P A 9, ' General Electric Standard Appitcation I for Reload Fuel', September 1988.
( 4. NEDO 31466, ' Technical Specification Screening Criteria Application and Risk Assest'wnt", November 1987. l i t i l i l O l-ABWR B 3.1-6 5/31/89 l O Amendment 9 16A.4-6 4
l INM 23A6100AP i Staadard Plant aa !; i 16AA.2 83.1.2 Control Rod Operability l Control Rod OPERABILITY B 3.1.2 B 3.1 REAR *.IVITY CONTROL SYST!MS B 3.1.2 Gentrol Rod OPERABILITY Baill BAi,KGROUND Control rods are components of the Control Rod Drive (rRD) system, which is the primary reactivity control system for the l- reactor. In conjunction with the 3eactor Protection System , l (RPS), the CR0 system provides the means for the reliable , control of reactivity changes to ensure under conditions of I -- normal operation, including anticipated operat'onal occur-rences, specified acceptable fuel design limits are not exceeded. in addition, the control rrds provide the capability , to hold the reactor core suberitical under all cenditions and to limit the potertial amount and rate of reactivity increase caused by a malfunction in the CRD system. The CRD system is designed to satisfy the requirements of General Design Criteria 26, 27, 28 and 2g. APPLICABLE The analytical methods and assumptions used in the evaluations l SAFETY involving control rods are presented in References 1, 2, 3 and ANALYSES 4. ~he control rods provide the primary means for rapid reactivity control (reactor scram), for maintaining the reactor vO suberttical and for limiting the potential effects of i reactivity insertion events caused by malfunctions in the CRD system. The capability to insert the control rods ensures the t assumptions for scram reactivity in the design basis transient and accident analyses are not violated. Since the SHUTDOWN MARGIN ensures the reactor WCl be subtritical with the strong. est cor. trol rod pair withdrm (assumed single failure of an HCU), the failure of an additional control rod to insert, if required, could invalidate the demonstrated SHUTDOWN MARGIN and potentially limit the ability cf the control r3d drive system to hold the reacter suberitical. Therefore, the requirement that all control rods are OPERAB E ensures the CRD system can perform its intended function. The control rods alsa protect the fuel from damage which could result in release of radioactivity. The limits protected are the Safety Limit MINIMUM CRITICAL N)WER RATIO (MCPR) (see Bases for LCO 3.2.2, MCPR), the 1% cladding p'astic strain fuel design limit (see Bases for LCO 3.2.1, AVERAGE PLANAR LINEAR
- HEAT GENERATION RATE) and the fuel daalage limit (see Bases for LCO 3.1.6, Rod Pattern Control) during reactivity insertion events.
(continued) ABWR B 3.1 7 5/31/89 O-U Amendment 9 16A.4-7 P
~.. .. . -- ABWR .
234 e r-Standard Plant au Control Roc OPERABILITY B 3.1.2 BASf5 fcontinued) APPLICABLE The negative reactivity insertion (scr19 provided by the CRD ; SAFETY system provides the anal /tical basis for determination of plant < iWALYSES . thermal limits and provides protection against fuel damage (continued) limits during a Rod Withdrawal Error (RWE) event. Bases for LCO 1.1.3 through LCO 3.1.6 discuss in more detail how the Safety Limits are pratected by the CRD system. Control Rod CTERABILITY satisfies the requirements of Selection Criterton 3 of the NRC Interim Folicy Sta " ment on Technical Specification leprovements as documented .. Reference 6. LCO OPERABILITY of an individual control rod is based on a combination of factors, primarily the scram insertion times, the associated control rod accumula%r status, the control rod coupling integrity and separation or ection capability, and the ability to determine the control rt position. Although not all control rods are required to be OPERABLE to satisfy the iatended reactivity control requirements, strict control over the number and distribution of inoperable control rods is required to satisfy the assumptions of the design basis transient and accident analyses. APPLICABILITY The contrul rod OPERABILITY requirements are applicable during MODES 1 and 2 whr..ever control rods may be withdrawn. In MODE 5, the OPERABILITY of withdrawn control rods is controlled by 100 3.g.5. In MODES 3 and 4, control rods are only allowed to be withdrawn under Special Operations LCO 3.10.3 (Control Rod Withdrawal - Hot Shutdown) and LCO 3.10.4 (Control Rod Withdrawal - Cold Shutdown) which provide adequate requirements for control rod OPERABILilY during these conditions. ACTIONS U A control rod is considered stuck if it will not insert by either FMCRD drive motor torque or by scram pressure. The failure of a control rod to insert during SR 3.1.2.2 or SR 3.1.2.3 alone, however, does not necessarily mean that the control rod is stuck, since failure of the motor drive would also result in a failure of these tests. Verification of a stuck rod can be made by attempting to withdraw the rod, if the motor is working and the rod is actually stuck, the travel-ing nut will back down from the bottom of the drive and a rod separation alarm and rod block will result (see LCO [ Rod Block (continued) ABWR . 3.1 8 5/31/89 O Ame:An 3: 9 16A.4-8
7 y' MM 23A6100AP Standard Plant m. a 4
'gs -
O , Control Rod OPERABILITY , B 3.1.2 BASES feontinued) _ , r ACTIONS M (continued) (continued), Instrumentation)). Conversely, if the motor drive is known to be failed, the rod is not necessarily inope-able since it is probably still capable of scram. However, at the next required performance of SR 3.1.2.2 or 3.1.2.3, there would be no way of , verifying insertability, except by scram. In this case, an ! individual scram shoLid be attempted. If the rod scrams, the ! rod is not stuck but should be considered inoperable and I bypassed in RC&ls since it cannot be withdrawn and a separation situation will exist until the motor is repaired and the traveling nut is run in to the full in position, if the rod fails to insert by individual scram, it should be considered stuck and the appropriate ACTIONS taken. The failure of a control rod pair to insert is assumed in the design basis j transient and accident analyses end therefore, with one with-drawn control rod stuck, some time is allowed to make the control rod insertable. With a fully inserted control rod stuck, no actions are required as long as the control rod remains fully inserted. As noted, a stuck control rod may bS bypassed in the Rod Control and Information System (RC&lS) to allow continued operation. LCO [ Control Rod Block Instrumen-
. tation) provides additional requirements when control rods are bypassed in RC&l5 to ensure compliance with the RWE analysis.
~
B . I . B . 2. B.3. BJ Vith one withdrawn control rod stuck for more than the allowed time, the cuntrol rod must be disarmed and isolated from scram pressure. The motor drive may be disarmed by placing the rod in RC&lS Bypass or by manually disconnecting its power supply. Isolating the control rod from scram prever.ts damage to the CRD and surr9unding fuel assemblies should a scram occur. The control rod can be isolated from scram by isolating it from its , associated hydraulic control unit. Two CRDs sharing an HCU can l be individually isolated from scram. Below 10% of RTP, the generic Banked Position Withdrawal sequence (BDWS, or Ganged Withdnwal Sequence Restrictions, GWSR, for ABWR) analysis requires inserted control rods not in compilance with B'WS to be separated by at le6st two OPERABLE centrol rods in all directions including the diagonal. Out of-sequence control rods may increase the potential reactivity worth of a control rod, or gang of control rods, during a RWE and therefore the distr wution of inoperable contrni rods must be controlled. (continued) ABWR B 3.1-9 5/31/89 i D ___ Amendment 9 16A.4 9
i
- 23A6100AP Standard Pin'nt no t
l-9 , r l- [ Control Rod OPERABILITY B 3.1.2 BASES feontinued) !- ACTIONS B.1. B.2. B.3. B.4 (continued) (continued) With a s' ingle control rod stuck tr. a withdrawn position, the i remaining OPERABLE control rods are capable of providing the requireo scram and shutdwn reactivity. Failure to reach MODE 4 is only likely if an additional control rod adjacent to the i stuck control rod also fails to insert during a required scram.
- Even with the postulated additional single failure of an adjacent control rod to insert, sufficient reactivity control e remains t4 reach and stintain MODE 3 conditions. Required l Action B.3 of LCO 3.1.2 performs a step test on each remaining iL withdrawn control roo to ensure that no additional control rods
- are stuck. Therefore, 72 hours is allowed to perform the l, analysis or test in Required Action B.4.
If the Required Actions and associated Completion Times of Condition B cannot be met, the reactor must be in MODE 3 within ( 12 hours. Insertion of the remainder of the control rods eliminates the possibility of an additional failurs of a control rod to insert. Prior demonstration of adequate l SHUTDOWN MARGIN ensures the reactor can be held subcritical l with only a single control rod withdrawn. ~ l D.1. D.2 l l With more than one withdrawn control red stuck, the stuck ! control rods should be disarmed and isolated from scram pres-sure as described in the Bases for Required Action B.) and the ' reactor must be in MODE 3 within 12 hours. The occurrence of
- more than one control rod without insertion capability may be i an indication of a generic problem in the control rod drive system that could potentially cause additional failures of I control rods to insert. Insertion of all insertable control rods eliminates the possibility of an additional failure of a
- j. control rod to insert.
i. (.1. E.2. f.3 l With less than or equal to 8 control rods inoperable for any of the reasons discussed in LCO 3.1.3 through LCO 3.1.6, operation may continue provided the control rods are fully inserted and disarmed (however, they do not need to be isolated from scram). Ir.serting a con'rol rod ensures the shutdrwn and scram capabilities are not adversely affected. The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations. The control rods can be disarmed by disconnecting power to the motor drive or by placing the rod in RC&lS bypass. Below 25% of RATED THERMAL POWER, the GWSR analysis requires (continued) ABWR B 3.1 10 5/31/80 4 Amendment 9 16A.4-10 -
ii iABWR . 2=>=^r Standard Plant Rn A J l Control Rod OPERABILITY S 3.1.2 jf$U fcontinued) ACTION $' D.1. D.2 (continued) With more than one withdrawn control rod stuck, the stuck control rods should be disarmed and isolated from scram pres-sure as described in the Bases for Required Action B.1 and the reactor must be in MODE 3 within 12 hours. Tbs occurrence of more than one cor. trol rod without insertion capability may be an indication of a generic problem in the control rod drive system that could potentially cause additional f ailures of control rods to insert. Insertion of all insertable control rods elir.inates the possibility of an additional failure of a control rod to insert. E.1. E.2. E.3 With less than or equal to 8 control rods inoperable for any of the reasons discussed in LCO 3.1.3 through LCO 3.1.6, operation may continue provided the control rods are fully inserted and disarmed (however, they do not need to be isolated from scram). Inserting a control rod ensures the shutdown and scram capabilities are not adversely affected. The control rod is disarmed to prevent inadvertent withdrawal during subsequent O operations. The control rods can be disarmed by disconnecting power to the motor drive or by placing the rod in RC&lS bypass, fd' Below 10% of RATED THERMAL POWER, the generic BPWS analysis requires inserted control rods, not in compliance with BPWS, to - be separated by at least two CPERABLE control rods in all directiois including the diagonal (Ref. 5). Inserted out of. sequence control rods may increase the potential reactivity worth of other e.ontrol rods during an RWE and therefore the nv*.ber and distribution of inserted inoperable control rods must be controlled. As noted, the control rods reay be bypassed ' in the RC&l$ if required to allow insertion of the inoperable control rods and cortinued operation. 1.lso, as noted, control rods declared inoperable with a failed motor drive can only be inserted Lj scram. Control rods with failed motor drives are not inoperable for this reason alone, but must be con-sidered so upon failure nf SR 3.1.;'.2 or SR 3.1.2.3, or when not in compliance with GWSR (see LCO 3.1.6). LCO [ Control Rod Block Instrumentation) provides additiot.al requiremen;s when the control rods are bypassed to ensure cocpliance with the RWE analysis. (continued) ABWR B 3.1-11 $/31/89 1i -
%./
Amendment 9 II j, ,
LABWR 22^61 MAP J Standard Plant Bsv A . , O Control Rod OPERABILITY B 3.1.2 BASES frontinued) , ACTIONS L,1 If the Required Actions and associated Completion Times of Condition E are not met or more than 8 inoperable control rods exist, the reactor is esquired to be in MODE 3 within 12 hours. This ensures all insertable control rods are inserted and - places the reactor in a condition that does not require the active function (i.e., scram or insertion) of the con *rol rods. The number of control rods permitted to be inoperable when operating above 10% of RATED THERMAL POWER could be more than the value specified, but the occurrence of a large number cf inoperable control rods could be indicative of a generic problem and investigation and resolution of the pot uttal problem should be done. Comoletion Times All Completion Times are based on industry accepted practice < 1 and engineering judgement considering the number of available systems and the time required to reasonably complete the Requir4d Action. SURVEILLANCE SR 3.1. 2.1 REQUIREMENTS Determining the position of each control rod is required to ensure adequate information on control rod position is available to the operator for determining CRD OPERABILITY and controlling rod patterns. Control rod position may be determined by the use of OPERABLE position indicators, by moving control rods to a position with an OPERABLE indicator, or by the use of other appropriate methods. , SR 3.1.2.2. SR 3.1.2.3 Control rod insertion capability is demonstrated by inserting each partially or fully withdrawn control rod at least one step and observing that the control rod moves. The control rod may then be re'urned to its original position. These surveillances are not required when below the actual LPSP of the RC&lS since the step insertions may not be compatible with the requirements of Rod Pattern Control (LCO 3.1.6) and the RC&l5 (LCO [ Control Rod Block Instrumentation)). Partially withdrawn control rods are not tested weekly because of the potential power reduction required to allow the control rod movement. (continued) ABWR B 3.1 12 5/31/89 O Amendment 9 16A.4-12
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- t. N) l I Control Rod OPERASILITY j B 3.1.2 ,
- l. BASES fcontinued)
SURVEILLANCE Surveillance Frecuencies RIQUIREMENTS (Continued) In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the i- unit conditions required to perform the test, the ease of , performing the test and a likelihood of a change in the system / component status. REFERENCES 1. NEDE 240ll P A, ' General Electric Standard Application for Reactor Fuel *, September 1988.
- 2. ABWR SSAR. Section 4.6.1,
- 3. ABWR SSAR, Section 7.7.1.2.
- 4. ABWR SSAR, Section 15.4.1.
- 5. NEDO 21231, ' Banked Position Withdrawal Sequer-?,",
January 1977, Section 7.2.
- 6. NEDO 31466,
- Technical Specification Screening Criteria p Application and Risk Assessment', November 1987.
ABWR B 3.1 13 S/31/89 v Amendment 9 16A.413
MM - 23A6100AP standard Plant n- A 16A.4.3 R3.1.3 Centrol Rod Scram 'thmes
- Control Rod Scram Times B 3.1.3 8 3.1 REACTIVITY CONTROL SYSTEMS
- B 3.3.3 Control Rod Scram Times t
BASES BACKGROUND The scram function of the Control Rod Drive (CRD) system
- reliably controls reactivity changes during abnormal opera.
tional transients to ensure specified acceptable fuel design limits are not exceeded (Ref. 1). The control rods are scrammed by positive means using hydraulic pressure exerted on
- the control rod drive piston.
APPLICABLE The analytical methods and assumptions used in evaluating the SAFETY control rod scram function are presented in References 2, 3, 4 ANALYSES and 5. The design basis transient and accident analyses assume all of the control rods scram at a specified insertion rate. The resulting negative scram reactivity forms the basis for the determination of plant thermal limits (e.g., MINIMUM CRITICAL POWER RATIO (MCPR)). Other distributions of scram times (e.g., several control rods scraming slower than the average time with several control rods scraming faster than the average time) can also provide sufficient scram reactivity. = Surveillance of each individual control rod's scram time ensures the scram reactivity assumed in the design basis transient and accident analyses can be met. The scram function of the CRD s MCPR (see Bases for LCO 3.2.2) and ystem the 1%protects claddingthe Safety Limit plastic strain fuel design limit (see Bases for LCO 3.2.1) which ensure no fuel damage if the limits are not exceeded. Above 950 psig the scram function is designed to insert negative reactivity at a raw fast enough to prevent the MCPR from becoming less than the Safety Limit MCPR during the limiting power transient analyzed in SSAR Chapter 15. Below 950 psig the scram function > 1s assumed to function during the Rod Withdrawal Error (RWE) event (Ref.5)andthereforealsoprovidesprotectionagainst violating fuel damage limits durin reactivity insertion accidents (see Bases for LC0 3.1.6 . For the SSAR Chapter 5 vessel overpressure protection ana ysis, the scram function along with the safety / relief valves ensures'the peak vessel pressure is maintained within the ASME Code limits. Control Rod Scram Times satisfies the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 6. (continued) ABWR B 3.1-14 5/31/89 O Amendment 9 16A.414 L _ ._._ _ . _ . - . _ , . . - . . _ _ _, _ . , . . . __ _ . _ , . . .. ._ - . .
1
- MM 23A6100AP Standard Plant aA ,
i f% Control Rod Scram Times I B 3.1.3 BASES feentinued) LCO The scram times specified in Table 3.1.31 are required to ensure the scram reactivity assumed in the design basis trans-l tent and accident analysis is met. To account for single i failures and ' slow' scramming control rods, the times specified in Table 3.1.31 are faster than those assumed in the design basis analysis. The scram times have margin to allow up to [ ] of the control rods to have scram times exceeding the limits (i.e., " slow" control rods) and also account for a single stuck control rod (as allowed by LCO 3.1.2) and an additional control rod pair (single failure criterion) failing 4 to scram. The scram times are specified as a function of ' reactor steam dome pressure to account for the pressure dependence of scram times. APPLICABILITY The CRD scram function is applicable during MODES 1 and 2 since a scram is only required when control rods are withdrawn, and . is assumed to function during transtants and accidents analyzed in these conditions. These events are assumed to occur during startup and power operation. In MODE 5. the scram capability of withdrawn control rods is specified by LCO 3.9.5. In MODES 3 and 4, control rods are only allowed to be withdrawn under
; Special Operations LCO 3.10.3 (Control Rod Withdrawal Hot 'N Shutdown) and LCO 3.10.4 (Control Rod Withdrawal . Cold Shut-down)-which provide adequate requirements for control rod scram capability during these conditions.
ACTIONS M For a control rod with excessive scram times (t [ ]secondsto 60% insertion from de energization of scram pilot valve sole-noids as time Zero) the control rod must be declared inoperable and therefore the control rod would be fully inserted and disarmed as required by LCO 3.1.2. Insertion of the control rod ensures the scram reactivity is not adversely affected by the failure of the control rod to scram, An example of a control rod with excessive scram times would be a control rod with a scram solenoid pilot valve that fails to open upon receipt of a scram signal. (continued) ABWR B 2.1-15 5/31/89 O Amendment 9 16A.4-15
23A6100AP Standard Plant u3 q j 9 Contrni Rod Scram Times B 3.3.3 BASES frontinued) ACTIONS U (continued) SR 3.1.3.2 is a periodic test to sample the control rod scram , times during the cycle to ensure the scram times have not degraded. Because only a sample of the control rods is tested (t 10%), a separate limit on the number of allowed " slow" control rods in the sample (20%) is specified. This limit is chosen such that if it is exceeded, it is an indication the total population of control rods, if tested, would exceed the allowed number of ' slow' control rods. Therefore, following the completion of SR 3.1.3.2 during which one or more control rods are discovered to be ' slow", the number of " slow" control rods in the sample must be deterstned and verified to be less than the sample limit. More than 10% of the control rods may be tested in SR 3.1.3.2 to provide a more representative sample. If a control rod is discovered ' slow" by means other than SR 3.1.3.2, this Jequired Action is not required. A.3. A 4 With a control rod whose scram time exceeds the scram time t-limits (a " slow" control rod) but is less than [ ] seconds. continued operation is justified if the number and distribution of ' slow" control rods is consistent with the assumptions of - the design basis transient and accident analyses. The scram times ofonly Table two3.1.31 are based on [ locations in any]
- slow' control rods, of which may occupy adjacent i direction and therefore no degradation of the design basis e
scram reactivity exists if these conditions are met. If scram-l time data already exist for the surrounding control rods, no ~ additional testing is required to determine if adjacent control rods are " slow". Control rods determined to be " slow" may alternatively be declared inoperable and the actions of LCO 3.1.2 followed. Inoperable control rods are not included in determining compliance with the requirements of Required Actions A.3 and A.4. M Multiple adjacent " slow" control rods or an excessive number of t ' slow' control rods can reduce the local scram reactivity f relative to that assumed in the design basis transient and accident analyses. Therefore, the reactor is required to be in MODE 3 within 12 hours. Also, if an unacceptable number of control rods are determined to be " slow" during SR 3.1.3.2, sufficient degradation of scram reactivity may be present and the reactor must be in MODE 3 within 12 hours. insertion of all control rods plates the reactor in a condition that does not require the scram function. l (continued) ABWR B 3.1-16 5/31/89 '- Amendment 9 16A.416 O
, n - - - - , , - - , - - . ---e- , - - , , - < , - --e a, m ,..n . , e--
Mhd- 21A6100AF Simadard Plant ,1 hA j d i D- i
= V .
Control Rod Scram Times < B 3.1.3 J BASES feontinued) l i 1 ACTIONS cannletion Times I (continued) i All Completion Times are based on industry accepted practice i and engineering judgement considering the number of available i systems and the time required to reasonably complete the J Required Action.
$URVIILLANCE SR 3.1.3.1 REQUIREMENTS The scram reactivity used in design basis transient and acci.
dent analyses is based on an assumed scram speed. Measurement of the scram times with reactor steam dome pressure greater than 950 psig dernonstrates acceptable scram times for the transients analyzed 'n References 4 and 5. Scram insertion times increase witt. increasing reactor pressure because of the competing effects of reactor steam dome pressure and stored accumulator energy. Therefore, demonstration of adequate scram times at reactor steam dome pressure greater than 950 psig ensures the scram times will be within the specified limits at higher pressure. Limits are specified as a function of reactor p- pressure to account for the sensitivity of scram insertion t
\ times with pressure and to allow a range of pressures during which scram time testing can be performed. To ensure scram time testing is performed in a reasonable tin,e following a refueling or after a shutdown greater than 120 days, all control rods are required to be tested before exceeding 40% of RATED THERMAL POWER following the shutdown.
SR 3.1.3.2 Additional testing of at least a 10% sample of control rods is required every 120 days of cumulative operation in MODE 1. For planned testing, the control rods selected for the 10% sample should be different for each test. Data from inadvertent scrams should be used if possible to avoid unnecessary testing at power even if the control rods with data may have been previously tested in a 10% sample. This frequency and number of tested control rods is based on engineering judgement considering the desire to minimize disturbances to normal plant operation, experience which shows scram' times do not significantly change over an operating cycle,' and the additional surveillances done on the control rod drives at more frequent intervals (LCO 3.1.2 and LCO 3.1.4). Testing of more than 10% of the control rods may be done to obtain a more representative sample. (continued) ABWR B 3.1 17 5/31/89
- A Amendment 9 16A.4-17
ABWR 3maour se==A=rd Plant aa . e-Control Rod Scram Times B 3.1.3 < BASES feontinued) SURVEILLANCE SR 3.1.3.2 (continued) REQUIREMENTS (continued) When possible, scram insertion time data can be determined frw full reactor core scrams. Otherwise, the scrar time data is obtained during individual rod pair scrams. As noted, for testing during rod pair scrams, the test shall be performed l' with the charging valve closed so that the influence of the CRD pump head does not affect the control rod pair under test (during a fall core scram, the CRD pump head would be seen by all control rods and would have a negligible ofTect on the scram insertion times). l SR 3.1.3.3. SR 3.1.3.4 for FSAR Chapter 15 analyses in the startup range (primarily l-the RWE), low pressure ( < 950 psig) scram times are assumed to be the same as the scram time limits at high reactor pressure (= 950 psig). This provides assurance that the low pressure scram times will be acceptable if the more stringent high l: pressure scram time limits are met during surveillance testing at t 950 psig. Should work on a control rod or the CRD system potentially affect the scram insertion time, testing must be done to demonstrate adequate scram performa. ice over the range of applicable reacter pressures. Specific examples of work that could affect the scram times are (but not limited to) the , following: removal of ary control rod drive for maintenance or l modification, replacement of a control rad. and maintenance or modification of a scram solenoid pilot whlve, scram valve, I accumulator or isolation / check valves in the piping required l 1 for scram. For work done while the reactor is at less than 950 psig, the scram testing must be performed before declaring the control roo uPERABLE and again at reactor steam come pressure greater - than or equal to 950 psig. This testing ensures prior to withdrawing the control rod for continued operation, the control rod scram function will be demonstrated both for startup and high reactor pressure conditions. Where work has been performed at high reactor pressure, both of these require-ments will be satisfied with one test. However, for a control rod affected while shutdown, a zero pressure and high pressure
- l. test :nay be required. Alternatively. 6 test during hydrostatic l pressure testing could also satisfy both criteria. To account i: for the variability in scram time; at different reactor pres-sures, specific scram time limits are specified in Table 3.1.3 1 at high reactor pressure conditions.
(continued) ABWR B 3.1-18 5/31/89 Amendment 9 16A.418 O
23A6100AP Remndard Plant hv. A , _ A -_ Control Rod Scram Tims B 3.1.3 j RASfs feontinued) SURVEILLANCE Surveillance Frenuancies REfulREMENTS (continued) In ger.eral, surveillance frequencies are based on industry - accepted practice and engineering judgement considering the I unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the i system / component status, i I J REFERENCES 1. 10 CFR Part 50, Appendix A, General Design Criterion 10.
- 2. NEDE 24Dll-P A, ' General Electric Standard Application for Reactor Fuel *, September 1988. '
- 3. ABWR SSAH, Section 4.6.1.
- 4. ABWR $$AR, Chapter 15.
l- 5. ABWR SSAR, Section 15.4.1
- 6. NEDO 31466, ' Technical Specification Screening Criteria
,/ -
Application and Risk Assessment", November 1987.
'\
l i-l- l l l-l l i, 1 ABWR B 3.1 19 5/31/89 [ l Amendment 9 16A.419
t JABWR zwioo^r : Standard Plant n4 I n ; 16A.4.4 R3.1.4 Contori Rod Scraai Accusmulators a f Control Rod Scram Accumulators B 3.1.4 8 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.4 control kod Scram Accimulators BASIS e BACKGROUND The control rod drive (CRD) scram accumulators are part of the CRD system and are provided to ensure the control rods scram I under varying reactor conditions. The control rod scram accumulators store sufficient energy to full control rods at any reactor vessel pressure.y The insert two accumulator is a hydraulic cylinder with a free floating piston. The piston separates the water used to scram the control rods from I ' the nitrogen which provides the required energy. The scram accumulators are necessary to scram the control rods within the required insertion times of LCO 3.1.3. l APPLICABLE The analytical methods and assumptions used in evaluating the ' SAFETY control rod scram function are presented in References 1, 2, 3 ANALYSES and 4. The design basis transient and accideat analyses assume
- all of the control rods scram at a specified insertion rate.
OPERABILITY of each individual control rod scram accumulator ensures (along with LCO 3.1.2, LCO 3.3.3 and LCO 3.1.5) the-scram reactivity assumed in the design basis transient and accident analyses can be met. The existence of an inoperable accumulator may invalidate prior scram time measurements for the associated control rods. The scram function of the CRD system, and therefore the-OPERABILITY of the accumulators, protects the Safety Limit MINIMUM CRITICA!. POWER RATIO (MCPR) (see Bases for LCO 3.2.2) and the 1% cladding plastic strain fuel design limit (see Bases for 100 3.2.1) which ensure no fuel damage if the limit is not exceeded (see Cases for LCO 3.1.3). Also, the scram function at low reactor vessel pressure (startup conditions) provides protection against violating fuel damage limits during reactivity insertion accidents (see Bases for LCO 3.1.6). Control Rod Scram Accumulators satisfies the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 5. (continued) ABWR B 3.1-20 5/31/89 O Amendment 9 16A.4-20
- ~ . - - . _ ._ , - . ,- - _ . , ._._ . - , . - . - ._ -_
i ABWR . 3minore j
,, Standard Plant u )
I fT I V !
)
Control Rod Scram Accumulators 8 3.1.4 BASES fcontinued) ' LCO The OPERABILITY of the control rod scram accumulators is required to ensure adequate scram insertion capability exists when needed over the entire range of reactor pressures. APPLICABIL11Y The control rod scram accumulators are required to be OPERABLE during MODES I and 2 whenever control rods are withdrawn and the scram function may be required. in MODE 5, the required OPERABILITY of accumulators associated with withdrawn control rods is specified in LC0 3.9.5. In MODES 3 and 4, control rods are only allowed to be withdrawn under Special Operations LCO 3.10.3 (Control Rod Withdrawal - Hot Shutdown) and LCO 3.10.4 (Control Rod Withdrawal Cold Shutdown) which provide adequate requirements for control rod scram accumulator OPERABILITY - during these conditions. ACTIONS L.L 1.2 The accumulator provides the primary scram force during opera. tion at any reactor steam dome pressure. However, because of the large number of control rods available for scram and the
.( Q assumed single failure of an HCU in the safety analysis, time is allowed to restore a single failed accumulator to OPERABLE status. If the accumulator cannot be restored to OPERABLE status within the required Completion Time, the associated control rods must be declared inoperable and the requirements of LCO 3.1.2 followed. This will require insertion of the contral rod thereby completing the intended function of the control rod.
IL1 c With two or more control rod scram accumulators inoperable, continued operation can only be justified for a short time. The importance of the accumulators in providing the scram force is such that the scram function could become severely degraded should more than one of the rcram accumulators, associated with fully or partially withdrawn control rods, be inoperable at the same time. The most likely cause for multiple control rod scram accumu-lators to be inoperable would be if there was not a CRD pump in operation supplying adequste pressure to keep the accumulators charged. However, the RPS will initiate a scram on ser. sed low (continued) ABWR 8 3.1-21 5/31/89 Amendment 9 16A.4 21
LAB M . - 2mione Standard Plant hA 'G Control Rod Scram Accumulators B 3.1.4 BASES fcontinued) ACTIONS L1 (continued) (continued) pressure in ths CRD charging water header so that all rods ate
-inserted prior to reaching a level of degradation beyond the safety analysis assumptions.
Of course there may be situations where two or more accumu-lators are inoperable for other reasons. In this case, a short time is allowed for corrective action to be taken (consistent with A.1 or A.2). However, if multiple accumulators associated with withdrawn control rods exist simultaneously for longer than the required completion time, then the Reactor Mode Switch must be placed in the Shutdown position. This ensures all insertable control rods are inserted and the reactor is in a condition that does not require the active function (scram and insertion) of the control rods. Comolation Times All Completion Times are based on industry accepted practice ) and engineering . judgement considering the number of available systems and the time required to reasonably complete the Required Action. SURVEILLANCE SR 3.1.4.1 l REQUIREMENTS The primary indicator of accumulator OPERABILITY is the accumu-lator pressure. A minimum accumulator pressure is specified, below which the capability of the accumulator to perform its intended function becomes degraded and the accumulator is considered inoperable. The minimum accumulator pressure of [ 1850 ) psig is well below the expected pressure of approxi-mately 2150 psig. Declaring the accumulator inoperable when the minimum pressure is not maintained ensures significant degradation in scram times does not occur. SR 3.1.4.1 requires the accumulator pressure be checked weekly to ensure adequate accumulator pressure exists to provide sufficient scram force. Operating experience has demonstrated that a seven day frequency for this surveillance is adequate. (continued) ABWR B 3.1-22 5/31/89 O Amendment 9 I M 22
MM 23s.i are Riandard Plant mi eD .t Control Rod Scram Accumulators B 3.1.4 RASf5 fenntinued) REFERENCES 1. NEDE 24011 P-A.' General Electric Stendard Application for Reactor Fuel'. September 1988.
- 2. ABWR $$AR. Section 4.6.1,
- 3. ABWR $$AR Chapter 15,
- 4. ABWR S$AR. Section 15.4.1.
- 5. NEDO 31466. ' Technical Specification Screening Criteria Application and Risk Assessment *. November 1987. i l
l l l O ABWR B 3.1 23 5/31/89 16A.4-23 Amendment 9
t MM 23A6100AP Res=A=ed Plant aa - 16A.4J B3.1.5 Control Rod Drive Coupling i CR0 Coupling - B 3.1.5 8 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.5 Control tod Drive toucline BASES . BACKGROUND The control rod drive (CRD) and control rod are coupled through a bayonet type coupling at the top end of the CRD where the drive tube locks into a mating socket at the base of the ' control rod. The coupling requires a 45' rotation for either engaging or disingaging. This positive means of coupling the control rod to the drive tube allows the position of the control rod to be varied and detected through the movement of the traveling nut and drive tube. CRD coupling is a requirement for control rod OPERABILITY since the lack of , coupling could result in loss of control rod position information and potentially result in a large reactivity insertion should the control rod become stuck and later drop from the core. However, this could only occur in conjunction with other failures such as failure of the redundant Class IE separation detection devices (see LCD (Rod Block Instrumentation)). APPLICABLE The analytical methods and assumptions used in evaluating the , SAFETY CRD system are presented in References 1, 2 and 3. The CRD I ANALYSES system provides the negative scram reactivity required to ensure specified acceptable fuel design limits are not exceeded durin abnormal operational transients (see Bases for LCO 3.1.3 . Also, if a control rod is not coupled, the position of the control rod cannot necessarily be determined. The control rod l position (specifically the control rod pattern) is an initial l condition of the Rod Withdrawal Error (RWE) event analysis. Demonstrating coupling of all control rods therefore provides protection against violating fuel damage limits during reactivity initiated accidents (see Cases for LC0 3.1.6). Additionally, the positive means of coupling, along with the l use of redundant separation detection devices, precludes the possibility of occurrence of a Control Rod Drop Accident. Control Rod Drive Coupling satisfies the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 4. I (continued) ABWR B 3.1-24 5/31/89 Amendment 9 16AM O
~
23A6100AP Standard Plant u.a
. 't )
CRD Coupling 8 3.1.5 BASES feontinued) LCO . The requirement for all cortrol rods to be coupled to their drive mechanisms is required to ensure OPERABILITY of the control rods such that their intended reactivity control functions can be performed. APPLICABILITY The CRD coupling requirements are applicable during MODES I and 2 since under these conditions control rods or control rod pair can be withdrawn and criticality achieved. The ORDA is not applicable during MODES 3, 4 and 5 since only a single control rod or control rod pair can be withdrawn from a core cell containing fuel assemblies and SHUTDOWN MARGIN ensures the reactor is suberttical under thess conditions. ACTIONS A.I. A.2 Continued operation with an uncoupled control rod should not be allowed because of the increased probability of a CRDA and therefore only a short period of time is allowed to establish
.[ ,% . and verify recoupling. Since the allowable time with an Q uncoupled control rod is short, and control rods do not always recouple on the first try, multiple attempts to recouple a control rod may be necessary.
If recoupling cannot be achieved, the control rod must be declared inoperable and the requirements of LCO 3.1.2 must be satisfied. This will ensure if the :ontrol rod cannot be recoupled in a short time, it will be inserted and disarmed thereby maintaining the shutdown capabilities of the control l rod. A CRDA involving this uncoupled control rod would still ' be precluded by the redundant separation detection devices. , However, to reduce the probability that the rod may become stuck, it should be inserted and disarmed. l l- Additionally, to prevent potential damage to the rod / drive i t coupling, after insertion, the control rod should be isolated . I from scram. The control rod can be isolated from scram by I I isolating it from its associated hydraulic control unit and the i l CRD charging water header, j 1 (continued) ABWR B 3.1 25 5/31/89 g l l- Amendment g 16AA 25 l l
)
M 2m6100AP Standard Plant a-2 CRD Coupling-B 3.1.6 IL'.SES icontinued) ACTIONS . Comolation Times (continued) All Completion Times are based on industry accepted practice and engineering judgement considering the number of available (~ systems and the time required to reasonably complete the Required Action. SURVEILLANCE SR 3.1.E.1 REQUIREMENTS Coupling verification is performed by verifying a control rod does not go to the overtravel position when it is fully with-lJ drawn. The overtravel position feature provides a positive ! check on the coupling integrity since only an uncoupled CRD can reach the overtravel position. (Since the traveling nut will go to the overtravel position, it will be temporarily separated from the rod and drive as the control rod is backseated at the l full out position. Therefore, this test also allows verifica-tion of the proper functioning of the redundant separation detection devices. See LCO (Control Rod Block Instrumentation).) This verification is required to be i performed the first time a control rod is withdrawn to the
' Full Out" position following refueling and prior to declaring the control rod OPERABLE when work on the control rod or CRD >
system could affect coupling. l REFERENCES 1. NEDE 240ll P A 9-US, ' General Electric Standard Application for Reactor Fuel *, Supplement for United States, September 1988,
- 2. ABWR SSAR Section 4.6.1.
l l 3. ABWR SSAR, Section 15.4.1. I l
- 4. NEDO 31466, " Technical Specification Screening Criteria
!. Application and Risk Assessment", November 1987. i I-i 9 ABWR B 3.1 26 5/31/89 O Amendment 9 16A.4-26
LABWR zwiom
- Standard Plant MA 16A.44 B3.14 Rod Pattern Control
' C/ -
Rod Pattern C6ntrol B 3.1.6 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.6 tod Pattern Control BASES 1 BACKGROUND Control rod patterns during startup conditions are controlled ] by the operator and the Rod Worth Minimizer (RWM) (LCO (Control Rod Block Instrumentation)), such that only specified control I rod sequences and relative positions are allowed over the l operating range from all control rods inserted to in of RATED 1 THERMAL POWER. The sequences effectively limit the potential ' amount and rate of reactivity increase that could occur during a control rod withdrawal, specifically the Rod Withdrawal Error (RWE) event, l APPLICABLE The analytical methods and assumptions used in evaluating the SAFETY RWE are summarized in References 1 and 2. RWE analyses assume ANALYSES the reactor operator follows prescribed withdrawal sequences. These sequences define the potential initial conditions for the RWE analysis. .The RWM (LCO [ Control Rod Block Instrumenta-tion)) provides backup to operator control of the withdrawal f sequences to ensure the initial conditions of the RWE analysis
'V are not violated.
Control rod patterns analyzed in References 1 and 2 follow the GWSR which is the same as the BPWS described in Reference 6. The GWSR is applicable from the condition of all control rods fully inserted to 10% of RATED THERMAL POWER. For GWSR, the control rods are required to be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions. The banked positions are defined to minimize the maximum incremental control rod worths without L being overly restrictive during normal plant operation. Prevention or mitigation of positive reactivity insertion events is necessary to limit energy deposition in the fuel to l prevent significant fuel damage which could result in undue release of radioactivity (Ref. 5). Since the failure conse-i quences for UD, have been shown to be insignificant below fuel l- energy depositions of 300 cal /gm, the fuel damage limit of 280 l cal /gm provides a margin of safety to significant core damage and release of radioactivity (Ref. 3 and 4). Generic analysis of the GWSR (BPWS, Ref. 6) has demonstrated the 280 cal /gm fuel damage limit will not be violated during a postulated reactiv-ity transient while follcwing the GWSR mode of operation. The 1- generic analysis also evaluated the effect of fully inserted l- inoperable control rods not in compliance with the sequence to ! allow a limited number (8) and distribution of fully inserted l: inoperable control rods. l (continued) ABWR B 3.1 27 5/31/89 Amendment 9 16A.4-27 l i i l
i
-MM 23A6100AP
. Rimadard Plant - n- A O
4 f Rod Pattern Control j B 3.1.6 BASES fcontinued) APPLICABLE Rod Pattern Control satisfies the requirements of Selection SAFETY Criterion 3 of the NRC Interim Policy Statement on Technical, ANALYSES Specification Improvements as documented in Reference 7. (continued)
- LCO Compliance with the prescribed control rod sequences minimizes the potential consequences of a RWE by limiting the initial conditions to those consistent with GWSR. This LC0 only applies to OPERABLE control rods. For inoperable control rods required to be inserted, separate requirements are specified in LCO 3.1.2, consistent with the allowances for inoperable control rods in GWSR.
APPLICABILITY Compliance with GWSR is required in MODES 1 and 2 when THERMAL POWER is less than or equal to 10% of RATED THERMAL POWER. When THERMAL POWER is greater than 10% of RATED THERMAL POWER, there is no possible control rod configuration that results in a control rod worth that could exceed the 280 cal /gm fuel damage limit during a RWE. In MODES 3, 4 and 5, since only a total of one control rod or control rod pair can be - withdrawn from core cells containing fuel assemblies, adequate SHUTDOWN MARGIN ensures the reactor will remain subtritical. (continued) J' t I i 4 1 4 i ABWR B 3.1-28 5/31/89 I O j Amendment 9 16AM8 4 4
w MM . 23A6100AP Standard Plant En A O M; Rod Pattern Control B 3.1.6 BASES feontinued) ACTIONS A.1. A.2 With a limited number of OPERABLE control rods not in compit- ! ance with the prescribed control rod sequence, actions may be taken to correct the control rod pattern. Noncompitance with the prescribed sequence may be t h result of failed synchos, drifting from a CRD purge water transient, leaking scram valves or a power reduction below 10% of RATED THERMAL POWER before establishing the correct control rod pattern. The number of OPERABLE control rods not in compitance with the prescribed l sequence is limited to 8 to prevent the operator from i attempting to correct a control rod pattern that significently deviates from the prescribed sequence. When the control rod pattern is not in compliance with the prescribed sequence, all control rod movement should be stopped except for scram or those moves needed to correct the rod pattern. As noted, control rods may be bypassed in RC&l5 to allow the affected control rods to be returned to their correct position. This 4 ensures the control rods will be moved to the correct position. !! Alternatively, if the affected control rod is not moved to its correct position in the Required Completion Time, it may be l declared inoperable and the requirements of LCO 3.1.2 followed. A control rod not in compliance with the prescribed sequence is ! h
'd not considered inoperable except as required by Required Action A.2. OPERABILITY of control rods is determined by compliance with LCO 3.1.2 through LCO 3.1.5.
B.I. B.2 More than 8 out of sequence OPERABLE control r;,ds indicates the 4 control rod pattern significantly deviates from the prescribed seauence. Control rod withdrawal should be suspended to prevent the potential for further deviation from the prescribed sequence. Limited (I hour) control rod insertion to correct control rods withdrawn beyond their allowed position is allowed 1 since, in general, insertion of control rods has less impact on l control rod worths than withdrawals beyond the prescribed i h.it s . if the control rod pattern requirements cannot be restored within I hour, the Reactor Mode Switch must be placed in the Shutdown position. l- Comoletion Times L l All Completion Times are based on industry accepted practice l and engineering judgement considering the number of available systems and the time required to reasonably complete the ( Required Action. l (continued) ABWR B 3.1 29 5/31/89 Amendmut 9 16AA-29 I l
i
~
23A6100AP Standard Plant - a- i i
.q
(- O l l Rod Pattern Control B 3.1.6 l- RASES fcontinued)
- SURVEILLANCE 1R 3.1.6.1 REQUIREMENTS The primary check on compliance with GWSR is performed by the RWM (LC0 [ Control Rod Block Instrumentation)) which provides i
control rod blocks to enforce the required sequence. The RWei is required to be OPERABLE when operating at less than or equal to 10% of RATED THERMAL POWER. Should a control rod be j bypassed in RC&lS the RWM will not block movement of this 3 control rod and therefore LCO [ Control Rod Block Instrumenta-tion) requires the bypassing and movement of the bypassed control rod to be verified under these conditions to ensure the bypassed control rod is returned to its correct position. Therefore, a daily check of the control rod pattern compliance with GWSR is adequate. l-L REFERENCES 1. ABWR SSAR Section 15.4.1.
- 2. NEDE 240!! P A 9 US, " General Electric Standard Application for Reactor Fuel - Supplement for United States', Septebmer 1988.
t- l l 3. NUREG 0800, Revision 2, Standard Review Plan, July 1981, i Section 15.4.1. 1-
- 4. Regulatory Guide 1.77. *
[ 5. 10 CFR Part 100,
- 6. NEDO 21231, "Bahked Position Withdrawal Sequence", January 1977.
, 7. NEDO-31466, " Technical Specification Screening Criteria l Application and Risk Assessment', November 1987. i 1 l i l L I-i l. l f l ABWR B 3.1 30 5/31/89 O l Amendment 9 16A.4-30 l l
~ MM 21A6100AP Standard Plant mi 16A.4.7 83.1.7 Standby Uquid Control Systema A
SLCS B 3.1.7 B 3.1 REACTIVITY CONTROL SYSTEMS ' B 3.1.7. Standhv Lieutd control system BASES BACKGROUND The Standby Liquid Control System (SLCS) is designed to provide the capability of bringing the reactor, at any time in a cycle, from full power and minimum control rod inventory (which is defined to be at the peak of the xenon transient) to a subcritical condition with the reactor in the most reactive i menon free state without control rod movement. The system is designed for the conostions when no control rods can be inserted from full power conditions. The SLCS consists of a boron solution storage tank, two positive displacement pumps, two motcr operated injection valves which are provided in parallel for redundancy and associated piping and valves used to transfer borated water from the storage tank to the reactor pressure vessel (RPV). The borated solution is discharged through the 'B' High Pressure Core Flooder (HPCF) subsystem sparger. D)
-( APPLICABLE SAFETY The SLCS is manually initiated from the main control room as directed by the Emergency Operating Procedures if the operator ANALYSES believes the reactor cannot be shutdown or kept shutdown with the control rods. The SLCS is used in the highly improbable event that not enough control rods can be inserted to '
accomplish shutdown and cooldown in the normal manner. The SLCS provides borated water to the reactor core to compensate for the various reactivity effects during the required condi-tions. 10 meet this objective it is necessary to inject a quantity of boron which produces a concentration of 850 ppm of natural boron in the reactor core at 68'F. To allow for potential leakage and imperfect mixing in the reactor system, an additional 25% (220 ppm) is added to the above requirement (Ref. 1). The volume and concentration limits in Figure 3.1.7 1 are calculated such that the required concentration is achieved accounting for dilution in the RPV with normal water level and including the volume in the Residual Heat Removal shutdown cooling piping (additional 250 ppm). This quantity of solution (with a total required concentration of 1320 ppm) is the amount which is above the pump suction shutoff level in the tank thus allowing for the portion of the tank volume which cannot be injected. (continued) ABWR S 3.1 31 L 5/31/89 ; Amendment 9 16A.4-31 l
. 1 1- l
t MM 23A6100AP , Standard Plant a-a 1 SLCS B 3.1.7 BASES frontinued) APPLICABLE The NRC Interim Policy Statement (Ref. 2) requires the SLCS be SAFETY retained in the technical specifications even though none of ANALYSES the Selection Criteria were satisfied (Ref. 3). * (continued) LCO The OPERABILITY of the SLCS is required to provide backup capability for reactivity control independent of normal reactivity control provisions from the control rods. The l '. OPERABILITY of the SLCS is based on the conditions of the borated solution in the storage tank and the availability of a flow path to the reactor pressure vessel, including the OPERABILITY of the pumps and valves. Two SLCS subsystems are required OPERABLE, each containing an OPERABLE pump, an explosive valve and associated piping and valves to ensure an OPERABLE flew path, b APPLICABILITY The SLCS specifications are applicable during MODES 1 and 2 since during these conditions the reactor can be critical. In MODES 3 and 4, control rods are only allowed to be withdrawn under Special Operations LCO 3.10.3 (Control Rod Withdrawal - ' Hot Shutdown) and LCO 3.10.4 (Control Rod Withdrawal Cold Shutdown) which provide adequste controls to ensure only a single control rod or control rod pair can be withdrawn, in MODE 5, only a single control rod or control rod pair can be withdrawn from core cells containing fuel assemblies, and demonstration of adequate SHUTDOWN MARGIN (LC0 3.1.1) ensures the reactor will not be critical. Therefore the SLCS is not required to be OPERABLE during these conditions when only a single control rod or control rod pair can be withdrawn. ACTIONS Ad With one SLCS subsystem inoperable, the remaining OPERABLE subsystem is adequate to perform the shutdown function. Howeser, the overall reliability is reduced because a single failure in the remaining OPERABLE subsystem can result in no SLCS shutdown capability. For this reason, continued operation is permitted for a limited time only, 7 days. (continued) ABWR B 3.1 32 5/31/89 O Amendment 9 16A.4-32
ABWR nanoose , Standard Plant n. u J l
)
l SLCS B 3.1.7 BASES fcontinued) ACTIONS . B.1. B.2 (continued) titth both SLCS subsystems inoperable, no SLCS shutdown capability remains to handle the postulated event. Continued operation is justified because of the low probability of an ' event which would require SLCS. However, continued operation of the plant is permitted for only a limited period of time. 8 l hours, during which time at least one subsystem must be restored to OPERABLE status. Additionally, the initial inoperable subsystem must be restored to OPERABLE status within 7 days from initial discovery consistent with the Completion Time of Required Action A.I. L1 Pith one or both SLCS subsystems not restored to OPERABLE status and the associated Completion Times not met, the reactor must be in MODE 3 within 12 hours. Comoletion Tires All Completion Times are based on industry accepted practice
,.- and engineering judgement considering the number of available d( systems and the time required to reasonably complete the Required Action. 5 i
~ SURVEILLANCE SR 3.1.7.1. SR 3.1.7.2. SR 3.1.7.3 REQUIREMENTS
!- SR 3.1.7.1 through SR 3.1.7.3 are daily surveillances verifying portions of the SLCS OPERABILITY without disturbing normal I plant operation. The surveillances ensure the proper solution vclume and temperature (including the temperature of the pump i I' suction piping) are maintained. The solution temperature is important in ensuring the boron remains in solution and does not precipitate out in the storage tank or pump suction piping. Failure to meet SR 3.1.7.1, SR 3.1.7.2 or SR 3.1.7.3 will make both SLCS subsystems inoperable since the storage tank and the majority of the pump suction piping is common to both l subsystems. (continued) l-ABWR B 3.1 33 5/31/89 A Amendment 9 IM33
t ABWR z w ioorr I se dard Plant *p 9 SLCS B 3.1.7
&&$D fcontinued)
SURVEILLCCE SR 3.1.7.4. SR 3.1.7.5 REQUIREMENTS (continued) SR 3.1.7.4 and SR 3.1.7.5 provide additional demonstration of the OPERABILITY of the SLCS, SR 3.1.7.4 provides a detailed , examination of the sodium pentaborate solution by using chem-ital analysis to ensure the proper concentration of boron solution exists in the tank. SR 3.1.7.5 verifies each valve in , the system (not otherwise locked, sealed or secured in post-tion) is in its correct position to ensure the flow path is available to support operation of the subsystem. SR 3.1.7.4 must be performed anytime boron or water is added to the tank solution to establish acceptability of the new values for the solution concentration. Also, SR 3.1.7.4 must be performed anytime the temperature is restored to within the limits of Figure 3.A.71 to ensure no significant boron ' precipitation occurred. SR 3.1.7.6 SR 3.1.7.6 demonstrates proper operation of the FLCS pumps and requires each pump be demonstrated to meet the minimum flow rate requirement of 50.0 gpm at a discharge pressure of 1223 psig. The minimum pump flow rate requirement ensures that when combined with the sodium pentaborate solution concentra-tion requirements, the rate of negative reactivity insertion from the SLCS will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of the moderator and xenon decay. SR 3.1.7.7. SR 3.1.7.7 SR 3.1.7.7 and SR 3.1.7.8 ensure a complete flow path from the storage tank to the reactor pressure vessel. The pump and injection valve tested should be alternated such that both complete flow paths are tested every 35 months. Surveillance Freauencies in general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and,a likelihood of a change in the system / component status. (continued) ABWR B 3.1 34 5/31/89 16A.4 34 O Amendment 9
l e i i 23A6100AP -. _ _R_eandard Plant mi t t
$LCS I B 3.1.7 l'
RASES fcontinued) REFERENCES 1. ABWR SSAR, Section 9.3.5.3. ;
- 2. 52FR3788, ' Proposed Policy Statement on Technical Specification leprovements for Nuclear Power Reactors "
February 6, 1987.
- 3. NED0 31466, " Technical Specification Screening Criteria Application and Risk Assessment", November 1987.
O V l l I l i l ABWR B 3.1 35 5/31/89 O Amendmen 9 16A.4-35
L' l
~ABWR ^"^' '
Standard Pljint Rev A SECTION 16A.5 CONTENTS -
-- m . :
16M,1 R311 Avernpr Planar unmar Heat Generntlem Eafe 16 M .1 16 M .2 R3.2.2 Mlaimum Cdtical Power Ratio 16M.5 16MJ R3.2.3 unmar Heat Genernilon Rate 16M 9 l l l l O 16A.5-ii Amendment 9
. 23A61M i Standard Phnt a.a !
- 16A.5 B3.2 Power Distribution Limits 1 p-- '
16A.S.1 R3.2.1 Average Planner Linear Heat Generator Rate APLHGR B 3.2.1 B 3.2 POWER DISTRIBUTION LIM 115 8 3.2.1 AVERAAE PLANAR LINEAR HEAT AENERATION RATE RAsts . BACKGROUND The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is a measure of the average linear heat generation rate of all the fuel rods in a fuel assembly at any axial location. Limits on APLHGR are specified to assure that the fuel design limits i identified in Reference I will not be exceeded during anticipated operational occurrences and that the peak cladding temperature (PCT) during the postulated design basis loss of coolant accident (LOCA) will not exceed the limits specified in 10 CFR 50.46. APPLICABLE- The ar.alytical methods and assumptions used in evaluating the SAFETY fuel design limits are presented in SSAR Chapter 4 and in ANALYSES Reference 1. The analytical methods and assumptions used in evaluating Design Basis Accidents, anticipated operational transients and normal operation that determine the APLHGR limits are presented in the FSAR, Chapters 4, 6 and 15 and in
,O Reference 1.
Fuel design evaluations are performed to demonstrate that the cladding 1% plastic strain and other fuel design limits described in Reference 1 are not exceeded during anticipated > operational occurrences for operation with LINEAR HEAT GENERATION RATES (LHGR's) up to the operatino limit LHGP. APLHGR limits are equivalent to the LHGR limit for each fuel rod divided by the local peaking factor of the fuel assembly. APLHGR limits are developed as a function of exposure and the various operatirig core flow and power states to ensure adher-ence to fuel design limits during the limiting anticipated operational occurrences. Flow dependent APLHGR limits are determined using the three dimensional BWR simulator code ' (Ref. 2).to analyze slow flow runout transients. The flow-dependent multiplier, MAPFACf . is dependent on the maximum core flow runout capability. The maximum runout flow is dependent on the existing setting of the core flow limiter in the Recirculation Flow Control System. Based on analyses of limiting plant transients (other than core flow increases) over a range of power and flow conditions, power-dependent multipliers (MAPFAC p
) are also generated. Due to the sensitivity of the transient response to initial core flow levels at power levels below that where turbine stop valve (continued)
ABWR C 3.2 1 5/31/89 b V Amendment 9 gA3,3
, . , , . , , . ~ . , . - . , ~ . . - , , .-
MM 23A6100AP l Staad=rd Plant m. A O. APLHGR B 3.2.1 RASES fcontinued) APPLICABLE closure and turbine control valve fast closure scram trips are SAFETY bypassed both high and low core flow MAPFAC limits are ANALYSES P (continued) provided for operation at power levels between 25% of RATED i THIRMAL POWER and the previously mentioned bypass power level. The exposure dependent APLHGR limits are reduced by NAPfAC and. p , [ etAPFACy at various operating conditions to ensure that all fuel design criteria are met for normal operation and anticipated operational occurrences. A complete discussion of the analysis code is provided in Reference 4. LOCA analyses are then performed to ensure the above determined APLHGR limits are adequate to meet the PCT and maximum oxida-tion limits of 10 CFR 50.46. The analysis is performed using GE calculational models which are consistent with the require. ments of 10 CFR 50. Appendix K. A complete discussion of the analysis code used in the analysts is provided in Reference 1. The PCT following a postulated LOCA is primarily a function of i the average heat generation rate of all the rods of a fuel i= assembly at any axial location and is not strongly influenced i by the rod to rod power distribution within an assembly. The l APLHGR limits specified are equivalent to the LHGR of the I highest powered fuel rod assumed in the LOCA analysis divided by its local peaking factor. A conservative multipiter is applied to the LHGR assumed in the LOCA analysis to account for the uncertainty associated with the measurement of APLHGR. APLHGR satisfies the requirements of Selection Criterion 2 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 3. LCO The APLHGR limits specified in the [ CORE OPERATING LIMITS REPORT) are the result of the fuel design analysis and design basis accident and transtent analysts. The limit is determined by multiplying the smaller of the MAPFAC, and MAPFAC factors times the exposure dependent APLHGR limits. P (continued) ABWR B 3.2 2 5/31/89 O Amendment 9 16A.5-2
. . ,. . ..-.-,- -..- - . - .. ~.. , . - . . . . - - . . . . . - . - - . . .
. ... . . ~
23A6100AP Standard Plant w r l
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APLHGR j B 3.2.1 BASIS icontinued) APPLICABILITY The APLHGR limits are primarily derived from fuel design evaluations, LOCA and transient analysis that are assumed to ' occur from high power level conditions. Design calculations and operettog experience have shown that as power is reduced. margin to required APLHGR limits increases. This trend cc - tinues down to the power range of 6 151 of itATED THERMAL POWER (RTP) where entry into MODE 2 occurs. When in MODE 2, the Startup Range Neutron Monitor (SRNM) scram function will provide prompt scram initiation during any significant transi-ent thereby effectively removing any APLHGR limit compliance concern in MODE 2. Therefore, at THERMAL POWER levels less than or equal to 25% of RTP. the reactor will be operating with substantial margin to APLHGR limits and the specification is not required. l l ACTIONS M !; Should any APLHGR exceed the required limits, an initial condition of the design basis accident and transtant analyses , may not be met. Therefore, orompt action should be taken to ,
.' restore the APLHGR's to within the reGuired limits such that the plant will be operating within analyzed conditions and b.
! within design limits of the fuel rods. M l If the APLHGR cannot be restored to within the required limits lf- in two hourss it is required to reduce THERMAL POWER to < 25% l of RTP. As discussed in the Bases for Applicability, operation below 25% of RTP results in sufficient margin to the required limits. Comoletion Times The Completion Times are based on industry accepted practice and engineering judgement considering the time to reasonably complete the Required Action. (continued) ABWR B 3,2 3 5/31/89 Amendment 9 16AS-? l
23A6100AP S1AD d ard Plant ,, i O APLHGR B 3.2.1 ] RASES fenntinued) SURVEILLANCE 1R 3.2.1.1 REQUIREMENTS ' APLHGR's are reeutred to be initially calediated within 12 hours after THERMAL POWER has exceeded 25% of RTP and then daily thereafter. They are compared to the specified limits to assure that the reactor is operating within tne assumptions of the safety analysis. The daily survelliance requirement is based on engineering judgement considering the slow changes in power distribution. The 12 hour allowance after exceeding 25% of RTP is acceptable given the large inherent margin to operating limits at low power levels. REFERENCES 1. NED0 240ll P A, ' General Electric Standard Application for Reactor Fuel *, September 1988.
- 2. NED0 301300 A, ' Steady State Nuclear Methods'. May 1985.
- 3. NEDO 31466, ' Technical Specification Screening Criteria Application and Risk Assessment,' November 1987.
- 4. NEDO 24154. " Qualification of the One Dimensional Core Transient Model for Bolling Water Reactors *, October 1978.
4 ABWR B 3.2-4 5/31/89 O Amerd nent 9 16A34
i 2nA61AAP Standard Plant m_ i l
.p 164.$J B3.2.2 Minimman Critical Power Ratio l
l MCPR B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS 8 3.2.2 R1ginal f11TitAL poltti RATIO ;
&asts - !
+
BACKGRDUND The MINIMWM CRITICAL POW [R RATIO (MCPR) is a seasure of the operating fuel assembly power relative to the fuel assembly , power that would result in the onset of boiling trsnsition. The $4fety Lleit MCPR is set such that 99.9% of the fuel rods will avoid boiling transition if the liett is not violated. > (refer to the Bases for LC0 2.1.2). The operating limit MCPf1 , is estabitshed to assure that no fuel damage resultt' during anticipated operational occurrences. Although fuel damage would not necessarily occur if a fuel 'd actually esperienced
, the crittu t power at which ;
boiling boiling transition transition(Ref.1) is calcu lated to occur has been adopted as a fuel design criterion. The onset of transition boiling is a phenomena that is readily detecte$ @: ring the testing of various bundle designs. Based on th% tJ.;vimental data, correlations have been develeped that # 6 tm d to credttt critical bundle power (i.e., the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., pressure, mass flue, l l O'. subcooling,etc.). Since plant operating conditions and bundle power levels are relatively easily monitored and determined. i t l monitoring MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur. i l APPLICABLE The analytical methods and assumptions used in evaluating ! $AFETY the anticipated operational occurrences to establish the , ANALY5t$ operating limit MCPR are presented in the $5AR, Chapters 4, 6 and 15. and in keference 2. To assure that the Safety Limit MCPR is not exceeded during any moderate frequency transient event, limiting transients have been analyzed to detersine the ; largest reductlon in Critical Power Ratio (CPR). The type of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion and coolant temperature decrease. The limiting transient yttids the largest ACPR. When the largest ACPR is a(ded to the safety Limit MCPR, the required operating limit MCPR is obtained. The MCPR operating limits derived from the transient an 1ysis are dependent on the operating core flow and power state (MCPRg i and MCPR p respectively) to ensure adherence to fuel design (continued) ABWR B 3.2 5 5/31/29 O i O Amendment 9 34 T
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w ,,.,.-m_ _ . , , ,. , , , _ . _ . . _ _ .. . _ . _ _ , , - . . _.._ ...-.,.i
1 MN 2M6100AP am i Standard Plant j i l 9\ l
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MCPli , 8 3.2.2 i KAME feentinued) j APPLICABLE limits during the worst moderate frequency transient. Flow 5AF ETY dependent MCPR limits are deterstned by steady state thermal ANALYSIS hydraulic methods with key physics response inputs benchmarked ; (continued) using the three dimenstenal DWR simulator code (Ref. 3) to > analyze slow flow runout transients. The operating limit is dependent on the maximum core flow limiter setting in the f Recirculation Flow Control System. Power dependent MCPR limits (MCPR p
) are determined mainly by j the one dimensional transient code (Ref. 4) for the anticipated transients that are significantly affected by power. Due to .
the sensitivity of the transient response to initial core flow levels at power levels below that where the turbine stop valve closure and turbine control valve fast closure scram trips are bypassed, a high and low flow operating limit MCPR p is . provided for operating between 25% of RATED THIRMAL POWIR (RTP) and the previously mentioned bypass power level. MCPR satisfies the requirements of Selection Criterion 2 of the NRC Interim Policy 5tatement on Technical Specification leprovements as documented in Reference 6. I LC0 TheMCPRoperatinglimitsspecifiedinthe[COR[ OPERATING LIMIT 5 REPORT) are the result of'the design basis accident and transient analysis. The operating limit MCPR is determined by the larger of the MCPRg and MCPR p limits. APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur from high power level conditions. Below 25% of RTP. the reactor will be operating at , minimum reactor internal pump speed and the moderator void content will be very small. Surveillance of thermal limits . below 25% of RTP is unnecessary due to the large inherent margin that assures that the Safety Limit MCPR will not be exceeded even if a 11miting transient should occur. Statis-tical analyses documented in Reference 5 indicate that the nominal value of initial MCPR expected at 25% of RTP is in excess of 3.5. Studies of the variation of limiting transient behavior have been performed over the range of power / flow conditions. These studies encrepass the range of key actual plant parameter values important to typically limitir.; transients. (continued)
. ABWR B 3.2 6 $/31/89 O
Amendment 9 16AS 6 s--+,-- - -w,- - , -- --,
- p. w - ,
M 2146100AP I R*==dard Plant m. a L MCPR i 8 3.2.2 ! v RASEK frontinued) APPLICABILITY Tha results of these studies demonstrate that margin is , (continued) expected between performance and MCPR requirements, and that margins increase as power is reduced to 25% of RTP. This trend ' is expected to co'ittnue to the 515% power range where entry into MODE 2 occurs. When in MODE 2, the Startup Range Neutron Monitor ($RNM) provides rapid scram inttistion for any signif. ' icant power increase transient which effectively eliminates any MCPR compliance concern. Therefore, at THERMAL POWER levels less than 25% of RTP, the reactor will be operating with substantial margin to MCPR limits and the specification is not required. ACTIONS M ' Should any MCPR be outside the required limits, an initial condition of the design basis transient analyses may not be met. Therefore, prompt action should be taken to restore the MCPR's to within the required Italts such that the plant will be operating within analyzed conditions. ' O If the MCPR cannot be restored to within the recuired lielts in two hours, it is required to reduce 1HEfqMAL POWER to < 25% of i' RTP. As discussed in the Bases for Applicability, operation below 25% of RTP results in sufficient margin to the required limits. Comohtien Times The Completion Times are based on industry accepted practice and engineering judgement considering the time to reasonably l complete the Required Action. t SURVElLLANCE $R 3.2.2.1 i REQUIREMENTS MCPR is required to be initt.11y calculated within 12 hours after THERMAL POWER has exceeded 25% of RTP and then daily thereafter. It is compared to the specified limits to assure l that the reactor is operating within the assumptions of the l safety analysis. The daily surveillance requirement is based on engineering judgement considering the slow changes in power i distributton. The 12 hour allowance after exceeding 25% of RTP '. is acceptable given the lar ! limits at low power levels.ge inherent mergin to operating (continued) ABWR B 3.2 7 5/31/89 O j Amendment 9 16AS 7 1
._ _- - . . - . - -- - - - -~-- - _ -_ -. .
23A6100AP f Stamulard Plant u.i I O i MCPR B 3.2.2 RAlti fenntinued) REFERENCE $ 1. NURIG 0562
- Fuel Rod failure as a Conseguence of i Departure From Nucleate Bolling or Dryout'. June 1979.
- 2. NED0 24011 P A ' General [lectric Standard Appittation for Reactor Fuel'. September 1988.
- 3. NEDD 30131 A. *$teady $ tate Nuclear Methods'. May 198$. .i 4 NEDO 24164 ' Qualification of the One Dimensional Core Transient Model for toiling Water Reactors.* Detober 1978. !
, 6. 'BWR/6 Generic Rod Withdrawal Error Analysis'. Appendix 368, General Electric Standard Safety Analysis Report (GESSAR ll).
- 6. NEDD 31466 ' Technical $pecification $creening Criteria Application and Risk Assessment *, November 1987. ;
Ot E ABWR B 3.2 8 5/31/89 O Amendment 9 16A.5-8
1 MM zu61ooAP i Stand =ed Plant ui i 16A.$.3 R3.2.31 Ameer Heat Generstlom Rate I LNGR 8 3.2.3 8 3.2 POWIR DI$TRIBUT10N LIM 11$ 8 3.2.3 LINEAR REAT RENERAfl0N RATE RA$f5 BACKGROUND is a measure of the heat The LIN[AR generation rateN[A1 of a GEN fuel rod (RAfl0N in a fueRAff. (LNGR)l assembly at an axial location. Limits on LNGR are specified to assure that fuel , design limits will not be exceeded anywhere in the core during normal operation including 6.iticipated operational occurrences. Exceeding the LHGR limit could potentially result in fuel damage and subsequent release of radioactive materials. Fuel design limits are specified to assgre that fuel system damage, fuel rod failure or tr. ability to cool the fuel will not occur during the anticipated operating conditions identified in [Ref.1). APPLICABLE The analyttral methods and assumptions used in evaluating fuel i
$ Aft 1Y system design are presented in $$AR, Chapter 4 and in ;
ANALY$[$ Reference 1. The fuel ensembly is designed to ensure (in conjunction with the core nuclear and them)) hydraulic design, i plant equipment, instrumentation and protection system that fuel damage will not result in the release of radioact we materials in excess of the guidelines of 10 CFR 20, $0 and 100. The mechanisms which could cause fuel damage during operational transients and which are considered in fuel evaluations are (1) rupture of the fuel rod cladding caused by strain from the relative expansion of the UO, pellet and (2) severe overheating of the fuel rod cladding cauted by inadequate cooling. A 'salue of 1% plastic strain of the 21rceloy claddin has been defined , as the limit below which fuel damage caused y overstraining of the fuel cladding is not expected to occur (Ref. 2). The ensures that Safety fuel Limit damage MINIMUM caused by' severe CRITICAL POWER overheettr.g of the RATIO (HCPR) fuel rod cladding is avoided and is discussed separately in the Bases for LCO 3.2.2. Fuel design evaluations have been performed and demonstrate that the 1% plastic strain fuel design liniit is not exceeded during continuous operation with LHGR's up to the operating limit specified in the [CORI OPERATING LIMITS REPORT). The analysis also includes allowances for short ters transient operation above the operating limit to account for anticipated operational occurrences, plus an allowance for densification power spiking. (continued) ABWR B 3.2 9 $/31/89 l O l_ Amendr* tat 9 IMS9
2M6100AP handard Plant m.2 O i LNGR i B 3.2.3 kHfi ftantinuedi _ APPLICABLE LNGR satisfies the requirements of Selection Criterion 2 of the
$AFETY NRC Interim Policy 5tatement on Technical Specification ANALY5t$ leprovements as documented in Reference 3. l (continued)
LCO LHGR is a basic assumption in the fuel design analysis. The . fuel has been designed to operate at rated core power with ' Sufficient design margin to the LHGR calculated to cause 1% cladding plastic strain. The operating limit to accomplish this objective is specified in the [ CORE OPER/. TING LIMITS REPORT). APPLICABILITY The LHGR liett is dertved from fuel design analysis that is Itatting at high power level conditions. At core thermal power levels less than 25% of RATED THERMAL POWER (RTP), the reactor will be operating with substantial margin to LNGR limits and therefore, the specification is only required when operating at or above 26% of RTP. ACTIONS M O:
$hould any LHGR exceed the required limits, an initial condition of the fuel design analysis will not be met. There-fore, prompt action should be taken to restore the LHGR to within the required limits such that the plant will be opersting within analyzed conditions.
I U l if the LHGR cannot be restored to within the required limits in two hours, it is required to reduce THERMAL POWER to < 25% of RTP. Operation below 25% of RTP results in sufficient margin I to the required limits. comoletion Times t The Completion Times are based on industry accepted practice and engineering judgement considering the time to reasonably l complete the Required Action. I (continued) , ABWR B 3.2 10 $/31/89 O Amendment 9 3g310
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I ABM * * ' " r i Standard Plant Rev.A ? I LHGR B 3.2.3 RAtt$ frontinued) f
$URV[lLLANC[ 1R 3.2.3.1 t REQUIREMENTS LHGR is required to be initially calculated within 12 hours ,
after THERMAL POWER has onceeded 25% of RTP and than daily r thereafter. It is compared to the specified limits to assure ! that the reactor is operating within the assumptions of the safety analysis. The daily surveillance requirement is based : interingjudgementconsideringtheslowchangesinpower ! onen!.bution. distr The !! hour allowance after exceeding 25% of RTP is acceptable given the large inherent margin to operating limits at lower power levels. REFERENC(5 1. [To be provided for non CE fuel).
- 2. NUR[G 0800, Standard Review Plan 4.2 ' Fuel System .
Design'. Section ll.A.!(g).
- 3. NEDO 31466. ' Technical Specification $creening Criteria Application and Risk Assessment *, November 1987. ,
O n i l 1 ABWR B 3.2 11 5/31/89 O ll Amendment 9 g g gg l l
t 1 b M
- i. === l
- r. m..a.o..t hA ;
16A.6 B3.3Instrunnentation . (m i
\ (IJter) i: i t
i I f f i t I
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i t i 4 ( t k t 1 i I l i ,= t 1 i
- 1. \
l Amendment 9 gggy
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i i- MN 23A6100AF Standard Plant m. A j SECTION 16A.7 .' O coureurs Section 21111 East 16A.7.1 R3.4.1 Racimdatlan Pumps Operatiaqr 16A.71 J 16A.7.2 R3.4.2 Safen/ Relief Yahes 16A.7 5 16A.73 R3.4.3 Operational h 16A.7 9 5 16A.7.4 R3.4.4 SpaelSc AetMtv 16A.713 16A.7.5 R3.4.8 Realdmal Heat Rasmoval Shutdown 16A.716 R3.4.6 Reactor Coolant System Pmasure/ Temperature i 16A.7.6 Limits 16A.719 , f 16A.7.7 B3.4,7 Reactor Steam Dome Pressure 16A.7 26 I h i O IM.7-ii Amendmem 9
ABWR zwiew Standard Pla'nt u 16A.7 B3A Reactor Coolant System
,m 16A.7.1 53A.1 Recierulation Pumps Operating Recirculation loops Operating 8 3.4.1 B 3.4 REACTCA COOLANT $Y$i[M B 3.4.1 kacirculation pamas Daarating Ratti BACKGROUND The reactor coolant recirculation system is designed to provide a forced coolant flew through the core to remove heat from the
- fuel. The reactor coolant recirculation system consists of ten rectreviation pumps internal to the reactor vessel. The !
reactor internal pumps (RIPS) are driven by wet motors that . protrude from the bottom of the reactor vessel. Each motor has ; its own external heat exchanger that is coolsd by reactor ! butiding cooling water. The pump motor casings are part of the reactor coolant pressure boundary and are located in the lower drywell area. The pumps are reactor vessel internals, f The retirculated coolant consists of saturated water from the steam separators and dryers that has been subcooled by incoming feedwater. inis water passes down the annulus between the reactor vessel rall and the core shroud to the inlet of the ten RIPS that are located equidistant around the bottom of the core shroud (or pump deck). The total core flow passes through the RIPS into the lower plenum, up through the orifices of the lower core plate, and then up through the core. Each pump G motor is driven by an Adjustable Speed Drive (ASD) that is individually started and controlled from the main control room. The total core flow can also be controlled automatically, with all pumps in parallel, by the master flow controller. The ten RIPS are powered by four separate electrical busses. Two busses each supply three RIPS and the other two busses each , supp'y 2 RIPS. There is also an WG set between each of the l busses supp' lying 3 RIPS and the respective ASDs (2 M/G sets total). This provides for slower flow coastdown character. 1stics during certain transtant events. APPLICABLE The operation of the reactor coolant retirculatien system is an
$AF[TY initial condition assumed in the design basis Loss of Coolant I ANALY$[$ Accident (LOCA) (Ref. 1). During a LOCA the operating RIPS are all assumed to trip at time zero due to a coincident loss of offsite power. The subsequent core flow coastdown will be immediate and rapid because of the relatively low inertia of the pumps. Also, for conservatism, no credit is taken for the inertia of the two WG sets that feed six of the RIPS.
(continued)
)
ABWR B 3.4 1 $/31/89 l O LV l Amendment 9 16A31 l
M Reamulayf plaget Rav. A ; l i Recirculation loops Operatin i B 3.4.
- RAtte fenntinued)
APPLICA%LE With at least nine of the ten RIPS in operation the LOCA 1 SAFETY analysis includes all potential power and flow operating points i ANALY$[$ free which an event might be initiated. With eight or less (Continued) RIPS in operation the LOCA analysts assumptions 60 not include all potential operating states so that additional restrictions are necessary regarding reactor power based on the number of pumps actually operating. The recirculation system is also assumed to have sufficient flow coastdown characteristics to maintain fuel thermal margins during abnormal operettoral transients (Ref. 2) which are ; analyzed in Chapter 16 of the F$AR. Recirculation Loops Operating satisfies the requirements of Selection Criterion 2 of the NRC Interim Policy $tatement on Technical Specification leprovements as documented in Reference 3. LC0 At least nine RIPS are required to be in operation to ensure ' during a LOCA the assumptions of the LOCA analysis are satts. , fled without restriction. With less than nine RIPS in opera-tion, all potential power and flow operating states have not been accounted for in either the LOCA or transient analysts. , Therefore, certain restrictions apply depending on the number of RIPS operating. APPLICABILITY Requirements for operation of the reactor coolant recirculation system are necessary during MODES I and 2 since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur. During other conditions, the consequences of an accident are reduced and the coastdown characteristics of the reactor internal pumps are not important. ACTIONS A.1. B.1. C.). D.1 With less than nine RIPS in operation the reactor power level must be restricted so that the assumptions of the LOCA and transient analyses are met. With only seven or eight RIPS operating the power level is restricted to 90% and 95% of RTP, respectively. However operation may continue indefinitely. As noted. LCO 3.0.4 is not applicable for Required Actions A.1 or B.l. With less than seven pumps operating power level is e restricted even further and operation may only continue for a short time. (continued) ABWR B 3.4 2 5/31/89 O Amendment 9 1R72
. MM 23A6100AP Remmulard Plant u ,m b
Retirculation loops Operating B 3.a ! RAlti fremtinued) ACT10h5 A.1. a.1. c.1. D.1 tenetinune) (tontinutd) for the case of $ or 6 pumps running, reatter power must be reduced to less than in RPT because of potential stability torterns. With less than $ pumps operating, power must be reduced to less than $$ RPT due to the lett of detailed analysis of the actual flow distribution with less than half of the pumps in operation providing forced flow at higher poner levels. f.f.1. f.P.f. D.2.1. D 2.2 With less than seven RIPS operating the steady state power and flow theratteristics of the core have not been fully analyted. Therefore, even at reduced power levels, continued operation is allowed for only a short ties while an attempt is made to restore at least seven pumps to operating status. With less than seven pumps able to be restored to operating status within the Required templetion line, the reactor is required to be in MDD[ 3. In this condition, the RIPS are not required to be operating because of the reduced severity of design basis accidents and einimal dependence on the forced flow tharacteristics. O Comeletion Times All Completion Times are based on industry attepted practice and engineering judgement considering the nuober of available systers and the time required to reasonably complete the Required Action.
$URY(ILLANC[ 1R 3.4.1.1 RIOUIRIMINTS This surveillante requirement ensures that the reatter power level is within the assumptions of the applicable analyses based on the number of pumps attually operating. Operating emperience has demonstrated that a 24 hour frequency for this type of surveillante is adequate.
l (Continued) ABWR B 3.4 3 $/31/89 Amendment 9 MA73 l
i N Rem 'ryd Plant 2M610MP u.e . l I Gi Recirculation loops Operatin B 3.4. kAlft frontinued) l RtF(RENC(5 3. ABWR 55AR. Section 6.3.3.
- 2. NtD0 31466 ' Technical Specification Screening Criteria Application and Risk Assessment'. November 1987.
l ? ( l l t ast' L. , t ABWR B 3.4 4 $/3}/e9 O Amendment 9 1(A.7 4
i MM 21A6100AP Standard Plant ha i 16A.7.2 33.4J SafeQ/ReuefValves , V i
$/RVs j B 3.4.2 i B 3.4 R[ ACTOR COOLANT $YSTEM f
B 3.4.2 Safetv/talief Walvet RAtts BACKGROUND The A$ME Boiler and Pressure Vessel Code requires the reactor . pressure vessel be protected from overpressure during upset conditions. As part of the nuclear pressure relief system, the size and number of safety / relief valves ($/RVs) are selected such that peak pressure in the nuclear system will not exceed j the ASME Code limits for the reactor coolant pressure boundary. 1 The $/RVs are located on the main steam lines between the reactor vessel and the first isolation valve within the , drywell. The $/RVs can actuate by either of two modes the i safety mode or the relief mode. In the safety mode (or spring note of operation), the direct action of the steam pressure in ., s the main steam lines will act against a spring loaded disk that 1 will pop open when the valve inlet pressure exceeds the spring force plus the weight of the disk, in the relief mode (or ' power actuated mode of operation), a pneumatic piston / cylinder and mechanical linkage assembly are used to open the valve by
- overcoming the spring force; however, a steam inlet pressure of at least 50 psig is required to overcome the weight of the disk which is not directly connected to the spring. The pneumatic operator is arranged 50 that its malfunction will not prevent the valve disk from lifting if steet inlet pressure reaches the spring lift set pressures. Eacn 5/RV discharges steam through a discharge line to a the suppression pool. point Eight below the minimum of the $/RVs waterthe that provide level in relief function are part of the Automatic Depressurization System (ADS) specified in LCO 3.5.1.
APPLICABLE The overpressure protection system must accommodate the most l SAFETY severe pressuritation transient. Evaluations have determined i ANALYS[$ that the most severe transient is the closure of all main l steam line isolation valves (MSlys) followed by reactor scram l on high neutron flux (i.e., failure of the direct scram associ. ated with MS!V position) (Ref.1). Twelve of the eighteen
$/RVs are assumed to operate in the safety mode and no credit is taken for the relief mode of operation. The analysis results indicate the design $/RV capacity is capable of main.
taining reactor pressure below the ASME Code limit of 110% of vessel design pressure (110% x !!50 psig = 1375 psig). Refer- ' ence ! discusses additional events which are expected to actuate the $/RVs. From an overpressure standpoint, these events are bounded by the MSIV closure with flux scram event described above. (continued) ABWR B 3.4 5 5/31/89 lO Aspendment 9 16A.7 3 l
MM 21A6100AP ; Reamulmed Piment n<A O'
$/RVs ;
B 3.4.2 RASES frontinued) APPLICABLt $/RVs satisfy the requirementa of Selection Criterion 3 of the
$AFETY NRC Interim Policy Statement on Technical Specification ANALYSES Improvements as documented in Reference 3. ,
(continued)
?
Twelve $/RVs are required to be OPERABLE in the safety mode. LC0 The requirements of this Lt0 are applicable only to the capabil-ity of the $/RVs to mechanically open to relieve excess pres. suse. In Reference 1. an evaluation was performed to establish the parametric relationship between the peak vessel pressure and the number of OPERABL[ $/RVs. The results show that with a minimum of 12 $/RVs OPERABLt in the safety mode, with setpoints in a distribution equivalent to, or conservative with respect to, the minimum requirements of SR 3.4.2.1. the ASME Code limit of 1375 psig is not exceeded.
- The $/RV setpoints are established to ensure the A$ME Code t limit on peak reactor pressure is satisfied. The A$Mi code specifications require the lowest valve be set at or below vessel design pressure (1250 psig) and the highest safety valve be set so the total accumulated pressure does not exceed 110% !
of the design pressure for overpressurization conditions. The transient evaluations in the ISAR are based on these setpoints, but also include the additional uncertainties of + 1% of the i nominal setpoint to account for potential setpoini drift and to provide an added degree of conservatism. Operation with less valves OPERABLE than specified, or with
- setpoints greater than specified, could result in a mere severe reactor response to a transient than predicted, possibly resulting in the ASME Code limit on reactor pressure being
- exceeded.
APPLICABILITY The specified number of $/RVs must be OP[RABLE in MODES 1. 2 and 3 since there is considerable energy in the reactor core and the limiting design basis transients are assumed to occur. = The $/RVs may be required to provide pressure relief to ' discharge energy from the core until such time that the Residual Heat Removal (RHR) system is capable of dissipating the heat being generated. In MODE 4. decay beat levels are low enough such that the RHR system is adeguate, and reactor pressure levels are low enough such that the overpressure limit t (continued) ABWR B 3.4 6 e: 5/31/89 Amendment 9 1W6
3 l MM 21A610QAP u Remmulard Plant i n N I i
$/kVs B 3.4.2 RAlf1 frontinued)
APPLICABILITY cannot be challe ed by assumed operational transients or (continued) accidents. In E 5, the reactor vessel head is unbolted or removed and there is no reactor coolant pressure boundary : (RCPB). The $/RV function is not needed duriPg these conditions. ACTIONS A.1. A.2 With any of the required $/RVs inoperable, the reactor must be in MODE 3 in 12 hours and in MODE 4 in 36 hours. With less than the minimum number of $/RVs OPERABLE, a transient may result in the violation of the ASME Code limit on reactor pressure. It is therefore necessary for the plant to be in a - condition where the $/RVs are not required. The Completion . Times allow for a controlled Shutdown of the reactor without placing undue stress on plant operators or plant systems. SURVE!LLANCE' 1R 3.4.2.1 REQUIREMENTS This surveillance requirement demonstrates that the $/RVs will I open at the pressures assumed in the safety analysis of Refer-ence 1. The demonstration of the $/RV lif t settings must be performed during shutdown and in accordance with the provisions of SR 3.0.5. The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.
$R 3.4.2.2 A manu.1 actuation of each $/RV is performed to verify the valve is mechanically functioning properly and no blockage r i
exists in the valve discharge line. This can be demonstrated by the response of the turbine control or bypass valves or by change in the measured steam flow or any other method suitable to verify gross steam flow through the discharge line to the suppression pool. Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. 4 Sufficient time is therefore allowed after the required pres. l sure is achieved to perform this test. Adequate pressure at u whichthistestistobeperformedis[ }psig(thepressure recommended by the valve manufacturer). Plant startup is (continued) l ABWR B 3.4 7 5/31/89 Amendment 9 IIA 7-7 l
Mkk- 2M6100AP . had-rd Plant hA
$/RVs B 3.4.2 RAlfs fcentinued)
$URiflLLANCE 1R 3.4.f.2 fenntinued)
REQUIREMENTS (continued) allowed prior to perforwing this test because valve OPERABILITY : and the setpoints for overpressure protection are verified, per ! ASME requirements, prior to valve installation. If the valve fails to actuate due only to the failure of the solenoid but is capable of opening on overpressure, the $/RV is considered OP[RABLt. Surveillance Frenuancies in general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the i systeVcomponent status. REFERINCES 1. ABWR $$AR, Section 5.2.2. .
- 2. ABWR $$AR, Chapter 15.
- 3. NEDO 31466, *1echnical Specification Screening Criteria Application and Risk Assessment', November 1987. ,
l t ( ABWR B 3.4 8 5/31/69 O Amendment 9 1M.7-8 l
r MM 21A6100AP me==dard Plant ' u 16A.7.3 R3.4.3 Opeestlesal Lankage V , e Operational Leakage B 3.4.3
- B 3.4 R(ACTOR COOLANT SYSTEM !
B 3.4.3 Onerational Lankane BAlf$ BACKGROUND The reactor coolant system includes systems and components that. contain or transport fluids to or from the reactor core. These systems fort a major portion of the nuclear system process . barrier. The pressure containing components of the reactor 1 coolani system, including the portions of the system out to and 4 including isolation valves, are defined as the Reactor Coolant ! Pressure Boundary (RCPB). Limits on leakage from the RCPB are required to ensure appropriate action can be taken before the integrity of the nuclear system process barrier is impaired ! (Ref.1). The safety significance of leaks from the RCPB can very widely < depending on the source of the leak as well as the leakage rate and duration. Therefore, detection of leakage in the drywell : is necessary. Identified Leakage is defined as the leakage into closed systems, such as pump seal or valve packing leaks that are captured flow metered and piped to a sump or collect. , ing tank. Also, leaka e into the containment atmosphere from b sources that are spect itally.lecated and known not to inter. fere with the operation of Untientified Leakage detection or not to be a flaw in the RCPB are considered Identified leakage. Unidentified leakage is collected in the drywell floor drain ' sump along with the normal design leaksge from the control rod drives, valve flange leakage, floor dratns, closed cooling ; water system and drywell cooling unit drains. Methods for separating the Identified Leakage from the Unidentified Leakage are necessary to provide prompt and quantitative information to the operators to permit them to take corrective action. A limited amount of leakage is expected from auxiliary systems ' within the drywell that cannot be made 1001,leaktight. If leakage occurs from these paths. it should be detectable and isolated from the drywell atmosphere if possible, so as not to 1 mask any potentially serious leak should it occur. 1 APPLICABLt The allowable leakage rates from the reactor coolant system l
$AFfiY have been based on the predicted and experimentally observed l j
ANALY$[$ behavior of pipe cracks. The norsally expected background leakage due to equipment design and the detection capability of the instrumentation for determining system leakage were also ; considered. The evidence obtained from experiments suggests for leakage somewhat greater than specified for Unidentified (continued) ABWR B 3.4 g 5/31/89 l Amendment 9 IfA.7 9 l' l I 1 i
' MM 21A6100AP w
Riandard Plant O Operational Leakage B 3.4.3 BASES frontinued) APPLICABLE Leakage. the probability is small the imperfection or crack
$AFETY associated with such leakage would grow rapidly. The ANALYSES Unidentified teakage rate limit is estabitsbed at i gpm to (continued) allow time for corrective action before the process barrier could be significantly compromised. This limit is a small fraction of the calculated flow from a critical crack in the primary system piping. Based on track behavior from experi-mental programs (Ref. 2 and 3) it is estimated that leak rates of hundreds of gpm will precede crack instability (Ref. 4).
There are no applicable safety analyses that assume the Total Leakegt Ilmit. The Total Leakage limit is specified based on consideration of inventory makeup capability and sump 8 capacities. Operational Leakage satisites the requirements of Selection Criterton 1 of the NRC Interim Policy Statement on Technical Specification Improtements as documented in Reference 5. LCO No Pressure Boundary leakage is allowed since the potential exists for a break in the PCPB and a loss of substantial inventory. Leakage should not he allowed to increase signif t. cantly without a thorough examination of the source of the leak. , The Total Leakage rate consists of all leakage, identified and [ unidertified, that flows to the drywell floor drain and equip. I ment drain sumps. The Unidentified Leakage rate limit is based I on the leakage rate above background leakage rates. 1-l APPLICABILITY The potential for RCPB leakage is greatest when the reactor is pressurized. Under these conditions, high stresses are j spplied to the system piping resulting in the potential for t crack growth and possible failure of the RCPB. Therefore, detection of RCPB leakage is recuired during MODES 1, 2 and 3. In MODES 4 and 5, operettonal b akage limits are not reoutred, since the reactor is not pressurized and the potential for leakage and possible pressare boundary failure is reduced. l (continued) I ABWR B 3.4 1D 5/31/89 Amendment 9 IfA710
~ .-.. - -- -- .-
i 2M6100AP { Readerd Plant u l i t )
\/ l l
Operational Leakage ! B 3.4.3 RA$tt feartinued) AC110N$ L1 With either the Total Leakage Unidentiflod Leakage, or both i greater than the required limits, actions should be taken to ' identify the source of the leak and detemine the significance. Because the leakage limits are conservatively below the leakage ; that wop 1d constitute a critical crack site, a limited time is
- allowed to evaluate the situation. If a change in Unidentified !
Leakage has been adequately identified and quantified, it may l be reclassified and considered as identified Leakage. However, the Tctal Leakage limit would remain unchanged. B.1. B.2 } If the LEAKAGE cannot be restored to within the required limits, the reactor should be in MODE 3 and subsequently in MODE 4. If Pressure Boundary leakage occurs there is the potential that the flaw in the RCPB could eventually result in a pipe break or cther LOCA. Since the area being monitored is inaccessible, the reactor must be in MODE 3 and subsequently in F MODE 4 to allow a visual inspection to determine the source of the leak. Comoletion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available , systems and the time required to reasonably complete the ' Required Action. SURVIILLANC[ SR 3.4.3.1 REQUIREMEN15 The reactor coolant system LEAKAGE is monitored by a variety of systems designed to provide alarms when leakage is indicated and to quantify the various types of leakage. Leakage detec. , tion is discussed in more detail in the Bases for LCO [ Leak ! DetectionInstrumentation). Sump level or flow rates are typically monitored to determine actual leakage rates. How. ever, any method may be used which quantifies leakage within the guidelines of Reference 1. Operating experience has ; demonstrated that a 12 hour frequency for this surveillance is adequate. (continued) ABWR B 3.4 11 $/31/89 [t Amendment 9 IM111
- .. , -- , - - - , + , - , - . . . .
. - * . . _ ,n ,-- ,
MM 21A6100AP !a*==tard Plant a- A i i Ol , Operational Leakage B 3.4.3 , t RAlti fenntinued) REFER [NCl3 1. Regulatory Gu1Ce 1.45, ' Reactor Cociant Pressure Boundary Leakage Detection Systems', May 1973. '
- 2. GEAP 5620, ' Failure Behavior in ASTM A106 Pipes Containing i Axial Through Wall Flaws,' April 1968. l
- 3. NUR[G 76/067, ' Investigation and Evaluation of Cracking in ;
Austenitic $tainless steel Piping of Boiling Water Reactor . Plants,' October 1975.
- 4. ABWR $5AR, Section 5.2.5.5. ,
- 6. NED0-31466, ' Technical Specification Screening Criteria -
Application and Risk Assessment', November 1987. i l l O. i 1 i f i i ABWR B 3.4 12 6/31/e9 O Amendment 9 IfA 712
. . - . . - . . _ . . . . . . . . . . - . . _ . . _ . . . _ ~ . . _ _ _ _ _ _ _ _ _ . . . . . . _ _ . _ _ _ _ . _ . _ _ . , _ ~ . . . _ . . _
MM 25A61%AP Standard Plant _ m : 16A.7A R3A.4 SpecWe Activisy v Spectitc Activity I B 3.4.4 - B 3.4 REACTOR COOLANT $YSTEM , B 3.4.4 13pcifte Activity kAtti ' SACKGROUND During circulation, the reactor coolant acquires radioactive matertal due to release of fission products into the coolant and activation of crud particles in the reactor coolant. These t radioactive materials in the reactor coolant could contribute to release of radioactive materials into the environment during design basis accidents. ' i limits on the maximum allowable level of radioactivity in the reactor coolant are established to assure, in the event of a release of any radioactive material to the environment during a design basis accident radiation doses are maintained within the limits of 10 CFR 100. APPLICABLE Analytical methods and assumptions involving radioactive
$AftTY material in the prinary coolant are presented in $$AR Chapter i ANALYSES !$. Accident Analyses. The specific activity in the reactor / coolant (source tenn) is an init14) condition assumed for l \ evaluation of the consequences of an accident due to a main '
steam line break (MSLB) outside the containment. No fuel damage is postulated in the MSLB accident, and the release of radioactive matertai to the environment is postulated to be terminated by complete closure of the main steam line isolation + valves (MSIVs). This release forms the basis for determining off site doses (Ref. 1). The limitations on the specific activity of the primary coolant ensure the 2 hour thyroid and whole body doses resulting from a s.atn steam line failure ' outside the containment during steady state operation will not - exceed 10% of the dose guidelines of 10 CFR 100. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that soecific site parameters, such as site boundary location and meteorological conditions, were not considered in this evaluation. Specific Activity satisfies the requirements of Selection Criterion 2 of the NRC Interim Policy Statement on Technical , Specification improvements as documented in Reference 2. (continued) , 1 ( ABWR B 3.4 13 $/31/89 Amendment 9 IIA 7'I3 i I l l
.M 2146100AP Riandaral Plant m.i 1 l
O; . i
$pecific Activity f t 3.4.4 ;
i taits fenetinued) , LCD The primary coolant specific activity level of r 0.2 micro-curies per gree D0st EQUIVAttWT l 131 is reoutred to ensure the source ters assumed in the safety analysis of the MSLB is not exceeded such that any release of radioactive material to the environment does not exceed 10 CFR 100 limits. APPLICABILITY Liettations on levels of primary coolant radioactivity are l applicable during MODE 3 and MODt3 2 and 3 with any main steam > line not isolated since there is an escape path for release of radioactive material from the coolant to the environment in the ' event of a MSLB outside of the primary containment. During MODES 4 and 6. no lietts are required since the reactor is not pressurized and the potential for leakage is reduced. I ACTIONS A.1. A.P. B.1. B.2 A primary coolant specific activity level > 0.2 pCi per gram DOSE EQUIVALENT l 131 indicates the presence of some abnormal. ity in plant operations. The range between 0.2 #C1 and 4.0 #Ci 7 I is acceptable for up to 48 continuous hoars to account for
- potential todine spiking that may occur following changes in i THERMAL POWER. Increased surveillance of the reactor coolant specific activity during this period is required to closely >
monitor the condition and determine if additional limits are exceeded. If coolant specific activity cannot be restored s 0.2 #Ci within 48 hours, or when coolant activity is e 4.0 #Ct. the main steam lines are required to be isolated. This action precludes the possibility of release of radioactive material to the environment in excess of the requirements of 10 CFR 100, during a postulated MSLB accident. Comoletion Times All Completion Times are based on industry accepted practice I and engineering . judgement considering the number of available systems and the time required to reasonably complete the Required Action. (continued) t
'd ABWR B 3.4 14 $/31/89 G
Amendnent 9 16A114
i
'MN uwow Standard Plant m.v. A e
h
/'~'T !
(s_,/ - l Specific Activity B 3.4.4 kAits (continuedi
$URV[lLLANC[ 1R 3.4.4.1 R[0VIREMEN15 Isotopic analysis for D0ft t0VIVAltNT l.13) concentration is >
necessary to determine that the specified maximum primary
~
coolant activity is s 0.2 microcuries per gram DO$E E0V! val [NT i 1 131 during steady state operation. Operating experience has demonstrated that a 31 day frequency for this surveillance is ; adequate. ' turve111ance Freauencies In general. surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of 6 performing the test and a likelihood of a change in the l' system / component status. R[f[RENtt$ 1. ABWR $$AR. Section 16.6.4.
- 2. NIDO.31466 ' Technical Specification Screening Criteria Application and Risk Assessment', November 1987. j
, O i i t i I l i l ABWR B 3.4 l$ 5/31/89 O V
/unendment 9 IEA.7-15 i
l l I l '
MM 21A6WQAP w Remmulard Plant ; i 164.7.5 R3.4.5 Reeldest Heat Rosmoval. Shutdown e; , RHR + $hutdown i 8 3.4.5 ! B 3.4 'REAC10R COOLANT $YSTEM : i 8 3.4.5 hesidual Heat tamoval - shutdown Batts BACKGROUND 1rradiated fuel in the reactor pressure vessel (RPV) generates decay heat during nomal and abnormal shutdown conditions, , potentially resulting in an increase in the temperature of the ! reactor coolant. This decay heat is required to be removec such that the reactor coolant temperature can be reduced to or maintained at s 200*F in preparation for performing refueling, ' maintenance operations or for maintaining the reactor in cold shutdown conditions. Systems capable of removing decay heat j tre therefore required to perform these functions. J The three shutdown cooling loops of the Residual Heat Removal (RHR) system provide decay heat removal. Each loop consists of a motor driven pump, a heat exchanger, and the associated , piping and valves. Each loop has its own dedicated suction i from the RPV. [ach pump discharges the reactor coolant, after , it has been cooled by circulation through the respective heat , exchanger, to the reactor via feedwater line 'A' for RHR ; Loop 'A' and via the RHR Low Pressure flooder $pargers for
- Loops 'B' and 'C'. The RHR heat exchangers transfer heat to e the Reactor Building Cooling Water System (LCO 3.7.1 and LCO 3.7.2). The RHR shutdown cooling mode is a manually controlled system.
APPLICABLE Decay heat removal by operation of the shutdown cooling aode . SAFETY of the RHR system is not required for mitigation of any events ANALY$ES or accidents evaluated in the safety analyses. However, the NRC Interim Policy Statement (Ref. 1) requires the RHR system be retained in the Technical $pecifications even though none of the selection criteria (Ref. 2) were satisfied. LCO Two shutdown cooling subsystems are required to be OPERABLE. An OPERABLE RHR shutdown cooling subsystem consists of one RHR pump, one heat exchanger, and the associated piping and valves. Additionally, each shutdown cooling subsystem is considered in the OPERABLE if it can shutdown cooling modebe for manually removalaligned (remote of decay heat. InorMOD local)ES 3 and 4, one shutdown cooling subsystem of the RHR can provide (continued) ABWR B 3.4 16 $/31/89 Amendment 9 IfA716
l l MM 2M6100AP !
. Standard Plant a" , ;[~} J
(/ RHR Shutdown B 3.4.5 j llAtFK fecntinued) i LCO the required cooling to maintain the desired temperature. Two (continued) subsystems are required to be OPERABLE to provide redundancy. Operation (either continuous or intermittent) of one subsystem can maintain and reduce the reactor coolant temperature as ; required. , APPLICABILITY Decay heat removal at reactor pressures above the RHR cut.in permissive pressure is typically accomplished by condensing , steam from the RPV in the main condenser. When the reactor ' pressure is below the RHR cut.in permissive pressure, the RHR system may be operated in the shutdown cooling mode. Operation of the RHR system in the shutdown cooling mode above this , pressure is not allowed because the coolant pressure may exceed the design pressure of the shutdown cooling piping. Operation of a subsystem to remove the decay heat may be required to either reduce or maintain coolant temperature. Continued operation of a shutdown cooling subsystem may be required to reach and maintain reactor coolant temperatures s 200*F, which corresponds to MODT 4. The requirements for decay heat removal ; in MODE 5 are discussed in LCO 3.9.8 and LCO 3.9.9. ACTIONS M With one of the required RHR shutdown cooling subsystems inoperable for decay heat removal, the remaining OPERABLE subsystem can provide the necessary decay heat removal. However, the overall system reliability is reduced. For this reason, continued operation is allowed for a limited time only. 2 hours. B B.I. P.2 If one of the inoperable subsystems cannot be restored to OPERABLE status or with all three subsystems inoperable, an alternate method of decay heat removal is required to be made i available for each recutred, tut inoperable, subsystem to l restore cooling capability. The required cooling capacity of the alternate method should be ensured by verifying (by calcu. lation or demonstration) its capability to maintain or reduce l.- temperature. Decay heat removal by ambient losses can be > l considered as contributing to the alternate method capability. (continued) ABWR B 3.4 17 5/31/89 Amendment 9 16A.717
t MM 21A610aAP Rianulard Plant w l l RHR Shutdown B 3.4.6 , hift fenntinuedi ACTIONS ramalation Timet 6 (Continued) All Completion Times are based on industry accepted practice ; and engineering h egement considering the number of available systems and the time required to reasonably complete the Required Action. , SURVIILLANCE 1R 3.4.5.1 + REQUIR[M[Ni$ Verification that all valves of the required RHR shutdown cooling subsystem are in the correct position ensures the proper flow path. Valves not in the correct position must be capable of manual realignment either from the control room or ; at the valve location. Verification of this capability is , provided b." ..tuation of the valve from the control room or the current in6ervice inspection reports. Operating experience has demonstrated that a 31 day frecuency for this surveillance is adequate. Because some of the required valves are interlocked closed when above the CHR cut.in permissive pressure, an allowance is provided to test the valves within !! hours after . pressure has been reduced below the cut.in permissive pressure. , This allows conditions to be established under which the test may be performed.
- R(FIR (NC[$ 1. 52FR3788, Proposed policy Statement on Technical Specification Improvements for Nuclear Power Reactors, February 6, 1987.
- 2. NEDO 31466,
- Technical Specification $creening Criteria l Application and Risk Assessment,' November 1987.
ABWR B 3.4 18 5/31/89 O Amendment 9 16A.718 , . - . _ . . _ - _ , . -. . . _ . __ _ . _ ~ . . _ - , _ . _ . . - ~ _ _ _
ABM iwiour Re==A-eti Plant m-2 1 164.74 R3.44 Reactor Coolant SystemPevssure/Tempersture units !
/3 i
() {, l RC$ Pressure / Temperature Limits , 8 3.4.6 t 8 3.4 REACTOR COOLANT SYSTEM j B 3.4.6 ""*ter Caelant tratan Pressura/Tsamaratura Limits hMft # BACKGROUND All components in the reactor coolant system (RC$) are designed d to withstand the effects of cyclic loads due to system temper- ; sture and pressure changes. These cyclic loads are introduced by normal load f.ransients, reactor trips, and startup and t shutdown operations. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic ; operation. ; The purpose of this specification is to establish operating limits that provide a wide margin to brittle f ailure of major ! piping and pressure vessel components of the Reactor Coolant Pressure Boundary (RCPB). Of the siajor components within the ! RCPB, the reactor vessel is the component most subject to ; brittle failure and therefore the component for which the technical specification ilmits is most pertinent. ; The basis of the pressure and temperature (PT) ilmits is found Os in Appendix G to 10 CFR 50 (Ref. 1). Appendix G requires that limits be estabitshed, and that the limits be based on specific l fracture toughness requirements for RCPB materials such that an t adequate margin to brittle failure will be provided during operational occurrences. 10 CFR 50 Appendix G mandates the use of ASME 5ection Ill. Appendix G (Ref. 2). The concern addressed by Appendix G is that undetected flaws could exist in the RCPB components, which if subjected to , i unusual pressure and/or thermal stresses, could result in non ductile (brittle) failure. Certain reactor coolant system PT combinations can cause stress concentrations at flaw loca. tions which in turn could cause flaw growth, resulting in failure before the ultimate strength of the material is attained. Flaw growth is resisted by the material toughness. I Toughness is a property that varies with temperature and is l lower at room temperature than at power operation. Further- t more, the material toughness is affected by neutron fluence l which causes the steel ductility to decrease. The effect of I fluence is cumulative, and ductility steadily decreases with exposure time. Only the vessel beltline region is in a high fluence area. Toughness is also dependent on the chemistry of . the base metal and its impurities. l (continued) ABWR B 3.4.lg 5/31/89 Amendment 9 IM 719
1 MM 21A6100AP
, Standard Plant m t i
i RCS Pressure / Temperature Limits B 3 4.6
-{MESfcontinued)
BACKGROUND- Linear elastic fracture anchanics (L(FM) methodology, following (continued) the guidance given by 10 CFR 50 Appendix G. ASME Section !!! Appendix G, and Regulatory Guide 1.99, is used to determine the stresses and meterlal toughness at locations within the RCPB. '; Although any region within the pressure boundary is subject to non ductile failure, the regions that provide the most restric. tive limits are the vessel closure head flange, the feedwater nozzles, the control rod drive nozzles, and the vessel. beltline. One indicator used to indicate the temperature effect on ductility is the nil ductility temperature (NDT). The NDT for the steel alloy used in vessel fabrication has been established . by testing. The NOT is a temperature at which it can be said that fracture failure may occur below the NDT. Ductile failure may occur above the NDT. The exact temperature value-is not very precise. Consequently a reference temperature (RT n ) has been established by experimental means. TheneutroneJr,ittle-ment effect on the material toughness is reflected by increas- g ing the RT es exposure to neutron fluence increases. In effectthegeperatureatwhichbrittlefailurecanoccur N increases. The actual shift in RT of the vessel material will be establis?.edperiodicalgiduring operation by removing and ' evaluating in accordance with ASTM E185 73 and 10 CFR 50, Appendix H. irradiated reactor vessel material specimens installed near the inside wall of the reactor vessel in the core area. The operating limit curves shall be adjusted, as required, on the basis of the specimen data and recommendations of Regulatory Guide 1.99 (Ref. 3 and 4). l The PT curves are composite curves established by superimposing limits derived from stress analyses of those portions of the , reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, ons location within the geometry of the reactor vessel will dictate the most restrictive limit. Across the entire pressure / temperature span of the limit curves, some locations . are more restrictive, and thus the curves are composites of the most restrictive regions. The curves have been developed for heatup, in service leak and hydrostatic testing, and cooldown in conjunction with stress analyses for a large number of operating cycles and provide a conservative margin to non. ductile failure. Although they have (continued) ABWR B 3.4 20 5/31/89 O Amendment 9 16A.7 20
n, , ABWR
. ge 2mione.
dard Plant. h^
.p l.
RCS Pressure / Temperature Limits B 3.4.6 BASES fcontinued) i BACKGROUND been created to provide limits for these specific normal (continued) operations, they also can be used as a basis for determining if evaluations are necessary for abnormal transients which can begin from power operation. ASME Section XI Appendix E (Ref. 5) provides a recommended methodology for evaluating operating events which cause an excursion outside the normal limits. APPilCABLE The limits are not derived from design basis accident analyses SAFETY presented in the FSAR. but are prescribed as guidance to be ANALYSES used during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions which might cause undetected flaws to propagate, in turn causing non-ductile failure of the RCPB. RCS Pressure / Temperature Limits satisfies the requirem nts of the NRC Interim Policy Statement on Technical Specification Impri;vements for Nuclear Power Reactors Ref. 6). While none of the three Selection Criteria (Ref. 7)(directly apply, this specification preserves limits defining important boundaries for safe operation derived from the RCS stress analysis. l.. [' Criterion 2 is the most appropriate criterion because operation i outside of these boundaries is unanalyzed and may result in (' . RCPB failure. LCO Compliance with the following PT limits is required by this LCO:
- 1. Operation within the PT limit curves specified in the
$$AR, <
1 i 2. A maximum reactor cuolant heatup or cooldown of 100*F in l any one hour period, i
- 3. A maximum temperature change of 5 10' in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves in the SSAR,
- 4. The reactor vessel flange and head flange temperature 2 70*F when reactor vessel head bolting studs are under tension, (continued)
ABWR B 3.4 21 5/31/89 Amendment 9 16A.7 21
- MM 23A6100AP mndard Plant Ra A O
RCS Pressure / Temperature Limits B 3.4.6 BASE 1 frontinued) LCO 6. A temperature difference between the bottom head coolant (continued) temperature and the reactor pressure vessel coolant temperature of g 14b'F during reactor internal pump , startup, and The above limits define allowable operatino regions and permit a large number of operating cycles while also providing a wide margin to non-ductile failure. APPLIC?BILITY The potential for violating the PT limits exists at all times , when the reactor coolant system can be pressurized. The temperature rate of change limit can be potentially violated any time the reactor vessel material is a different temperature ,_ from a cooling source, . 1 ACTIONS A.1. A.2 As noted, Required Actions A.! and A.2 must be completed whenever Condition A is entered. The purpose of the Note is to give additional emphasis to the need to restore operation to ' the allowable condition and to also perform an evaluation of the effects of any excursion outside of the allowable limits. Restoration alone is insufficient because higher than analyze.1 stresses may have occurred and may have affected the RCPB integrity. l Restcration within the limits is appropriate because the RCPB l 1s placed in a condition that has been verified by stress I analysis. The action is in the proper direction to reduce RCPB L stress. If a recirculation pump has been started and the PT limits were not met, the pump does not need to be turned off unless the PT limits are not currently being met with the pump operating. The Completion Time limit of 30 minutes is based on engineering judgement. Most violations will not be so severe that the activity cannot be accomplished in this time in a controlled manner; however, if the activity cannot be accomplished, then a controlled shutdown must be initiated per Required Actions B.1 and B.2. (continued) ABWR B 3.4-22 5/31/89 O Amendment 9 16A.7-22
!AB M : asiaa^r Standard Plant ner .p j
r RCS Pressure / Temperature Limits B 3.4.6 BASES fcontinued) ACTIONS . A.1. A.2 (continued) . (continued) _In addition to restoration, an evaluation to determine if RCS operation may proceed is required. The purpose of the evalu-ation is to deterwine if RCPB integrity is acceptable and must-be accomplished before the event is reconciled. . ASME Section XI Appendix [ may be used to support the evalu-
-ation. If the evaluation cannot be accomplished in 72 hours, or if the results of the evaluation are indeterminate or unfavorable..then the next appropriate action is to further reduce pressure and temperature as required in Condition B.
The 72 hour Completion Time is based on engineering judgement and is reasonable to accomplish the activities necessary. For a slid violation the evaluation sho;:1d be possible within this time. More severe violations may require special, event specific stress analyses and/or inspections, which are appro-priately carried out while the RC5 is in a reduced pressure and temperature condition as required by Condition B. (' B.1. B.2 l- If the Required Actions and associated Completion Times are not L . met, a controlled shutdown must be initiated. This is a ! prudent action when the RCS remained in an unacceptable region for an extended period of increased stress or a sufficient l severe event caused entry into an unacceptable region. Either. Possibility indicates a need for more careful examination of the event, which is best accomplished while the RCS is in a low pressure and temperature state. With the plant at reduced pressure conditions the possibility of propagation of undetec-ted flaws is reduced. The times allowed for a controlled-shutdown to MODE 4 are reasonable and avoid placing undue stress on plant operators or plant systems. l l l SURV[lLLANCE SR 3.4.6.1 l REQUIREMINTS l Verification that operation is within limits is an appropriate surveillance when RCS temperature and pressure conditions are undergoing planned changes. The time period of 30 minutes is based on engineering judgement. Since temperature rate of change limits are specified in hourly increments, a half hour tt ae period permits assessment and correction of minor deviations within a reasonable time. (continued) l ABWR B 3.4 23 5/31/89 l-Amendment 9 16A.7 23 l
t MM 23A6100AP.
- !Ihandard Plant h*
0; RCS Pressure / Temperature Limits. I B 3.4.6 BASES feontinued) , SURVEILLANCE SR 3.4.6.2 REQUIREMENTS (continued) A separate limit is used when the reactor is critical. Conse-quently, it is appropriate to verify that the RCS pressure and
^
temperature are within the appropriate limit prior to the withdrawal of control rods that might make the reactor critical. SR 3.4.6.3 A differential temperature within the limit of this survell-lance will ensure that thermal stresses resulting from an idle l reactor internal pump startup will not exceed design allow- ! ances. Performing the surveillance within 15 minutes before starting the idle pump provides adequate hssurance that the limits will not be exceeded between the time of the surveil-lance and the time of the idle pump start. As noted, this surveillance is only required with less than five pumps already operating. With at least half of the pumps in~ operation, there is already sufficient mixing that this verification is not necessary. SR 3.4.6.4 Limits on the reactor vessel flange and head flange temperature (required wh6n the vessel head is tensioned) are generally bounded by the other PT limits during system heatup and cool-down. However, during operation in MODE 4, with RCS tempera-ture less than or equal to 100*F surveillance of the fit.rge temperatures is required to ensure the 70'F temperature limit is not violated. With RCS temperature less than or equal to 80'F, a more frequent check of the flange temperatures is required because of the reduced margin to the limit. The flange temperatures must also be verified to be above the limit prior to and during tensioning of the vessel head bolting studs to ensure that once the head is tensioned the limit is satisfied. Surveillance Frecuencies In general, surveillance frequencies are based o industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test ant a likelihood of a change in the system / component status. (continued) ABWR B 3.4-24 5/31/89 O Ameadment 9 16A.7 24
. l MN 2M6100AP -
Standard Plant m. a ;
/t \ =\
i RCS Pressure / Temperature Limits 8 3.4.6 Basts feontinued) REFERENCES 1. Code of Federal Regulations, Title 10. Part 50, Appendix 6 ' Fracture Toughness Requirements.'
- 2. American Society of Mechanical Engine 6rs (ASME), Boiler and Pressure Vessel Code, Section ill, Appendix G.
' Protection Against Non Ductile Fatiure.'
- 3. USNRC Regulatory Guide 1.99, Revision 1, ' Effects of Residual Elements on Predicated Radiation Damage to Reactor Vessel Materials,' April 1977.
- 4. USNRC Regulatory Guide 1.99, Revision 2, ' Radiation Embrittlement of Reactor Vessel Materials," May 1988.
- 5. American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code, Section 11, Appendix E,
' Evaluation of Unanticipated Operating Events.'
- 6. 52FR3788, Interim Policy Statement on Technical Specification Improvements for Nuclear Power Reactors.
USNRC, February 6, 1987.
- 7. NED0 31466, " Technical Specification Screening Criteria
( Application and Risk Assessment," November 1987. A8WR B 3.4 25 5/31/89
/ \
Amendment 9 16A.7 25 a e
-M M .. . 2M6100AP Standard Plant ,_ Rw A 164.7.7 B3.4.7 Reactor Steam ' Jose Perssort ' .i Reactor Steam Dome Pressure B 3.4.7 B 3.4 REACTOR COOLANT SYSTEM B 3.4.7 kaactor steam name Pressure BASES BACKGROUND The reactor steam done pressure is an assumed initial condition of design bests accidents and transients and is also assumed in the determination of compliance with reactor pressure vessel overpressure protection criteria. APPLICABLE The reactor steam dome pressure is an initial condition of the i SAFETY vessel overpressure protection analysis of Reference 1. This ANALYSES analysis assumes an initial maximum reactor steam dome pressure and evaluates the response of the pressure relief system (primarily the safety / relief valves) during the limiting pressurization transient. The determination of compliance with the overpressure criteria is dependent on the initial reactor steam dome pressure and therefore, the limit on this pressure ensures that the assumptions of the overpressure protection i analysis are conserved. Reference 2 also assumes an initial reactor steam dome pressure for the analysis of design basis accidents and transients used to determine the limits for fuel cladding integrity (MINIMUM CRITICAL POWER RATIO, see Bases for - LCO 3.2.2) and 1% cladding plastic strain (see Bases for LCO 3.2.1). Reactor Steam Dome Pressure satisfies the requirements of Selection Criterion 2 of the NRC Interim Policy Statement on Technical Specification leprovements as documented in Reference 3. LCO The specified reactor steam dome pressure limit assures the plant is operated within the assumptions of the transient analyses. Operation above the limit may result in a transient response more severe than analyzed. APPLICABILITY The reactor steam done pressure is required to be less than or equal to the limit in MODES I and 2 where the reactor is generating significant steam and the design basis transients and accidents are bounding. The limit may be exceeded during anticipated transients since the evaluations of References 1 and 2 demonstrate that appropriate reactor and fuel limits are not exceeded. (continued) ABWR B 3.4-26 5/31/89 O Amendment 9 16A.7 26
ABM . 234 i ase
. Standard Plant n... A j' .
L) Reactor Steam Dome Pressure B 3.4.7 BASES fcontinued) APPLICABILITY The limit is not applicable in MODES 3. 4 and 6, because in (Continued) these modes the reactor is shutdown. The reactor pressure is well below the required limit and no anticipated events will challenge the overpressure limits. ACTIONS A.I. B.1 If the reactor steam dome pressure is greater than the limit, prompt action should be taken to reduce the pressure to below the limit. If the operator is unable to reduce the reactor steam dome pressure to the limit, then the reactor is required to be in MODE 3. , Comoletion Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action.
.h.
D SURVEILLANCE REQUIREMENTS SR 3.4.7.1 The reactor steam dome pressure'is verified to be less than or equal to the limit every 12 hours. This ensures that the initial conditions of the design basis accidents and transients are met. Operating experience has demonstrated that the 12 hour frequency for this surveillance is adequate. REFERENCES 1. ABWR SSAR, Section 5.2.2.
- 2. ABWR $$AR, Section 15,
- 3. NEDO.31466, ' Technical Specification Screening Criteria Application and Risk Assessment.' November 1987.
ABWR B 3.4 27 5/31/B9 Amendment 9 16A.7 27
23A6100AP '
.Se==dard Plant a ,
SECTION 16A.8 b Section CONTENTS 31 Lit East 16AA.1 R16.51 ECCS Operatlag 16AA1 16A.8.2 R33.2 ECCS. Shutdown 16A.810 O , 1 0 16A.8-ii Amendment 9
ABWR z w ioo^P Reandard Plant a-a 16A.8 B3.5 Emergency Core Cooling Systems 16A.S.1 R3J.1 ECCS Operation l ECCS Operating 8 3.5.1 B 3.5 EMERGENCY CORE COOLING SYSTEMS B 3.5.1 tecs - emaratine ; BASES BAttGROUND The ECC5 is designed, in conjunction with the primary and secondary containment, to limit the release of radioactive materials to the environment following a loss of coolant < accident (LOCA). The ECCS is supported by other systems which provide automatic ECCS initiation signals (LCO ((CC5 Actuation Instrumentation)),coolingwatertocooltheECC$pumpsandthe rooms containing ECCS equipment (LC0 3.7.1), and electrical power LCOs 3.8.1 and 3.8.4). The ECC5 injection network is compris(ed of the High Pressure Core Flooder (HPCF) system, the
- Reactor Core Isolation Cooling (RCIC) system and the Low Pressure Flooder (LPFL) mode of the Residual Heat Removal (RHR) system. The ECCS also consists of the Automatic Depressuriza-tion System (ADS). The Condensate Storage Pool (CSP) and suppression pool provide the sources of water for the ECCS although no credit is taken in the safety analyses for the C5P.
(' The ECCS injection systems are arranged in three separate divisions each comprised of a high pressure and low pressure system. ECCS Division I consists of the RCIC system, and LPFL subsystem 'A'. ECCS Division 2 consists of LPFL subsystem 'B' and HPCF st.tsystem 'B'. ECCS Division 3 consists of the HPCF subsystem 'C' and the LPFL subsystem 'C'. LPFL is an independent operating mode of the RHR system, in the LPFL mode, the RHR operates to provide the ECCS function. There are three LPFL subsystems. Each LPFL subsystem (Ref. 2) consists of a rotor driven pump, piping and valves to transfer water from the suppression pool to the reactor vessel. Each LPFL subsystem has its own suction and discharge piping. The water is injected into the reactor vessel outside the core shroud, via feedwater line 3 for LPTL subsystem A, and via the individual LPFL subsystem inlet nozzles for LPFL rubsystems B and C. The LPFL subsystems are designed to provice core cooling at low reactor vessel pressure. Upon receipt of an initiation signal, each LPFL pump is automatically started (from normal AC power if available; otherwise, the pumps start after emergency AC power becomes available). When the reactor vessel pressure drops sufficiently, LPFL flow to the reactor vessel begins. RHR system valves in the LPFL flow path are automatically positioned to ensure the proper flow path for water from the suppression pool, through the RHR heat exchanger, to injection into the reactor vessel. RHR system valves that service other RHR functions (e.g. shutdown cooling, fuel pool cooling) must be manually realigned when required (continued) ABWR B 3.5-1 5/31/89 O Amendment 9 16A.61
I 21A6100AP -
; Stawlard Plant - m. a ;
- I
[CCS Operating B 3.5.1 ) BASES (continued) : BACKGROUND following LPFL subsystem operation. A discharge test line is (continued) provided to route water from and to the suppression pool to allow full flow testing of each LPFL' pump without injecting ' water into the reactor vessel. The HPCF system is comprised of two HPCF subsystems. A HPCF l subsystem (Ref. 3) consists of a single motor driven pump, a j flooding sparger in the upper plenum of the core shroud, and i piping and valves to transfer water from the suction source to l the sparger. Suction piping is provided from the CSP and the suppression pool. Pump suction is normally aligned to the CSP l source to minimite injection of suppression pool water, which is not likely to meet boiling water reactor (BWR) water quality requirements, into the reactor pressure vessel (RPV). However, i if the CSP water supply is low or the suppressicn pool level is I above a prescribed level, an automatic transfer to the suppres. l sion pool water source assures a water supply for continuous ; operation of the HPCF system. The HPCF system is designed to ! provide core cooling over a wide range of reactor vessel pressures (0 to ll7? psid, vessel to suppression pool air- .i space). Upon receipt of an initiation signal, the HPCF pump i automatically starts (from normal AC power if available, j otherwise, the pump starts after emergency AC power becomes ! available) and valves in the flow path begin to open. Since the HPCF system is designed to oprate over the full range of expected reactor vessel pressures, HPCF flow begins as soon as the necessary valves are open. A full flow test line is r provided to route water from and to the suppression pool to < allow testing of each HPCF subsystem during normal operation without injecting water in the reactor vessel. ] The RCIC system (Ref. 1) consists of a steam driven turbine- 1 pump unit, piping and valves to transfer water from the suction r source to the core via feedwater line A. Suction piping is provided from the CSP and the suppression pool. Pump suction l is normally aligned to the CSP to minimize injection of suppression pool water into the RPV. However, if the CSP water supply is low or the suppression pool level is above a pre-scribed level, an automatic transfer to the suppression pool assures a water supply for continuous operation of the RCIC l system. The steam supply to the turbine is piped from main l steam line B upstream of the inboard main steam line isolation l valve. The RCIC system is designed to provide core cooling , over a wide range of reactor pressures. RCIC is also designed I to operate following RPV isolation accompanied by a loss of coolant flow from the feedwater system to provide adequate core j cooling and control of reactor vessel water level. i (continued) ABWR B 3.5 2 5/31/89 9 Amendment 9 16A.8-2 1
) + - . _ -
MM _ 23A6100AP mndard Plant' mal p a ECCS Operating ' B 3.5.1 - BASES (continued) BACKGROUND . All ECCS pumps are provided with minimum flow bypass lines , (continued) which discharge to the suppression pool. The valves in these these lines automatically open to prevent pump damage due to overheating t+en other discharge line valves are closed er reactor vessel pressure is greater than the LPFL pump discharge pressure tc110 wing system initiation. To ensure rapid delivery of water to the reactor vessel and to ministre waterhammer effects, the ECCS discharge line keep fill systems are designed to maintain all pump discharge lines filled with water. The ADS (Ref. 4) consists of 8 of the 18 safety / relief valves ($RVs). It is designed to provide depressurization of the primary system during a small break LOCA if HPCF and RCIC fail to maintain required water level in the reactor vessel. ADS operation reduces the reactor vessel pressure to within the operating pressure range of the low pressure LPFL subsystems, so these subsystems can provide core cooling. Each of the SRVs used for automatic depressurization is equipped with a separate dedicated pneumatic accumulator and associated inlet check valve. The safety related accumulators provide the pneumatic power to actuate the valves. Nitrogen is normally supplied for operatten of the ADS valves from the non safety related
- O4 Atmospheric Control System.
t APPLICABLE The ECCS performance is evaluated for the entire spectrum of SAFETY break sizes for a postulated LOCA. The accidents for which ANALYSES ECCS operation is required are specifically listed in SSAR Chapter 15. The required analyses and assumptions are defined in Reference 5. The results of these analyses are described in Reference 6. . The ECCS subsystem design requirements ensure the criteria of Reference 7 are satisfied under all postulated LOCA conditions j; assuming the worst single active component failure in the ECCS. The limiting single failures are discussed in Reference 9. For any LOCA, failure of ECCS subsystems in Division 2 or 3 due to failure of its associated diesel generator is the most severe single failure. One ADS valve failure is analyzed as a limit-ing single failure for events requiring ADS operation. The remaining OPERABLE ECCS subsystems provide the capability to adequately cool the core and prevent excessive fuel damage. ECCS Operating satisfies the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 10.
)
'(continued)
ABWR B 3.5 3 5/31/89 I O Q Amendment 9 16A.8-3
.M M Riandard Plant 21A6100AP .
aa .
-1 O
ECCS . Operating 8 3.5.1 i BASES (continued) , LCOs The ECCS injection subsystems are defined as the three LPFL subsystems, the RCIC system and the two HPCF subsystems. The 7 divisional arrangement of the [CCS subsystems and a description of what is required for the [CCS to be considered OPERABLE is provided in the Background Section. With less than the required number of ECCS subsystems OPERABLE, the potential exists that during a limiting design basis LOCA
, concurrent with the worst single failure, the limits specified in Reference 7 could be escoeded. All ECCS subsystems must therefore be OPERABLE to satisfy the single failure criterion required by Reference 7.
A LPFL subsystem may be considered OPERABLE during alignment to and operation in the RHR shutdown cooling mode when below the - RHR cut in remissive pressure in MODE 3, if capable of being manually realigned to the LPFL mode and not otherwise inoper-l able. At these low pressures and decay heat levels (reactor is shutdown in MODE 3) a reduced complement of ECCS subsystems can provide the required core ccoling thereby allowing operation of an RHR shutdown cooling loop when necessary. 1 APPLICABILITY All ECCS subsystems are required to be OPERABLE during MODES 1, O r 2 and 3 when there is considerable energy in the reactor core and core cooling would be required to prevent fuel damage in the event of a break in the primary system piping, in MODES 2 and 3 the ADS function is not required when pressure is s 50-are psig because capable the low pressure of providing flow into ECCS subsystems the reactor vessel (LPFL) be low this pressure. ECCS requirements for these conditions are specified in LCO 3.5.2. The RCIC system is required to be OPERABLE in MODES 1, 2 and 3 with reactor steam dome pressure 2 150 psig since RCIC is a steam driven water source which provides water for core cooling when the reactor is isolated and pressurized. In MODES 2 and 3 with reactor steam done pressure < 150 psig, and in MODES ( and
- 5. RCIC is not required to be OPERABLE since the other ECCS subsystems can provide sufficient flow to the vessel.
(continued) ABWR B 3.5 4 5/31/89 0
. Amendment 9 16A.8-4
8 N 2M6100AP ma Standard Plant ~ i - ECCS Operating B 3.5.1 BASES (continued) p ACTIONS A.I. B.1. B 2. C.1. C.2. C.3 With three or less ECCS injection subsystems from Divisions 1.
- 2. and 3 inoperable, the remaining OPERABLE subsystems provide adequate core cooling during a LOCA. However, the overall ECCS reliability is reduced because a single failure in one of the remaining subsystems concurrent with a LOCA may result in the ECCS not being able to perform its intended safety function.
Therefore, continued eperation is only allowed for a limited , time. D.1. D.2 Should the Required Actions and associated Completion Times of Conditions A. B. or C not be met. the reactor is required to be in MODE 3 and subsequently in MODE 4 Of course for the specific case where RClf is the only inoperable subsystem, once reactor pressure is reduced below 150 psig, the LCO is no
' longer applicable and no further ACTION is required. In MODE l
- 4. the ECCS requirements are specified in LCO 3.5.2. If unable to attain MODE 4. the reactor coolant temperature should be maintained as low as practical by use of alternate heat removal methods.
3 E.1. F.1. F.2. C.). G.2 The'LCO requires B ADS valves to be OPERABLE to provide the ADS function. Analysis demonstrates that only three ADS valves are required for the ADS to successfully perfom its function. With one or two ADS valves inoperable, the reliability of the ADS is relatively unaffected. Operation is allowed until the next outage of 30 days initial planned length, or until the refueling outage, whichever comes first. As noted, the provisions of LCO 3.0.4 are not appitcable for this Condition to allow recovery from shutdowns of short duration prior to the
- l. next refueling outage and/or extended maintenance outage. When three ADS valves are inoperable, operation is allowed to continue for 30 days to allow time to attempt repair and/or to plan for a plant outage to affect repairs. Three ADS valves out of service is considered indicative of a generic problem, and continued plant operation beyond the 30 days is not allowed. If the required number of inoperable ADS valves cannot be made OPEkABLE. the plant is required to be in MODE 3 and the reactor pressure reduced to s 50 psig. At these conditions the ADS function is no longer required since the lJ reactor pressure is low enough :;uch that the low pressure ECCS subsystems can perform their designed safety function.
j- (continued) ABWR B 3.5-5 5/31/89 l l Amendment 9 16A.8 5 l
i s ABWR =>= ;
. Standard Plant " ,
O i ECCS - Operettop B 3.5.1 BASES (continued) AC110NS These action statements assume that the valve failure is in an (continued) inaccessable area of the plant. Since the intent of this LCO is to have all ADS valves operable at all times, all ADS valve failures which can be repaired in areas of the plant accessable during normal operation should be repaired within a ressoreble period of time, cameletion Times The ECCS Completion Times are based on the results of a study which evaluated the impact on ECCS unavailability assuming various components and subsystems were taken out of service. The results were used to calculate the average availability of ECCS equipment needed to mitigate the consequences of a LOCA as l a function of allowable outage times (A0T). A0Ts were then chosen to provide comparable levels of ECCS availability. SURVE!LLANCE SR 3.5.1.1 REQUIREMENTS The pump discharge lines of the RCIC system, and the LPFL and . HPCF subsystems are required to be kept full with water to - minimize potential waterhammer effects when the systems are initiated. Additionally, the lag between the receipt of the initiation signal and the actual injection into the reactor vessel is minimized. One acceptable method of ensuring the lines are " full" is to vent at the high points. SR 3.5.1.2 L Verification that all applicable valves are in the required position ensures proper flow paths for ECCS. However, a valve that is capable of automatic return to its ECCS position, when . an ECCS initiation signal is present, can be in position for another mode of operation. This is applicable only if the valve auto repositions and fully strokes withis. the time required for its ECCS function. For the RCIC system. this also includes the steam flow path for the RCIC turbine. This surveillance also includes the RCIC flow controller position. SR 3.5.1.3 The accumulator on each ADS valve provides pneumatic pressure for valve actuation. The designed pneumatic supply pressure requirements for the accumulator are such that following a failure of the pneumatic supply to the accumulator, at least one valve actuation can occur with the drywell at design (continued) ABWR B 3.5 6 5/31/89 O 16A &6
- Amendment 9
MM 2M6100AP Rev.A RennAnrd Plant (/ ! ECCS Operating B 3.5.1 BASES (continued) SURVElLLANCE pressurc, or five valve actuations with the drywell at REQUIREMENTS atmospheric pressure. The ECCS safety ar.alysis assumes only (continued) one actuation to achieve the depressurization required for operation of the low pressure ECCS and therefore this require. ment provides sufficient margin to satisfy the assumptions of the safety analyses. This minimum required pressure of 161 ' psig is provided by the Atmospheric Control System. 1R 3.5.1.4. 3.5.1.E. 3 E.1.6 The performance requirements of the ECCS pumps are determined through application of the 10CFR50 Appendix K criteria (Ref. 5). The pump flow rates, as deterwined by analysis, ensure that adequate core cooling is provided to satisfy the accept. , ance criteria of Reference 7. Periodic surveillance is per-formed (in accordance with ASME Section XI requirements) to verify these flow rates. The pump flow rates are verified against a system head that is equivalent to the reactor vessel pressure expected during a LOCA. The total system head devel-oped is adequate to overcome the elevation differsnces between the suction source and the vessel, friction losses and pressure differences present during LOCA. These values are established l during preoperational testing. The RCIC pump flow rates ensure that the system can maintain reactor coolant inventory during pressurized conditions with-the RPV isolated. The flow tests for the RCIC system are performed at two different pressure ranges such that system capability to provide rated flow is tested both at the higher l' and lower operating range of the system. Since the required reactor steam dome pressure must be available to perform SR 3.5.1.5 and SR 3.5.1.6, sufficient time is allowed after adequate pressure is achieved to perform these tests. 1 SR 3.5.1.7 l' The ECCS subsystems are required to actuate automatically to I ' perfore their designed function. These surveillance tests demonstrate that with the required system initiation signals, the automatic initiation logic of HPCF, RCIC and LPFL will cause them to operate as designed, including actuation of all automatic valves to their sequired position. This test also ensures that the RCIC system will automatically restart on a signal received reactor vessel subsequent low water to reactor level vessel (Level high water2) level (Level 8) trip. This test also ensures that the HPCF subsystem and RCIC system will automatically transfer suction from the CSP to the suppression pool. Since all active components are testable and (continued) B 3.5 7 5/31/89 Q ABWR V 16A.8-7 Amendment 9 I f l
l I i 21A6100AP
. Standard Plant w i
=
l L ] ECts Operating B 3.6.1 i BASES (continued) l j
$URVE!LLANCE full flow can be demonstrated by recirculation through the test REQUIREMENT $ line, coolant injection into the reactor vessel is not required ;
(continued) during the tests. -i j; MLa-Q The ADS designated SRVs are required to actuate automatically upon receipt of specific initiation signals. A system func. tional test (logic only) is perfomed to demonstrate that the }. ADS logic operates as designed when initiated, causing proper actuation of the required components. Actual ADS valve actu-ation is excluded to prevent a reactor pressure vessel , l blowdown. I 1R 3.5.1.9 L A manual actuation of ADS valves is performed to verify that I the valve and solenoids are functioning properly and that no blockage exists in the $RV discharge lines. This is demon-strated by actual valve stem movement (as indicated by LVDT) or by any suitable method to verify steam flow. Adequate reactor steam done pressure must be evallable to I perform this test to avoid damaging the valve. Sufficient time is therefore allowed, after the required pressure is achieved, to perform this test. Adequate pressure at which this test is ! to be performed is ] psig (the pressure recommended by the valvemanufacturer)[. Reactor startup is allowed prior to , performing this test because valve OPERABILITY and the set-l points for overpressure protection are verified, per ASME ' requirements, prior to valve installation. l Surveillance Fraouencies In general, surveillance frequencies are based on industry - accepted practice and engineering judgement considering the . unit conditions required to perform the test, the ease of I performing the test and a likelihood of a change in the ' system / component status. (continued) ABWR B 3.5 8 5/31/89 l. l Amendment 9 16A.8 8 i i
- d. '
M 23A6100AP ; Remndard Plant ui. L- 3 1 i ECCS . Operating B 3.5.1 BASES (continued) REFERENCl3 1. ABWR $$AR, Section 6.3.2.2.3.
- 2. ABWR $$Ak, Section 6.3.2.2.4.
- 3. ABWR $$AR, Section 6.3.2.2.1.
- 4. ABWR $$AR, Section 6.3.2.2.2.
- 5. 10CFR$0, Appendix K, 'ECCS Evaluation Models'.
- 6. ABWR SSAR, Section 6.3.3.
- 7. 10CFR$0.46, ' Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors'.
- 8. 52FR3788 Proposed Policy Statement on Technical Specification leprovements for Nuclear Power ReactorJ, February 6, 1987.
- 9. ABWR SSAR, Section 6.3.3.3.
p
- 10. ' Technical Specification Screening Criteria Application and Risk Assessment', NED0 31466, November 1987, i
ABWR 8 3.5 9 5/31/B9 :
.O l Amendment 9 16A.89
- ~ - - - . _ _ _ _ _ - - - _ - - - - _ _ _ _ - - - - - -
MN 21A6100AP
. Standard Plant u.a ,
i 164.8.2 R3.5.2 ECCS Shutdown ECCS - Shutdown B 3.5.2
,i B 3.5 EMERGENCY CORE COOLING SYSTEMS E 3.5.2 Ects - Shutdown BA$ts BACKGROUND and the f AdescriptionoftheHighPressureCoreFlooder(HPCF)
Low Pressure Flooder (LPFL) subsystems of the ResidualHeat Removal (RHR) system are provided in the Bases for LCO 3.5.1. APPLICABLE The ECCS performance is evaluated for the entire spectrum of SAFETY break sizes for a postulated Loss of Coolant Accident (LOCA). ,
' ANALYSES The lon term cooling analysis following a design basis LOCA-(Ref.I demonstrates that only one ECCS subsystem is required, >
post LO A, to maintain the peak cladding temperature below the allowable limit. To preserve the single failure criterion of L Reference 2, a minimum of two ECCS subsystems are required to be OPERABLE in MODES 4 and 5. Two OPERABLE ECCS subsystems also ensure adequate vessel inventory makeup in the event of an inadvertent vessel draindown. The consequences of an inadvert-ent draindown of the vessel, which may require ECCS subsystem operation during MODES 4 and 5, are bounded by these analyses. ECCS Shutdown satisfies the requirements of Selection Crt. terion 3 of the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 3. LCOs The ECCS injection subsystems are defined for the purposes of this specification as the three LPFL subsystems and the two HP.CF subsystems. Each LPFL subsystem consists of one motor driven pump, piping and valves to transfer water from the suppression pool to the reactor vessel. Each HPCF system consists of one motor driven pump, piping and valves to trans-fer water from the suppression pool or Condensate Storage Pool (CSP) to the reactor vessel. Any LPFL subsystem that may be aligned in the shutdown cooling mode of the RHR system in MODES 4 or 5, is considered OPERABLE for the ECCS function, if it can be manually realigned remotely i or manually to the LPFL mode and is not otherwise inoperable. l Because of low pressure and temperature conditions in MODES 4 and 5, sufficient time will be available to manually align and l initiate LPFL subsystem operation to provide core cooling prior j to postulated fuel uncovery. (continued) ABWR B 3.510 5/31/89 O Anwndment 9 1 9 .8-10
k MM 21A6100AP
- m. a
, - Standaed Plant -
ECCS - Shutdown B 3.5.2 BASES (continued) APPLICABILITY ECCS OPERABILITY is required in MODES 4 and 5 to assure adequate coolant inventory and sufficient heat removal capability for the irradiated fuel in the core in case of an inadvertent draindown of the vessel. Requirements for ECCS OPERABILITY during MODES 1. I and 3 are discussed in the Applicability section for LCO 3.5.1. ECCS subsystems are not required to be OPERABLE during M00E 5 with the spent fuel pool gate removed and the water level siaintained greater than or equal to 23' above the reactor pressure vessel flange. This provides sufficient coolant inventory to allow operator action to terminate the inventory loss prior to fuel uncovery in case of an inadvertent draindown. The Automatic Depressurization System (ADS) is not required to be OPERABLE during MODES 4 and 5 because the reactor vessel pressure is < 50 psig and LPFL and HPCF subsystems can provide core cooling without any depressurization of ths primary system being required. Since the Reactor Core isolation Cooling (RCIC) system requires steam to operate it is not required to be OPERABLE during MODES 4 and 5. , v ACTIONS A.1. B.1 a 1 With one of the two required ECCS subsystems inoperable, the remaining OPERABLE subsystem can provide sufficient vessel flooding capability to recover from an inadvertent vessel ; draindown. However, system reliability is reduced because a j single failure in the remaining subsystem concurrent with a vessel draindown could result in the ECCS not being able to I perform its intended function. Therefore, continued operation is only allowed for a limited time. With the inoperable I subsystem not restored to OPERABLE status within the required Completion Time, operations that have.the potential for drain, ing the reactor vessel must be suspended. This minimizes the 1 probability of a vessel draindown and the subsequent potential 1 for ECCS actuation, j l (continued) l . ABWR B 3.5-11 $/31/89 l Amendment 9 16A.B-11
ABM Asioo^r 2 Standard Plant mu
=
O 4 ECCS Shutdown 8 3.5.2 . . t
-BASES (continued) i
'ACil0NS C.1. C.2. D.I. D.2. D.3. D.4 (continued) r With both of the required ECCS subsystems inoperable, all J coolant inventory makeup capability may be unavailable and operations that have a potential for draining the reactor t
). vessel must bt suspended. If at least one ECC5 subsystem is not restored to OPERABLE status within the required Completion Times, additional actions are required to minimize any poten-tial release of radioactive materials to the environment. This includes ensuring Secondary Containment is OPERABLE, at least subsystem is OPERABLE one and atStandby least oneGas Treatment Secondary System (SGTS) Isolation Valve and Containment ' associated actuation instrumentation is OPEEABLE in each .- associated penetration not isolated. This may be performed by an administrative check, by examining logs or other inform-
. ation, to determine if the com
,1 maintenance or other reasons. ponents It does not aremean out oftoservice performfor the i surveillances needed to demonstrate OPERABILITY of the compon-ents. If however, any required component is inoperable, it
- s.ust be restored to OPERABLE status. In this case, survell-lance requirements may need to be performed to restore the
, component to OPERABLE status. j Cameletion Times - All Completion Times are based on industry accepted practice and engineering , judgement considering the number of available systems and the time required to reasonably complete the Required Action.
- i
! -. SURVE!LLANCE 1,R3.5.2.1 1 -. REQUIREMENTS l The minimum water level 14.63 feet required for the suppression pool is verified to ensure that the suppression pool will provide adequate net positive suction head (NPSH) for the RHR pumps, including consideration of a recirculation volume, vortex prevention and a safety margin for conservatism. The suppression pool water level is referenced to the inside bottom l< ef the suppression pool. With the suppression pool water level less than the required limit, all LPFL subsystems are inoperable, j (continued) i n ABWR B 3.5.]2 $/31/89
. Amendment 9 16A&l2
- _ _ , . ~ . _ -.
= ,
M' 1 M
~
23A6100AP _Riandard Plant - m.a l
-(
I: i'
'l ECC$ Shutdown j- 8 3.5.2 I BASES (continued)
$URVEILLANCE 1R 3.5.2.2 REQUIREMENTS 1 When suppression pool level is less than 14.63 feet. HPCF is (continued) considered OPERAtLE only if it can take suction from the CSP and the CSP water level is sufficient to provide the required NPSH for the HPCF pump. Therefore, a verification that either the suppression pool water level is t 14.63 feet or HPCF is aligned to take suction from the CSP, and the CSP contains l 285,264 gallons of water, equivalent to !!.8 feet water level, '
ensures HPCF can supply makeup water to the reactor vessel. Suppression pool water level is referenced to the inside bottom of the suppression pool. SR 3.5.2.3. 3.5.2.5. 3.5.2.6 The bases provided for $R 3.5.1.1, 3.5.1.4, and 3.5.1.7 are applicable. ; 1R 3.5.2.4 Verification that all applicable valves are in the required position ensures proper flow paths for ECCS. However, a valve s that is capable of automatic return to its ECCS position, when b an ECCS initiation signal is present, can be in position for another mode of operation, in MODES 4 or 5 the RHR system may operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor. Therefore, during MODES 4 or 5 RHR valves that are required for LPFL subsystem operation may be aligned for the shutdown cooling mode. The LPFL mode of the RHR however, may ' be considered OPERABLE for the ECCS function if all the required valves in the LPFL flow path can be manually realigned (remote or local) to allom injection into the RPV and the system is not otherwise inoperable. This will ensure adequate core cooling if an inadvertent vessel draindown should occur. Surveillance Frecuencies In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the system / component status. (continued) ABWR B 3.5-13 5/31/89 O Amendment 9 16A.813
IMM . Standard Plant
' 23A6100AP
- m. A l
G: ECCS - Shutdown 8 3.5.2 BASES (continued) REFER (NCES 1. ABWR 55AR, Section 6.3.
- 2. 10CFR50.46, ' Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors
- 3. ' Technical Specification Screening Criteria Application and Risk Assessment *, NEDO 31466 November 1967.
O t ABWR B 3.514 5/31/89 O Amendment 9 16A.8-14
i MN - nA61ooAP. '! . . .m Se==dard Plant w , SECTION 16A.9 h.-p CONTENTS ; Section Dgle Egg 16A.9.1 R34.1 Cantainament Svatens 16A.91 16A.9.1.1 B3.6.1.1 Primary Containment 16A.9-1 16A.9.1.2 B3.6.1.2 Containment Air Locks 16A.9 5
+
16A.9.13 B3.6.13 Primary Containment Pressure 16A.911 16A.9.1.4 B3.6.1.4 Drywell Average Air Temperature 16A.913 16A.9.1.5 B3.6.1.5 Primary Containment and Pressure Isolation Valves 16A.915 16A.9.1.6 Wetwell to-Drywell Vacumm Breakers 16A.9 21 16A.9.2 B3.6.2 Suppression Pool 16A.9 16A.9.2.1 B3.6.2.1 Suppression Pool Average Temperature 16A.9 26 16A.9.2.2 B3.6.2.2 Suppression Pool Water 1.cVel l'6A.9-31
/
k ' 16A.9.23 B3.6.23 Residual Heat Removal Suppression Pool Cooling 16A.9-34 16A.9.2.4 B3.6.2.4 kesidual Heat Removal Wetwell Spray System 16A.9-38 16A.93 B3.6.3 Hvdronen Control 16A.9-41 16A.93.1 B3.63.1 Hydrogen Recombiner System: 16A.9-41 16A.93.2 B3.63.2 Primary Containment Oxygen Concentration 16A.9 42 16A.9.4 B3.6.4 E,enadary cantninment Svstems 16A.9-45 16A.9.4.1 B3.6.4.1 Secondary Containment 16A.9-45 16A.9.4.2 B3.6.4.2 Secondary Containment Isolation Valves 16A.9 51 16A.9.43 B3.6.43 Standby Gas Treatment System 16A.9 55 16A.9-ii Amendment 9
.1 ABWR -
S w laarr Remndard Plant _ m- i 16A.9 B3.6 Containment Systems 16A.9.1 B3.6.1 Primary Containment Systeams 1 16A.9.1.1 B3.6.1.1 Primary Costalmament Primary Containment S 3.6.1.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.1 Primary containment BAsrs BACKGROUND The function of the Primary Containment is to isolate and j contain fission products released from the reactor primary system in the event of a Loss of Coolant (LOCA). The Primary Containment consists of a reinforced concrete containment vessel with a steel liner which provides a leak tight barrier surrounding the reactor primary system and includes the upper and lower drywell and the wetwell consisting of the suppression pool and wetwell airspace. To assure Primary Containment OPERABLE, leakage test requirements have been set forth by Reference 1. These test l requirements provide for periodic verification by tests of the I leak tight integrity of the primary containment and systems and I components which penetrate the primary containment. The purpose of the leak tests are to assure that leakage through the primary containment and systems and components penetrating
..the primary containments shall not exceed the allowable leakage rates specified in the technical specification and used in the safety analyses. Additionally the periodic tests performed y assure that proper maintenance and repairs are made during the i A.- --
service life of the plant. All leakage rate requirements and surveillance requirements are in conformance with 10 CFR 50 Appendix J (Ref. 1) and approved exemptions. Periodic Type A tests can use the peak calculated containment pressure of 45.0 psig (Ref. 2) or a reduced j pressure of Pt [ ]psig. l The maximum allowable leakage rate for the primary containment t is 0.5 percent by weight of the containment air per 24 hours at Pa,45.0psigor[ ]percentbyweightofthecontainment air per 24 hours at the reduced pressure of Pt [ ] psig (Ref. 2). The maximum allowable leakage rate is based on what is acceptable for nuclear safety considerations per 10 CFR 100. ' Reactor size, site location and m6teorology, as well as the possible mechanisms for radioactivity generation and transport within the containment vessel are all considered in specifying the allowable leakage rate for a given containment system. Generally the lowest Isakage rate which is tsadily obtainable and measurable is specified even though this may be lower than required by safety considerations. l (continued) i 1 ABWR B 3.6 1 5/31/89 j l O i l Amendment 9 16A.91
- l. J l
l
5MM
. Standard Plant.
. 33A6100AP
'" 4
.t
. 0..
Primary Containment-t 3.6.1.1 : RASIS feentinued) BACKGROUD The maximum allowable leakage rate is used as an input to the ' (continued) safety analysis of Reference 3 which confirms that the specified maximum allowable leakage rate will conform to the ' requirements of 10 CFR 100. To maintain primary containment integrity the drywell bypass' , leakage must be ministred to prevent overpressurization of the wetwell during the drywell pressurtaation phase of a LOCA. This requires periodic testing of the drywell bypass leakage and confirmation that the wetwell to drywell vacuum breakers are closed per LCO 3.6.1.6. This specification assures that the performance of the primary containment in the event of a LOCA meets the requirements of Reference 4 and the assumptions used in the safety analyses of References 2 and 3. L
,.4 APPLICABLE Analytical methods and assumptions involving the primary SAFETY containment are presented in Reference 2 and 3. The safety ANALYSES- analyses assume a non mechanistic fission product release following a LOCA which forms the basis for determination of .
offsite doses. The fission product release is in turn based on an assumed leak rate from the primary containment. OPERABILITY of the Primary Containment assures that the leak rate assumed in the safety analyses is not exceeded and that the site boundary radiation dose will not exceed the limits of 10 CFR 100 even if the non mechanistic release were to occur. Primary Containment satisfies the requirements of Selectici Criterion 3 of the NRC Interim Policy Statement on Tochtical Specification Improvements as documented in Reference 5. LCO The Primary Containment must be OPERABLE to assure that the containment conditions, passive features, and active features are consistent with those assumed in the safety ar.alyses, in order for the Primary Containment to be considered OPERABLE. only those SRs listed in this LCO must be rat. APPLICABILITY Maintenance of Primary Containment OPERABILITY is appitcable during MODES 1, 2 and 3 since it is rwired when sufficient energy is contained in the reactor coolant system to pressurtre the containment. Primary Containment is required and is assumed in the LOCA analyses of Reference 2, where pressurtration of the primary containment occurs. A LOCA in (continued) ABWR B 3.6 2 5/31/89 O 16A.9-2 Amendment 9
. . - - - - - . . -- -. - ~ - .
21A6100AP - fittaendard Plant m ; Primary Containment , B 3.6.1.1. RASES fenntinuedi 8 r APPLICABILITY MODE 4 or 5 will not pressurire the primary containment and (continued) therefore its OPERABILITY is not required. The analyses of Reference 3 which assumes and requires primary conteineent also assumes the reactor is pressurtred (i.e., M00L 1, 2 or 3). ACTIONS A.1. B.* and B.2 With Primary Containment inoperable the leak rate may exceed , that assumed in the design basis accident analyses. A short ' time is allowed to restore Primary Containment to OPERABLE
- status due to the low probability of an event which would pressurtre the primary containment. However, if Primary Containment cannot be restored to OPERABLE status, the reactor is required to be in MODE 3 and subsequently in MODE 4.
comolation Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonsbly complete the Required Action, n SURVEILLANCE 1R 3.6.1.1.1 REQUIREMENTS Maintaining the pressure suppression function of Primary Containment requirws limiting the leakage from the drywell to the wetwell. Thus, if an event were to occur which pressurized the drywell, the steam would be directed through the downcomers into the suppression pool. This SR measures drywell to metwell differential pressure over a ten minute period to ensure that the leakage paths which would bypass the suppression pool are within allowable limits. SR 3.6.1.1.2. SR 3.6.1.1.3 Maintaining Primary Containment OPERABLE requires compliance l with the leak test requirements of 10 CFR 50, Appendix J. ' Therefore, these two SRs reflect the leak rate testing requirements with regard to overall containment leakage (Type A and B leak tests). Other specific Appendix J requirements are addressed in the individual containment component Surveillance Requirements. Type A and B testing must be performed in accordance with 10 CFR 50, Appendix J or NRC approved exempt-ions to Appendix J. These periodic testing requirements verify that the containment leak rate does not exceed the leak rate assumed in the safety analysis. The surveillance frequency is required by Appendix J, and as such, SR 3.0.2 (which allows survelliance frequency extensions) does not apply. (continued) ABWR B 3.6 3 5/31/89 t l Amendment 9 16A.9-3
1 1 MM ' 23A6100AP. Standard Plant' m-# 1 Prismary Containment B 3.6.1.1 BASES fcontinued) SURVEILLANCE Surveillance Freauencies REQUIREMENTS (continued) In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the systesVcomponent status. REFERENCES 1, 10 CFR 50, Appendix J. - l 2. A8WR SSAR, Section 6.2. ABWR SSAR, Section 15.6.5. i 3.
- 4. 10 CFR 50, Appendix A, GDC 16.
i 5. NED0 31466, Technical Specification Screening Criteria Application and Risk Assessment,' November 1987. l O ABWR B 3.6-4 5/31/89 O Amendment 9 16A.9-4
ABWR zwar ! Standard Plant m. a
.~ 164.9.1.2 33.6.1.2 Contalement Air Lacks 1
I I Contatnment Air Locks ! B 3.6.1.2 i B 3.6 CONTAIN4[NT $Y$1[MS j B 3.6.).2 f.ontainment Air Lacks kutt BACKGROUND Two double door Containment Air Locks have been built into the i containment which provide access while also assuring that the containment remains isolated during the process of personnel i entering and exiting the primary containment. [ach air lock is i designed with a seal system for each door in the air lock as l well as an interlock mechanism that assure one door is closed and sealed before the other door is opened. The air locks are t designed to withstand the same leads, temperatures and peak e design internal and external pressures as the containment (Ref. I
!). As part of the containment boundary, the air locks Itmit the release of redteactive material to the environs during ,
norsal plant operation and through a range of incidents up to and including (postulated) design basis accidents. , The containment. including containment air locks, is designed to maintain its functional integrity euring and following the peak transient pressures and temperatures that would occur following any postulated loss.cf.coolar.t accident. O The air lock doors have inflatable seals which are maintained at[ ]psigb maintained at ! y the seal airdoor flaskhas
/ pneumatic psig. Each two seals sosystem that which is theyaresinglefai)ureproofinmaintainingtheleaktight l
boundary of primary containment. APPLICABLE Primary Containment OPERABILITY. and the itelting of radio.
$AFETY active material release to the environs. is a consideration ANALY$[$ in the evaluation of a number of accident analyses presented in FSAR Chapters 6 and 15. The Containment Air Locks are a part vf the Primary Containment boundary. Their design and maintenance are essential to maintaining Primary Containmen*
OPERABILITY which assures that the safety analysis assumpti: " are maintained. l Containment Air Locks satisfy the requirements of Selection Criterion 3 of the NRC Interim Polt y Statement on Technical
$pectftcation Improvements as documented in Reference 3.
i . (continued) i l ABWR B 3.6 $ 6/31/89 O i ifA9-5 t Amendment 9
l_ I l 31461W t Standard Pimpt m.a , O! Containment Air Locks ! B 3.6.1.2 Mitt frontinued) LCO The Containment Air Locks are required to be OPEMBLE with at
- 1 east one door closed. For the air locks to be considered OPinABLE, the doer seals must not teak, the interlock mechanism must be OPthABLt. the air lock must be in compliance with the '
Type B sir lock tests, and both doors in each air lock must be OPERABLt. This reeutrement is necessary to maintain contain. ment integrity whet, a; cess in and out of the containment is required. APPLICABILITY The centainment air incks are required to be operable during NODE $ 1. I and 3 consistent with the requirement for when t Primary containment must be OPERABLE. In WIDES 4 and 5 there is insufficient energy in the reactor to pressurtte containment such that a containment air lock fatture would not threaten plant safety. , ACTION $ A.I. A.f. A.3 When one air lock door is inoperable, or che door and the interlock mechanism on the same air lock are inoperable, then the air lock will not perfors its intended function of allowing ; personriel access while maintaining the primary containment boundary intact. It will however, maintain the boundary intact , indefinitely if the remaining OPERABLt door in the air lock is closed. Therefore, for this condition, the OPERABLE door is closed ismediately. If the components cannot be restored to the OPERABLE door must be locked and a OPERABLE periodic status, lance performed to confirm it is locked. surveil The i provisions of LC0 3.0.4 are not applicable and continued plant l operation is justified since there is another contatraent air lock which is OPERABit and because the primary containment boundary is maintained intact. (continued) ) i 1 ABWR B 3.6-6 5/31/89 1 0 ! Amendment 9 1R94
4 MM 2 M 610 M P Romantard Plant m-2 e i Containment Air Locks , 8 3.6.1.2 kAtti fremtinued) ACTION A.1. A.f. A.3 (cor,tieved) (continued) If the outer door is inoperable. then it may be easily accessed to repair. However. If it is stuck in position and must be , accessed through the containment via the inner door, or if the inner door is the one which is inoperable, then it s6y require when the containment boundary a is short time (during not intact. access abi The unlimited only)lity to open the OPERABLE ! door, even if it means the primary containment boundary is temporarily not intact is allon*d for 24 hours untti the OPERABLt door is locked closed. After that, the cumulative time the OPERABLE door may be open for access under admini. , strative control must not exceed one hour per year. This is i acceptable due to the low probability of an event which would pressurtre primary containment during the short time in which the OPERABLL door is open. a.1. B.f. R.3 - If the interlock mechanism is inoperable in one or both air locks, then either OPERABLE air lock door can be closed and the l primary containment boundary remains intact. Therefore, en f l OP[RABL[ door must be closed and if the interlock cannot be resto-ed to OP(RABLE status the door must be locked closed and ' ' periodically verified to be locked. However. personnel access underadministrativecontrolisallowedatanytimeduring continued plant operation where the administrative contria procedure will assure that when the locked door is unlocked the interlock function will be performed manually. Therefore, the integrity of the primary containment boundary will not be threatened by the failure of only the air lock interlock mechanism. , t.1. C.2 , If the air lock is inoperable for reasons other than Condition A or B (e.g. a problem with the barrel or problems with both doors in an air lock) then the containment boundary may not be I intact even when the doors are closed. One door in the air lock is closed to minimize the affect of any potential release
- to Primry Containment. Twenty four hours is allowed to restore the air lock to OPERABLE status due to the low probe.
bility of an event which could pressurite Primary Containment. (continued) l 1 l ABWR B 3.6 7 5/31/89 O Amendment 9 IMA7 l 1 _._ ~ _ ___ . . . . . _ . - . _ _ _ _ _ _ _ _ . . _ . _ _ _ _ ____ _ _ _ . _ _ _ . _ . _
2M61oQAP Remondmed Plant ,, a j i t Sl Containment Air Locks ! 8 3.6.1.2 RAlti fenntinued) p AETION$ D.1. D.f. D 3 STATEMENT $ (continued)- When Condition A (one door or one door and assectated interlock inoperable) applies to both str locks, the one OPtitABLE door in each air lock must be closed and locked. With a problem in both air locks, it is not possible to access primary tentain. ment through an OPERABLE air lock. Furthermore, since one de,or > 1s inoperable in each air lock, the primary containment boun. dary will not be intact during the short time period when the OPERABLE door is opened for access. It is not reabonable to continue plant operation with both air locks in this condition. ' Therefore, at least one air lock must be restored to OPERABLE l status within 7 days. U ; If Condition A applies to one air lock and Condition B applies to the other air lock, the Required Actions of both Conditions are to be followed. The actions apply to the respective air locks and provide appropriate controls for the specified conditions. ,! F.1. F.2 , If the Required Actions and associated Completion Times of Condition A. B. C, D or E are not met or both air locks are , inoperable for reasons other than Condition D or E (i.e., they
- are more severely damaged than en inoperable door and interlock function on one air lock and an inoperable interlock function in the other air lock), then the reactor is required to be in MODE 3 and subsequently in MODE 4 where the air locks are no i longer required to be OPERABLE.
h I canelation Timet All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. (continued) l [: 1 l ABWR B 3.6 8 5/31/89 9 Amendment 9 1M.9-8
MM zwtooAP , Remndard Plant i ! Containment Air Locks i B 3.6.1.2 RAtts feartinuen
$URyt!LLANCE 1R 2.6.1.2.1 REOUIREMENTS (continued) The everyseal air to 7 days flask pressure ensure the sealissystem verifiedremains to be tvia
[ ble]. psig 1R 3.6.1.2.2 The interlock mechanism is vertited to be OPERABLE to assure that both doors cannot be inadvertently opened simultaneously. This is not vertfted when the containment is inerted because of ' the oxygen which is released into containment when the inside air lock door is opened. It is performed prior to entry into the primary containment, unless it has been performed within the previous 6 months, if primary containment remains inerted for longer than 6 months. If primary containment remains inerted for longer than 6 months, then it is tested prior to subsecuent entry when the containment has been de inerted. , 1R 3.6.1.2.3
+
A seal pneumatic system test to assure that pressure does not decay > [ ] psi in a [ ] hour period is an effective leak rate test to vert y system performance. ER 3.6.1.2.4 . Maintaining Primary Containment recutres compliance with the leak test requirements of 10 CFR 50. Appendix J (Ref.1). Therefore, this sveveillance reflects the leak rate testing recuirements with regard to containment air lock leakage (Type B leak tests). Type B testing must be performed in accordance with Reference 1 or NRC approved exemptions. These periodic testing requirements verify that the containment leak rate does not exceed tse leak rate assumed in the safety analysis. The ER states the overall air lock leak rate and door seal leak i rate acceptance criterta as required by 10 CFR 50 Appendix J. The surveillance frequency is requirea by Appendix J, and as such, SR 3.0.2 (which allows surveillance frequency extensions) does not apply. i turve111ance Frenuancias in general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perfore the test, the ease of performing the test and a itkelihood of a change in the ! system / component status. (continued) ABWR B 3.6.g $/31/89 O Amendment 9 1(A9 9 l
,..L- ,-. . . , . _- ... --. - _ . . - -. :
ABWR
.mo,ar Standard Plant m. i G
tentainment Air Locks 8 3.6.1.2 RA$t$ fenetirued) REFlR(N;l$ l. 10 CFR 60. Appendia J.
- t. ADWR $$AR. Section 6.t.
- 3. hl00 31466. ' Technical $pecification screening trtteria Appilcation and Risk Assessment.' hovember 1987.
i t i l l O l I l l l ABWR B 3.6 10 $/31/89 O jus 30 Amendment 9
}
MN 21A610ur
- m. a l
**==dard Plant 16A.9.1J 334.1.3 Pehry Costalaament Pressum l O l t
I primary Containment Pressure B 3.6.1.3 ! l B 3.6 CONTAINM[N1 $YSTEMS e B 3.6.1.3 Primary tentainment pressure kAin , BACKGROUND The Itatt on the primary containment pressure was developed as ' a reasonable upper bound based on operating plant empertence. A positive internal pressure alluias that air does not leak in to a deinerted contatnment. The maximum liettations on the primary containment pressure ensure that the peak LOCA primary containment pressure does not exceed the maximum allowable pressure of 45 psig (Ref.1) and also prevents inadvertent strams due to high drywell pressure. i APPLICABLE Containment perfomance is evaluated *er the entire spectrum 5ATETY of break stres for postulated Loss of. Coolant Accidents (LOCA). ANALYSIS Among the inputs to the design basis analysts is the initial >
. Analyses assume an containment initial primaryinternal pressure containment (Ref. a1)f 0.75 psig. This pressure ,
limitation assures that the peak LOCA containment internal ! pressuee does not exceed the maximum allowable of 45 psig. ; Containment Pressure satisfies the rer,utrements of Selection ; Criterton 2 of the NRC Interim Policy 5t6tement on Technical Specification improvements as documented in Reference 2. LCD A maximum allowable primary containment pressure of 0.75 psig is required to assure that containment initial conditions are consistent with the safety analyses assumptions so that peak.LOCA internal containment pressures are within the design value. l APPLICABILITY The primary containment pressure limitation is appitcable during NODES 1, I and 3 consistent with the requirement for when the Primary Containment must be OPERABLE. In MODE 4 and 5 there is insufficient energy in the reactor to pressurite coritainment beyond the calculated design basis values even if the initial internal pressure is not within its Itmits. (continued) ABWR 8 3.6 11 5/31/89 O IM 9 ll Amendment 9 i
. _ , . . - - - . . , _ . - m.~.....,
I I g thM10W ; m.A-rd Plant n= A 1 Ol ! Primary Containment pressure S 3.6.1.3 ; Ratti tenatinuadi _ f AC110N$ A.I. k.1. R.2 , i With the primary containment pressure not within the limit, the , tenditions in the primary containment will not be consistent with those assumed in the Reference 1 anfety analysis. ; However, continued operetten is 411swed for a lietted time while the containment internal pressure is rettet9d to within t iloits due to the low probability of an event which would pressurite the primary containment, if the required primary , containment pressure is not restored to within the limit and the associated (espletion ilme has expired, then the reactor is required to be in MODE 3 and subsequently in MODE 4. j Ganelation Times t All Completion ilmes are based on industry accepted practice and engineering judgement considering the number of evallable systems and the time required to reasonably complete the Required Action. i :
$UdVIILLANC[ 1R 3.6.1.3.1 Rl0VIREM[N1$
The Reference 1 analysis assumes an initial primary containment pressure of 0.75 psig prior to a postulated LOCA. Surve111ance i l of the primary containment pressure assures consistency with this assumption. . f turvatilance Frenuenetes l In general, surveillance frequencies are based on industry ( accepted practice and engineertog judgement considering the ~ unit conditions required to perfore the test, the ease of performing the test and a itkelihood of a chang'e in the i systee/ component sistus. , ABWR $$AR. Section 6.2. REFERENCE $ 1.
- 2. NEDO.31466 ' Technical Specification $creening Criteria +
Application and Risk Assessment.' November 1987. . t ABWR B 3.6el2 F/31/89 } ! O Amendment 9 IfA912 _ __ ,~__ _ _ . . . ~ . _ ... _. _ _... _ _ ._._ ,_ _,. ._,- ,. _. _ _ _ , _ _ . _ _ _
MM 2M6touP u j RemmAard Plant i 16A.9.1A B34.1A Drywell Avempe AirTeenperaten v(3 ! I i i i Drywell Avg Air Temperature 8 3.6.1.4 , j-3 3.6 CONTAllM(NT $YSTEMS j t 3.6.1.4 br aali Araraan Air femmarature matte BACKGRDUND The drywell contains the reactor vessel and piping which add ' heat to the strspace. Drywell coolers remove heat and maintain a suitable environment. The average atrapace temperature effects eevipment OP[RAllLITY end itfe, personnel access, and , the calculated response to postulated design basis events. The limitation on the drywell average air temperature was developed as a reasonable upper bound based on operating plant 1 esperience. The limitation on drywell temperature is used in the Reference I safety analysis. Although used as an initial 4 condition for safety analyses, recent studies have shown that the initial drywell air temperature has negitgtble effect on ' the peak calculated drywell pressure and temperature. l APPLICABLE Containment performance is evaluated for the entire spectrum of
$AFETY break sites for postulated Loss.cf Coolant Accidents (LOCA)l ANAtY$l$ Among the inputs to the design basis analysts is the initia O Drywell everage air temperature (Ref. 1). Analyses assume an initial average drywell air temperature of Ill'I. Drywell Average Air Temperature satisfies the requirements of Selection Criterion 2 of the NRC Interim Policy $tatement on Technical 4 Specification leprovements as documented in Reference 2. l l
LCO A maximum allowable drywell air temperature of 135'F is required to assure that the initial drywell air temperature assumed for the safety analysis is not exceeded. f I APPilCABILITY The drywell everage air temperature limitation is appitcable during MODIS 1, 2 and 3 consistent with the requirement for when the Primary Containment must be OPERABLE. In PODE 4 and 6 there is insufficient energy in the reactor to heat up - containment beyond the calculated desten basis values even if the initial temperature was above its limit.
?
(continued) ABWR B 3.6 13 $/31/89 O Amendment 9 lu913 l l 1- . _ - - - . _ . - _ . . . _ . . . , _ . . _ . , _ . _ - . _ . . _ . . _ . _ , , _ . , _ _ _ . . . _ _ _ , . _ _.,._. . _ _ . -
t 1
-Renews.ed MN Plani 2.u6100AP n, i fi
)
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Oi f I Drywell Avg Air Temperature l B 3.6.1.4 l tatti fenntinued) f a.1. a.1. a.2 b ACTIONS i With the drywell average temperature limitation not set, en initial assumption in the analysis of the containment response l to a LOCA is not met. However, continued operation is allowed ' for a limited time while attempting to restore toeperature to ' within limits due to the law probability of an event which l-would pressurtre the primary containment. If the drywell average temperature is not restored within the 11stt and the ! I associated Ceapletion time has empired, the reactor is required > to be in MODE 3 and subsequently MODE 4 r - lation timet . All Completion ilmes are based on industry accepted practice ' and engineering judgement considering the number of available systems and the time required to reasonably complete the j Required Action. i
$URV[lLLANC[ 1R !.E.1.a.1
, REQUIREMENTS - Surveillance of the drywell average air t reture assures that the drywell air temperature remains be ow the specified 11eit. Drywell air temperature is monitored in all quadrants _i and at vertous elevations. The instruments are uniformly distributed so that en arithmetic everage is an accurate representation of the actual average 'semperature. Surveillance Freauencies in general, surveillance frequencies are based on industry
- accepted practice and engineering judgement considering the unit conditions required to perfors the test, the ease of performing the test and a likelihood of a change in the i
system / component status. , i REFER [NCES 1. ABWR $$AR, Section 6.2. 1
- 2. NED0 31466. ' Technical Specification Screening Criteria '
Application and Risk Assessment ' November 1967. J ABWR B 3.6 14 5/31/89 i i Anundment 9 3pt9 34 e i 4 7
1 I i M 2M610MP u Reasulard Plant MA.9.1.5 R3.6.1.5 Prismary Containawat and Pnessa loonstion Yalves l 4 Primary Containment and Pressure Isolation Valves B 3.6.1.5 ]
$ 3.6 CONTAINMENT SYSTEMS B 3.6.1.5 prinner e.antat===nt and preasure taniation vaivas ;
hAtti i BACKGRDUND The function of the Primary Containment isolation valves e (KlV), in cambination with other accident attigation systems, i is to Itatt fission product release durin postulated design basis accident (Ref. 4)g less to values andthan following the ; that which would result in offsite ooses greater than 10 CFR 10011mits. Containment isolation within the time limits specified for those isolation valves designed to close auto. natica11y ensures that the release of redloettive material to the environment will be consistent eith the assumptions used in the analyses for a LOCA. l The PCIVs are required by General Design Criterior. (GDC) 54, . 55, 66 and 57 (Ref. 2). These GDCs set guidelines for the isolation capability of lines that penetrate primary cor.tain. i ment. The requirements depen'j upon if the 1tne connects to the . containment air space, is part of the reactor coolant pressure boundary, or connects to some other component inside contain-most. The GDC includes requirements for design and leakage. This LCO is intended to ensure that those GDC requirements are
\ satisfied during plant operation. -
( The function of the Pressure Isolation Valves (P!Vs) is to protect the low pressure portion of systems which are connected to the primary reactor coolant system. All PlVs are destyned f to meet the requirements of Reference 3. APPLICABLE The analytical methods and assumptions used in evaluating the SAFETY radiological consequences of design basis accidents are pre. ARALY5[5 sented in Reference 4. The special assumptions used in Reference 4 regarding leakage of the M51Vs is discussed in the Bases for $R 3.6.1.5.g. The containment isolation valves are also assumed to function so that the bases for the Reference 5 containment response analysis is valid. The maximum isolation times for Primary Containment isolation valves are consistent with those used in the Reference 4 accident analysis. The Primary Containment isolation valves satisfy the require. ments of Selection Criterion 3 of the NRC Interim Policy ' Statement on Technical Spectittation leprovements as documented in Reference 5. (continued) ABWR B 3.6 15 5/31/89 l O 1M.915 Amendment 9 l_.- _ - .. _ _ . . . _ _ __- _ ___ ,_ _. - - , - _ - _ _ _ _ _ _ .- . _ _ . .,
m -. . _. ._. _ __ __ _ _ _- -- ._ _ - - _ _ ~ _ - - _. _--.__ _
?
' MM 21A6100AP
- m. a l
Remndard Plant O , Primary Containment and Pressure Isolation Valves , ! B 3.6.1.5 kitit frontinued) LC0 The OPERABILITY of the primary containment isolation valves ensure that the containment atmosphere will be 15014t4d from the outside environment during an event with the potential for release of radioactive material to the containment atmosphere or pressurtsation of the containment. The OPERABILITY of the Pressure Isolation Valves ensures that low pressure portions of plant systems will be protected. By checking valve OPERABILITY en any penetration which could contribute a large fraction of the design leakage the total leakage is maintained at less , than the destyn va,lue. The isolation valves are considered OPERABLE when their tsala-tion times are within Itaits, the valves isolate on a simulated , automatic isolation signal, the valves meet leakage rate requirements. and excess flow check valves actuate within the required differential pressure range. APPLICABILITY The Primary Containment and Pressure Isolation Valve functions are required for all modes which could involve significant releases of fission products to the enytrennent. This includes i MODIS 1. I and 3 consistent with the reestrement for when the Primary Containment must be OPERABLE. In M00tl 4 and 5 there l is insufficient energy in the reactor to cause a significant release even if the primary containment were not isolated. However, certain valves are required to be OPERABLE to prevent inadvertent reactor vessel draindown. These valves are those whose associated instrumentation is required to be OPERABLE per LCO [ Primary Containment isolation Instrumentation) (this does not include the excess flow check valves which isolate the associated instrumentation). ACT10NS L2 With one or more isolation valve inoperable, a remaining OPERABLE isolation valve in an effected penetration is adequate to perfors the intended isolation function. However, the j overall reliability is reduced because a single failure in the I remaining OPERABLE isolation valve could result in the loss of the isolation function for that penetration, if the affected penetration has two isolation valves, it must be vertited that the other valve is OPERABLE. This action is satisfied by examining logs or other information, to detemine if the valve is out of service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate OPERABILITY of the valve. If the affected penetration is designed with only one isolation valve. the (continued) l ABWR 8 3.6 16 $/31/89 O) Amendment 9 16A.9-16 _ . -~ __ _. _ . _ _ _ _ _ . _ _..____.o_...._ _. _ _ _ _ . . _ . _ _ -
t MM R*=ndard Plant 21A610MP mA [ i i O l I Primary Containment and Pressure Isolation Valves B 3.6.1.6 .
?
kActi frontinued) > AC110N$ penetration is such that plant safety is not Unduly threatened (Continued) by a temporary inab111ty to isolate the penetration. A.2.1. A.f.f.1. A.f.f.f. A.f.f.3 In addition to Required Action A.I. the inoperable valve (s) must be restored to OPERABLE status, or the 1tne must be manually isolated. Acceptable isolation methods for KlVs and FIVs are stated in the Required Actions. If a valve is both a ' PCIV and a PlV then the most 11miting of the two Required Actions must be met. Only those check valves which have been demonstrated to not exceed their allowable feaksee rate liett at their last surveillance may be used to meet the PIV requirement. Finally. it must be periodically vertfled that the affected penetration remains isolated, i R.1. a.f. c.1 If the Required Actions and associated Completion Times of Coedition A cannot be met in MODC$ 1. 2 or 3 then the reactor is required to be in MODE 3 and subsequently MODE 4. If the reactor is already in MODE 4 or 5. ar.d the valves are required to be OPERABLE, CORE Al1[ RATIONS must be immediately suspended. A i However, this does not preclude completion of movement of a component to a safe, conservative position. The OPERABILITY of valves in MODE $ 4 and 5 is governed by the requirements for the associated actuation instrumentation. Comeletion Timet All Completion ilmes are based on industry accepted practice s and engineering f,.edgement considering the number of available systems and the time required to reasonably complete the Required Action.
$URVt!LLANC[ 1R 3.6.1.E.1. $R 3.6.1.E.3 REQUIREMENTS The analyses in Refer: aces 1 and 4 are based on a leak rate which assumes that all primary containment penetrations that tre not closed by an automatic primary containment isolation valve which are reoutred to be closed during accident condt-tions are in fact closed. 1he manuti valves are those which do not receive an automatic isolation signal. Surveillance of
- these penetrations is required to assure that this assumption ts met. Those valves which are readily accessible (i.e.,
outside the primary containment) are surveilled every 31 days. (continued) ABWR B 3.6 17 $/31/89 1(A917 Amendment 9 I . 1 {
I MN 3M6100AP ! RemmAmed Plant m. i i 9: r Primary Containment and Pressure Isolation Valves l 8 3.6.1.5 , McF1 frantinued) i.e., inside ' SURVEILLANC[ Those valves which are or more difficult to need inspect on (ly be REQUIRtMINTS the drywell, containment, steam tunnel) ! survet)(Continued) led in MODES 4 and $ when the primary containment is de inerted and the valves have not been checked in the previous 92 days. l 1R 3.6.1.E.2 i The Transverstne In Core Probe (TIP) isolation valves are regular ball valves that are automatically actuated provided each associated flP 18 fully withdrawn (the withdrawal function also occurs automatically). As a backup, thre are also menu. ally inlttated sheer valves (one per Itne) that are actuated by l emplosive charges. Surveillance of emolestve charge continuity ' provides assurance that the valves will actuate when required. 1R 3.5.1.E.4. 1R 3.6.1.E.E , These surveillance requirements demonstrate 6sch power operated or automatic primary containment isolation valve is OPERABLE by vertfying the isolation time to be within limits when tested pursuant to SR 3.0.5. $R 3.0.5 reoutres periodic inservice testing of valves in accordance with Reference 3. Section IWV.3413 of Reference 3 requires that str6ke times be specified and periodically measured. The stroke time Itatts are spect. fled in Ref. 1. These surveillance requirements provide assurance that the closure time of automatic isolation valves and MS!Vs will not exceed trat used in References 1 and 4 so that predicted consequences will not be exceeded during the postulated events. 1R 3.6.1.E.6 This survetilence requirement provides assurance that the , isolation valves will actuate to the proper position upon '. receipt of the respective containment isolation signals. This requirement complies with Reference 3 to periodically test OPERABILITY of isolation valves. (continued) ABWR B 3.6 18 5/31/89 O Amendment 9 16A.918
ABWR m 2=r Remeulard Plant ari i i
. i O :
l Primary Containment and Pressure Isolation Valves B 3.6.1.5 l matf1 frantinued) SURVilLLANCE 1R 3.8.1.E.7 , Rt0VIRLMEN15 (continued) This surveillance requirement demonstrates that each reactor instrumentation line encess flow check valve is OPERABLE by verifying the valve reduces flow to s I gallen per hour on a simulated line break. This survel11ance requirement provides assurance that the instrus.antation line excess flow check valves will perform so that predicted radiological consequences > will not be orceeded during the postulated instrument line break event evaluated in Reference 5. ! 1R 3.6.1.E .8 The TIP manual isolation shear valves are actuated by explosive charges. An in place functional test is not possible with this design. The emplosive soutb is removed and tested to provide 4ssurance that the valves will actuate when required. The replacement charge for the emplosive squib shall be from the , same manufactured batch as the one fired er from another batch ', which has been certified by having one of that batch success-fu11y fired. . 1R 3.6.1.5.9 The analysis in Reference 4 is based on leakage less than the specified leak rate. Leakage through each M51V should be s 20 acf/hr when tested. However, the analysis of Reference 4 assumes that only five of the eight valves meet this requirement. It is assumed that two of the other three MSIVs have leakage of 100 scf/hr and the remaining valve 500 scf/hr. The worst combination of leakage and fatture of one M51V to isolate would then result in a tenbined leakage of 640 scf/hr. sR 3.6.1.E .10. ER 3.5.1.6.11 Leakage through each reactor coolant P!V must be < 0.5 gpm per nominal inch of valve size, though if the valve is 20 inches or larger, the maximus P!Y leakage is 5 gpm. Note that the Plv surveillance is only required in MODt51 and 2; MODE 3 may be used to perfore the required demonstration. Leakage through each hydrostatic 411y tested itne whicn penetrates Primary Containment is not to exceed 1 gpm when tested at ( psig. Surveillanceoftheseleakaosratesprovidesassurance)that Referer.ces 5 and 6 calculation assumptions are met. Note also that dual function valves must pass all applicable $Rs including the Type C leak rate test (SR 3.6.1.5.12)if appropriate. (continued) ABWR B 3.6 19 5/31/89 O Amendment 9 N I'
2M6100AP Standard Plant u.2 i L 9l ; i Primary Containment end Pressure Isolation Valves 8 3.6.1.6 RAlf1 ftintinuad) , i t
$URV[lLLANC[ 1R L6.1.1.12 R[OVlR[M[NTS (tentinued) Matttaining the 16 elation function requires compitance with the .
Type C leak rate tests of 10 tiR 60. Appendit J. These ' periodic testing requirements verify that the isolation valves wt)) not leak excessively when closed. ; f turve111anta Freausneten In general, surveillance frequencies are based on industry , accepted practice and engineering judgement considering the unit conditions required to perfore the test, the ease of i perforsing the test and a Itke11 hood of a change in the system / component status. R[f[RLNC[$ 1. ADWR $$AR, Section 6.2. ; 1
- 2. 10 CFR 50, Appendix A. l
- 3. ASM[ Code Section 11, subsection IW.
- 4. ABWR $$AR, Chapter 16.6.
- 5. N[D0 31466 ' Technical Specification screening Criteria Appitcation and Risk Assessment,' November 1987.
1 l l ABWR B 3.6 20 $/31/89 9 16A.9 20 Arncadment 9
I I MM Susieur ma I
$gymulard Plant )
16A.9.14 B34.14 Wetwell to DrywellYacewin Brookers l l l 1 l Wetwell to Drywell VB ) B 3.6.1.6 l B 3.6 CONTAINMENT $Y$TEMS j 8 3.6.1.6 blatum11 ta Drmua11 Vaema treakers I RAsts ShCKGROUND The function of the Wetwell to Drywell Vacuum Dreakers is to relieve vacuum in the drywell, 1here are 8 vacuum breakers located within the wetwell atrspace on separate lines penetrat. Ing the pedestal well at equal distances around the pedestal. l The vacuum breakers allow air and steam flow from the wetwell ' airspace to the lower drywell when the drywell is at a negative l pressure with respect to the wetwell. Each vacuum breaker is a , self actuating ulve similar to a check valve which can be l remotely operated for testing purposes. ; Negative pressure differentials across the drywell walls are caused by rapid depressurtsstion of the drywell. Events which cause this rapid deptassuritation are cooling cycles, inadvert. . ent drywell spray actuatier., and steam condensation from sprays i or subcooled water reflood of a break following primary systes > ruptures. Cooling cycles result in einer pressure transients i in tr.e drywell which occur slowly and are nomally controlled by heating and ventilating eculpeent. Spray actuation or spill
, of subcooled water out a break result in more significant pressure transients end becomes important in siting the external vacuum breakers.
$ team condensation following prteary ruptures within the drywell results in the most severe pressure transients, j following a primary systen rupture air in the drywell is purged into the w twell airspace Itaving the drywell full of stene. Subsequent condensation of the steam can be caused by '
two possible transients, ECCS flow from a Itne break or manual containment spray actuation following a Loss of Coolant (LOCAL. These two cases determine the maximum depressurization rate of the drywell. . In addition, the water leg in the ABWR vent system downtomer is controlled ley the drywell to suppression chamber differential pressure. If the drywell pressure 15 less than the suppression chamber pressure then there will be an increase in the vent water leg. This will produce an increase in the water clearing inertie during a postulated LOCA, Consequently, this will result in an increase in the peak drywell pressure and pool swell dynamic loads during a postulated LOCA. The internal vacuum breakers limit the height of the water leg in the vent system during normal operation. I (continued) ABWR B 3.6 2) 5/31/89 O l 1 4 .. i 9
- 22 l
l
__ . _ _ . _ _ ~-- _ _ _ _ _ . ~ . _ _ _ _ _ _ _ _ _ _ _ _ _
- Y !
MM 2M610MP
- m. a Remmulard Plant
. i . 9; i Wetwe11+te Drywe11 VB t 3.6.1.6 l hAtti fenntin w n 3 APPLICABLt Analytical methods and essumptions involving the wetwell. ! SAf[TY to Drywell Vacuum treakers are presented in Reference 1 , ANALY$ts as part of the accident response of the contatament systems. > Internal, wetwell to drywell vacuum breakers are provided as , part of the primary containsent to Itait the negative pressure ; differential across the drywell and wetwell walls. ; The safety analyses assume the internal vacuum breakers to be ! closed initially and to be fully open at a differenttal pressure of 0.6 psid. Additionall vacuum breakers 16 assumed to fsi)y, in a one of the closed pc31 eight tion internal (Ref. 1). The results of the analyses show that the design pressure i is not escoeded even with the worst possible accident scer.ario. l The vacuum breaker opening differential pressure Setpoint and t
- the requirement for seven vacuum breakers to be operational are >
a result of the requirement on the vtcuum breakers to limit the i' vent system water les height. The total cross soettonal area of the main vent system between the drywell and wetwell needed
- l to fulfill this requirement has been estabitsbed as a minimum
! of ( times the total break area (Ref.1). b turn, the 4
vacuum rei)tef capacity between the drywell and wetwell should (' be[ ] of the total main vent cross sectional area with the . valves set to operate (begin to open) at 0.2 pstd pressure differential and be fully open at 0.5 psid. lietwell to Drywell Vacuum Breakers satisfy the requirements of Selection Criterton 3 of the NRC Interim Policy Statement on I Technical 5ptcilication leprovements as documented in
- Reference 2.
I i l
, LC0 Seven vacuum breakers must be OPERABLE for opening, and all
, eight Wetwell to Drywell Vacuum treakers are regulred to be closed (except during testing or when the vacuum breakers are l performing their intended design function). The requirement , provides assurance that the drywell to suppression chamber ! negative pressure differential remains below the design value, i The closed requirement assures that there is no excessive
- bypass leakage should a LOCA occur.
1 APPLICABILITY The Suppression Chamber to Drywell Vacuum Breakers must be
' OPERABLE and closed during MODES 1. I and 3 consistent with the requirement for when primary containment must be OPERABLE. In MODES 4 and 6 there is insufficient energy in the reactor to
. heat up and pressurite the drywell to the point that a
- significant drywell depressurttation event could occur.
(continued) i ABWR B 3.6 !! $/31/89 Anwedment 9 HA9-22
MN zu61our Rar A Stedard Plant r Wetwell to Drywell VB
$ 3.6.1.6 katti febrtinued)
ACTION $ Q An open vetuum breaker results in leakage betwen the drywell and suppressten chamber and the potential for sueettssion thenber overpressure due to bypass leakage if a LtlCA were to occur. Therefore, the open vacuum breaker must be closed. A short time is allowed to close the vacuum bretker due to the low probablltty of an event which would pressurtre the primary containment. If vetuum breaker positten indication is not re11eble, en alternate method of vertfying the vacuum breakers are closed is to verify that a differential pressure of 0.5 pst ' between the torus and drywell is maintained for one hour without makeup. M With one of the required vacuum bretkers inoperable for reasons other than Condition A (e.g., the vetuum breaker is not open, and may be stuck closed or not within its opening setpoint Ilmit such that it would not function as designed were en event OP[RABLE vetuum breakers are capable of)providing the vacuumto occur that depre relief function. However, overall tystem reliability is ( reduced because a single failure in one of the remaining vacuum breakers could result in en excessive wetwell to drywell differential pressure during a design basis event. Therefore, with one of the eight required vacuum breakers inoperable, a limited time is 6110wed (consistent with other primary Contain-ment functions) to restore the inoperable vacuum breaker to OP[RABLE status so that plant conditions are consistent with those assumed for the design basis analysts, c.1. C.2 With one Suppression Chamber to Drywell Vacuum Breaker open and j one of the required vacuum breakers otherwise inoperable, the Required Actions of Conditions A and B must be satisfied. D.1. D.2 If the Required Actions and associated Completion Times of t Condition A. 8 or C are not met, the reactor is required to be l in MODE 3 and subsequently in MODE 4. i (continued) l ABWR B 3,6 23 $/31/89 O IMO Amendment 9
-+ ... - . - . , - - .. _ ._ ,, , , . - _ _ . - __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . ___
i MkN 2M610MP
- m. a Remadard Plant !
f I e; i Wetwell.to Drywell VB i b 3.6.1.f ' hits febatinued) ; ACTIONS r = 1ation Timat ; (continued) All Completion times are based on industry accepted practice , and engineering judgement considering the number of available , systems and the time required to reasonably complete the - Required Action.
$URV[lLLANCE 1R 3.6.1.6.1 REQUIREM[h'15
[ach vacuus breaker is vertited to be closed (except when being ; tested in actnrdance with $R 3.6.1.6.2 or when the vacuum ; breakers are pertersing their intended design function) to ensure that this potential large bypass leakage path is not present. This can be performed by observing the valve indica-tion or, if vacuum breaker indication is not M 11able, an alternate method of verifying the vacuum breakers are closed is to verify that a differential pressure of 0.5 psi between the torus and drywell is maintained for one hour without makeup. 1R 3.6.1.6.7 [ach vacuum breaker must be cycled to ensure it opens ade. ' i quately to perfors its design function and returns to the fully ( closed positton. This assures that the safety analysis assump. tions are valid. 1R 3.6.1.E.3 Verification of the vacuum breaker opening setpoint is neces. sary to assure that the safety analysis assumption regarding vacuum breaker full open differential pressure of 0.5 psid is valid. , Eurve111ance Freauencian - In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perfore the test, th9 ease of perforsing the test and a Itkelthood of a change in the system / component status. (continued) l 1^ 1 ABWR B 3.6 24 .$/31/8g 9 Amendment 9 IM9 24
- - - . _ _ _ _ _ . - . _ - . _ _ _ _ , _ _ _ _ _ . . ~ _ .
i 1 31A6100AP Remndard Plant nu ! l I b) ) htwellet0*Drywell VS B 3.6.1.6 j kkiti ftantinued) R[F[R[NC($ 1. ABWR $$AR. Section 6.2. !
- 2. NID0 31466, ' Technical Specification Screening Criteria ,
Application and Risk Assessment,' hovember l H7. i i i i i O + v ABWR B 3.6 t$ 5/31/89 O Amendment 9 16A.9-25
----,-, ..+,.--- -,--.. - --,.,..--. , - - - - - - - - -,- -- -
i MM Re==dard Pimut su6er mi f i 16A.92 R34.2 Suppression Pool I 16A.92.1 R34.2.1 Suppression Pool Average Tesaperature i i Suppression po:1 Avg tem S 3.6.t. ; B 3.6 CONTAllettW1 SYSTEMS {
$ 3.6.2.1 Ankare& Alan paal averama f u maratura l aatts BACKGROUND The suppressten pool is a concentric volume of water contained ;
within the stainless steel liner of the wetwell which is located at the bottom of the containment, separated free the ' upper and lower drywell by the diaphrage floor and reactor ' pooestal respectively. The suppression pool is designed to absorb the secay heat and sensible heat released during a reactor blowdown from safety /reitef valve discharges or from a
. i Less.cf.
steam from Coolant the beactor Accident (LOCA)letion Cooling (RCIC) systemThe pool must also Core Ise turbine exhaust and provides the main emergency water supply source for the reactor vessel. The amount of energy that the pool can absorb es it con 4enses steam 16 dependent upon the intitel pool temperature. The lower the inttle) pool temper. , ature, the more neat it can absorb without heating up exces, stvely. Since it 15 indirectly connected with the upper and . lower drywe11 through the connecting vents and vacuum breakers 1 its temperature will effect both containment pressure and i average air temperature. Using conservative inputs and methods, the maximum calculated containment pressure during and following a design basis accident must remain below the containment design pressure of 45 psig. In addition, the maximum containment everage air temperature sust remain telow 340*F. The allowable suppression pool temperatures are based on the following:
- 1. g5'F is the initial condition for the analysis detersining ,
pool performance and satisfies the post.LOCA long.ters ! peak pool temperature of 207'F. It is also the initial condition for the long tern suppression pool heatup analysts for Safety / Relief Valve (5/RV) transients.
- 2. The original basis for the reactor pressure vessel depressurization liett was the todega Bay Tests (Ref 5) which were conducted with a peak pool temperature of 170'F. The Reference 3 analysts indicates that a reactor blowdown will raise the pool water temperature approxt. ,
mately 50'F. So to stay below the 170'F limit, the operator is instructed to begin reactor depressurization when the pool temperature reaches 120*F. (continued) . ABWR B 3.6 26 5/31/89 O Asneedment 9 16A.9 26
, -.+ -
,-.r .
},
MM tu6s0MP Remsidard Plant u , t
.w Suppressten pool Avg temp 8 3.6.2.1 j ants tenettnued)
DArkGROUND 3. IlO'T and 106'T are derived from engineering judgment (continued) which considers operator reaction time and avoidance of unnecessary scrams and/or deprest-urtrations. l The limitations on suppressten pool temperature are necessary to meet the requirements of Reference 2. APPLICABLE Containment performance is evaluated for the entire spectrue of
$AFETY break sites for postulated LOCAs. Inputs to the safety ANALY$t$ analyses. Inputs to the safety analyses include initial ;
suppressten pool temperature. An initial suppression pool temperaturS of gl'F is assumed for the entlyses of References 1 i and 3. Reactor shutdown at a pool temperature of !!0'F and i initiation of reactor vessel depressurtretten at a pool temper. , ature of It0'F are assumed for the References 1 and 3 analyses. , The safety analyses assure that the 207'i design average pool temperature lisit are not esteeded during a postulated accident. Suppression Pool A,erage Temperature satisfics the requirements of Selection Criterion 2 and 3 of the NRC Interim policy O t $tatement on Techni:41 spettf tcation lepravements as documented in Reference 4. LCO A limitation on the suppression pool temperature is required to assure that the containment tenditions assumed for the safety analyses are met. This liettation subsequently ensures that peak containment pressuru and temperatures do not escoed design values during a postulated LOCA or any transient resulting in heatup of the suppression pool. The LC0 requirements are at follows: A. Temperature g gl'F when [ reactor power is > 15 of RTP] and when net testing equipment that discharges steam to the , suppression pool which ensures that itcensing bases j initial conditions are met. ; B. Temperatures 105'Fwhen(reactorpoweris>15ofRTP) i l and when testing soutpoent that discharges steam to the suppression pool which ensures the plant has testing j flexiblitty. The time period with temperature above g5'F - is short enough to not cause a significant increase in 1 plant risk. l (continued) ; I ABWR B 3.6 27 5/31/89 O l Amendment 9 1M.9-27
)
l 1
--*e e - _ __- --_ _ _ _ . _ . _ _. _ _., , . _ _ . _ , _ . . _ , , , , ,__ _ _ _ , ,
- .- .- - - - . -. . .- .~_ . . - . - - _ _ _ _ . ... . _ _ _
MM 21A610MP me ; Reasadard Plant i i Suppression pool Avg Temp B 3.6.2.1 RAlf! frontinued) LC0 C. 1emperature s 110'F when l whichensuresthatthepla[reactorpoweriss15ofRTP) nt will be shutdown upon esteed- ; ing Il0*F. The pool is designed to absorb decay heat and sensible heat with a pool cooling system operating, but
- could be heated beyond design limits by the steam generated if the reactor is not shutdown.
hotethat[1%ofRATEDTH(lW4ALPOWER(RTP))hasbeenchosenas a convenient measure of when the reactor is productog heat input apprortmately equal to normal system heat losses. [A parameter valve representative of it of RTP will be deterstned ' and substituted at a later date.) APPLICABILITY The suppression pool temperature limitations are appitcable ' during H00t$ 1, 2 and 3 consistent with the requirement for when the Primary Containment must be OPERABLE. In NODE $ 4 and 5 there is insufficient energy in the reactor to significantly heat up the suppression pool even if the initial pool 11stt is not maintained. ' ACTIONS A.1. A.2 With the suppression pool temperature above the specified limit I when not performing testing which adds heat to the suppression pool, the containment conditions onceed the conditions assumed for the Reference 1 analyses. However, containment cooling capablitty still exists and the containment pressure suppression function will occur at temperatures well above that assumed for safety analyses. Therefore, continued operation is allowed for only a limited time to ellow the suppression pool temperature to be restored below the Itait. Additionally, when pool temperature is above 95'F, increased monitoring of sup-pression pool temperature is required to ensure it remains at or below 110'f. L1 l If the Required Actions and associated Completion Times of Condition A are not met, then reactor power must be reduced [to below approximately 15 of RTP). The time specified is consis-tent with the time to be in MODE 3 for other containment actions. At this power level, heat input to the suppression pool is approximately equal to normal system heat losses. Also, the energy required to be absorbed by the suppression pool during a LOCA is greatly reduced. (continued) ABWR 8 3.6 28 5/31/89 O Anwedment 9 1M9 28
. ~ - . . - . . . - - . . , - - . .
MN Remndard Plant 21A6100AP m.a i suppress 1on Pool Avg Temp 8 3.6.2.1 AMrc feentinued) ACTIONS L1 (continued) Suppression pool temperature up to 106'F is allowed during testing which adds heat to the suppression pool. If tempera. ture exceeds 10$'F the testing must be tenediately suspended to preserve the pool's heat absorption capability. With testing suspended, Condition A is entered and the Required Actions and assectated Completion Times are applicable. D.1. D.2 Suppression pool temperature at or above 110'F requires that the reactor be shutdown immediately. This is accomplished by placing the Reactor Mode Switch in the Shutdown position. Additionally when pool temperature is above 110*F increased monitoringofpooltemperatureisrequiredtoensureitremains below 120 F. E.1. E.t Continued heat addition to the suppression pool with pool temperature above 120'F could result in exceeding the design basis maximum allowable values for containment temperature or pressure. Furthsmore, if a blowdown were to occur when i temperature was above 120'F the maximum allowable bulk and local temperatures could be exceeded very quickly. Therefore, the reactor must be depressurized and placed in MODE 4 to limit this potential. Enanistien Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. , SURVEILLANCE SR 3.6.2.1.1 REQUIREMENTS The Suppression pool Average Temperature is regularly monitored (24 hours) to ensure that the r& quired limits are satisfied. When hut is being added to the suppression pool by testing, it is necessary to monitor suppression pool toeperature every 5 minutes since temperature could be increasing rapidly. This . will ensure that allowable pool temperatures are not exceeded. (continued) ABWR B 3.6 29 $/31/89 O Asnendment 9 16A.9-29 ,
i MM Su6100AP Standard Plant m-. A ; i o! ! Suppression Pool Avg Temp ; 8 3.6.t.1 , hatte frantinued) I
$URV(ILLANCE turveillanen Franuanetas Rt0VIREMENTS (continued) In general, surveillance frequencies are based on industry ,
accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of ,
, performing the test and a likelihood of a change in the syste# component status.
l REFER [NCES 1. ABWR $$AR, Section 6.2 . f
- 2. 10 CFR $0, Appendix A. GDC 13.
- 3. ABWR $$AR, Chapter 15. ,
- 4. NEDO.31466, ' Technical Specification Screening Criteria l Application and Risk Assessment,' November 1987. _
i !
$. ' Preliminary Hazards Summary Report Bodega Pay Unit No.
- 1,' Pacific Gas and Electric Co., December 28, 1962.
l 1 l e'> l l l l ABWR B 3.6 30 $/31/89 Amendment 9 1M.9 30 l
.,..,a--..,..-.,, . - , -
-_- . -. ~ . . . -
9 l 2146100AP Semitdmed Plant ma 164.9.2.2 RM.2.2 Suppression PoolWaterlevel O Suppression Pool Water Level B 3.6.t.2 l 1 B 3.6 CONTAINMENT $YSTEMS 8 3.6.2.2 Sumarassion pani katar Laval l RAtts BACKGROUND The suppression pool is designed to absorb the decay heat and sensible heat released during a reactor blowdown from safety / discharges or from a Loss of Coolant reitef valve Accident ($/RV)he (LOCA). T pool must also condense steam from the Reactor Core Isolation Cooling (RCIC) turbine exhaust and ; providesthemainemergencywatersupplysourceforthereacto! - i vessel. ThesuppressionpoolvolumerangesbetweenIt6,gt?ft at the low water level alarn of 22.g7' and [ ]ft at the high water level alarm of 23.2g'. If the suppression pool water level is too low, it could , jeopardire steam condensation from the $/RV quenchers, main vents, or RCIC turbine exhaust. It could also be an inadequate emergency makeup source. The lower volume would also absorb less steam energy before heating up excessively. Therefore, a i minimue pool water level is specified. If the suppression pool water level is too high, it could O result in excessive clearing loadt from 5/RV discharges and excessive pool swell loads during a LOCA. A)PLILABLE Initial Suppression Pool Water Level effects containment pool SAFETY teorerature response calculations, calculated drywell pressure ANALYSES during vent clearing for a LOCA, calculated pool swell loads > for a LOCA, and calculated loads due to 5/RV discharges. Suppression Pool Water Level satisfies the requirements of Selection Criterion 2 of the NRC Interim Policy $tatement on Technical Specification leprovements as documented in Reference 1.
~
LCO A limit that Suppression Pool Water Level be from 22.g7' to t i 23.2g' is required to assure that the containment conditions l assumed for the safety analysts are met. The high or low water level limits was used in the safety analysts, depending upon which is conservative for a psrticular calculation.
- (continued)
ABWR B 3.6 31 $/31/89 O Amendment 9 1R9 31
+
MM me. a.,a pi. e 2146100AP m_ i : t 9 ! r Suppression Pool Water Level [ B 3.6.2.2 matti feantinued)
, APPLICABILITY The Suppression pool Water Level liettations are appitcable [
during MODES 1. I and 3 consistent with the requirement for >
..- when the Primary Containment must be OPERABLE. In MODE $ 4 and
$ there is insufficient energy in the reactor to place significant leads on the containment even if pool water level is not kept within lietts. However, the suppression pool sit 11 ,
provides the primary source of emergency makeup water during these conditions. Requirements for Suppression Pool Water Level during MODES 4 and $ are spectfted in LCO 3.$.t. j l ACTIONS M When suppression pool water level is outside the 11mits, the conditions essumed for the safety analysts are not met. If , water level is below the etnteum level, the pressure suppres. i sten function sit 11 entsts as long as main vents are covered. - RCIC turbine exhaust is covered, and $/RV quenchers are ! covered. If water level is above the maximum level, protection 5 l against overpressurtration still entsts due to the sargin in ' ( the peak containment pressure snelysts or as long as the l drywell and containment sprays are OPERABLE. Therefore, + continued operation for h limited time is allowed to restore , Suppression Poci Water Level to within limits. I i B.I. R.2 If water level cannot be restored to within limits then the reactor is required to be in MODE 3 and subseevently in MODE 4 Requirements for suppression pool water level during operetton i in MODE 4 are specified in LCO 3.$.t. . t tamolation Times -
)
All Completion Times are based on industry cceepted practice and engineering judgement considering the number of available systeas and the time required to reasonably complete the Required Action.
$URVEILLANCE 1R 3.6.f.2.1 REQUIREMENTS The Itcensing bases analysis assumes that initial Suppression Pool Water level is between the specified limits. This SR assures that the assumption is valid.
(continued) j ABWR B 3.6 32 $/31/89 O Amendment 9 1R9-32
ISA6100AP Riandard Plant .m l i I Suppression Pool Water Level ! B 3.6.2.2 RASES fenntinued) SURVEILLANCE turve111ance Franuaneten REQUIREMENTS (continued)* 'In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a likelihood of a change in the systes/ component status. REFERENCES 1. NEDO 31466, ' Technical Specification Screening Criteria Application and Risk Assessment,' November 1987. O ABWR B 3.6 33 5/31/89 Amendment 9 16A.9-33
. - . _. . _ _ _ . - - - - - - . . - - . .~ - - . _ - . . - - -
r 21A61 MAP i R*=adard Plant mi-1sA.9.2.3 B342.3 Residual Heat Removal Suppmsalem Pool Cooling 1-l RHR Suppression Pool Cooling B 3.6.2.3 8 3.6 CONTAINMENT SYSTEMS B 3.6.2.3 Residual Heat Removal Suparassion pool Caelins ! RAtts + l i BACKGROUND Following an accident, the Residual Heat Removal (RHR) system removes heat free the containment. The suppression pool is designed to absorb the sudden input of heat from the primary system and long ters, the pool continues to absorb residual-heat generated by fuel in the reactor core. Some means must be provided to remove heat from the suppression pool so that the temperature inside the primary containment remains within design limits. This function is provided by three redundant , RHR Subsystems. Initially, this is accomplished by the Low Pressure Flooding mode as water is drawn from the suppression pool and pumped through the RHR heat exchangers to the vessel. Eventually this water spills out the postulated break and drains back to the suppression pool. Bee.ause this recircula. tion path includes a pass through the RHR heat enhanger, enough containment cooling is accoglished to adequately control the short term containment response. However, long term pool temperature (and overall containment response) is controlled by-RHR in the Suppression Pool Cooling mode. Each of the three RHR subsystems contains one p and one heat exchanger which are independently controlled ant ' operators. The three RHR subsystems perform suppression pool cooling function by circulating water from the suppression pool, through the RHR heat exchangers, and returning it to the suppression pool. A suppression pool cooling subsystem is OPERABLE when the pump, h3at exchanger and associated otping and valves are OPERABLE. Reactor Building Cooling Water (RCW), [ circulating through the shell side of the heat exchangers, ression pool water, dischar es this-i exchangesheatwiththesupp$rBuildingServiceWater(SW) heat to the associated React I !- loop, and ultimately to the external heat sink. Each 1 subsystem, including the associated RCW and R$W loops, is - I separate and independent, including separate power supply ) l- divisions. ; l I l - The combined heat removal capability of two subsystems is i sufficient to meet the long term post Loss of Coolant Accident (LOCA) pool cooling requirement. Transient events such as turbine trip or a stuck open safety / relief valve ($/RV) also result in discharge of steam from the reactor to the supores-l sion pool, increasing suppression pool water temperature. S/RV
- leakage increases cuppression pool temperature more slowly.
1 The suppression pool cooling mode of the RHR system is used to lower the suppression pool water bulk temperature follewing such events, a. (continued) . ABWR B 3.6 34 5/31/89 O1 1 Asundment 9 16A.9-M
MM
. Sittandard Plant 23A6100AP ma O
RHR Suppression Pool Cooling 8 3.6.2.3
~
RA$fs fenntinued) APPLICABLE Reference I contains the results of analyses used tc predict SAFETY containment pressure and temperature following large and small ANALYSES break LOCAs. The intent of the analysis is to demonstrate that the heat removal capacity of the RHR heat exchangers is ade-quote to maintain the containment conditions within its design limits. The time history for suppression pool temperature is followed to demonstrate that the maximum suppression pool i' temperature remains less than the design limit following the accident event.
- RHR Suppression Pool Cooling satisfies the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification leprovements as documented in ~
Reference 2. LCO Three subsystems of the RHR Suppression Pool Cooling system must be OPERABLE. Each subsystem consists of one pump, one ! heat exchanger, and the necessary piping and valves to circu-late water from and back to the suppression pool to cool the suppression pool. The LCO assures that redundant RHR Suppres. 1 sion Pool Cooling subsystems are available to perform the heat removal function. l APPLICABILITY The RHR Suppression Pool Cooling function is only required to j be OPERABLE in MODES 1. 2 and 3 consistent with the requirement for when the Primary Containment must be OPERABLE. In MODES 4 and 5 there is insufficient energy in the reactor to overheat the suppression pool even if suppression pool cooling was not available. (continued) l t I l { ABWR B 3.6 35 5/31/89 l Amendment 9 16A.945 i L.
MM 2nA6100AP
.StanA=rd Plant - w
- e. ,
RHR Suppression Pool Cooling . , B 3.6.2.3 { BASE 1~fenntinued) _ , ACTIONS L1 When one RHR Suppression Pool Cooling subsystems are 2 inoperable, the remaining OPERABLE subsystems are adequate to perfore the cW.ainment cooling function. However, the overall reliability is reduced because a single failure in the remaining subsystem could result in significantly reduced- ' containment heat removal capability. Continued operation is permitted for a 11mited time due to the low probability of an unplanned event which would require suppression pool cooling be OPERABLE. As noted, the provisions of LCO 3.0.4 are not applicable. p B.1. R.2
- With two RHR Suppression Pool Cooling subsystems inoperable, <
reduced (and possibly ina6 equate) containment heat removal capability remains to handle either accident or transient events. Continued operation is justified because of the low probability of an event which would rsquire suppression pool cooling and because some level of containment heat removal 4 capability still exists. However, continued operation of the plant is permitted for only a limited period of time. The , initial inoperable subsystem is also required to be restored to j OPERABLE status consistent with Required Action A.I. c.1.'t.2. c3 With all three RHR Suppression Pool Cooling subsystems , inoperable, no containment heat removal capability remains to handle either accident or transient events. Continued l operation is justified because of the low probability of an , event which would require su pression pool cooling. However, continued operation of the ant is permitted for only a
- limited period of time. Th initial and subsequent inoperable subsystems are also required to be restored to OPERABLE st6tus consistent with Required Actions A.1, B.1 and B.2.
l D.I. D.2 If the Required Actions and associated Completion Times of Condition A. B or C are not met, then the reactor is required to be in MODE 3 and subsequently MODE 4 where the RHR Suppres. ston Pool Cooling system is not required. If the reactor cannot attain MODE 4 because of inoperable RHR subsystems, the - reactor coolant temperature should be maintained as low as practicable using an alternate decay heat removal method. l (continued) ABWR B 3.6 36 5/31/89 I O Amendment 9 16A.9 36 d 4
- - , - , . - - . ,- - . , , . , _ - - . - . ~ . , _ _ . _ , . - . . . _ . . , , . .. .~ ,_,..,,m yn,4m-,m .
2146)(BAP - Standard Plant mw ; I l V j l RHR Suppression Pool Cooling i B 3.6.2.3 ' RASES fenniinued) ACTIONS Cannletion Times (continued) A11 Completion Times are based on industry accepted practice I and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. SURVEILLANCE 1R 3.6.2.3.1 REQUIREMENTS Successful initiation of suppression pool cooling requires the operation of certain valves. It is verified that each manual. I automatic or power operated valve not locked, sealed, or 1 otherwise secured in position, is in its correct position or ) can be aligned to the position required to operate suppression I pool cooling. l sR 3.6.2.3.2 , I. The components of RHR Suppression Pool Cooling subsystems A B and C are designed to be tested during normal plant operation. Each subsystem is tested to verify that the required flow of 4200 gpm (Ref.1) through the associated heat exchangers is "O achieved. Each subsystem is tested to ensure that all will provide the requisite flow. Surveillance Frenuancies in general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a itkelihood of a change in tiie system / component status.
)
REFERENCES 1. ABWR $$AR, Section 6.2.
- 2. NEDO 31466, ' Technical Specification Screening Criteria Application and Risk Assessment," November 1987, i
l l l l ABWR B 3.6 37 5/31/89 LO Anwndment 9 16A.9-37
, ... . - . _ - - . .__ -- __________~_ __
MM 3sa.ioorr -
**==dard Plant om i
.i I
16A.9.2A R3.6.2A Residual Heat Removal Wetu-It Spray System O' RHR Wetwell $ pray B 3.6.2.4 B 3.6 CONTAINMENT SYSTEMS I B 3.6.2.4 Residual Heat Renoval Matwell inrav RASES , BACKGROUND Wetwell L Following an accident, spray system the Residual removes heat from the Heat Removal wetwell airspace. (RHR)The suppression pool is designed to absorb the sudden input of heat from the primary system. The heat that results in steam in the l' metwell increases primary containment pressure. Sees means must be provided to remove heat free the wetwell 60 that the > pressure and temperature inside the primary containment remain within design limits. This function is provided by two , redundant RHR Wetwell $ pray subsystems. Each of the RHR subsystems contains a pump and a heat exchanger
- l. which are independently controlled by plant operators.
' Subsystems B and C only perfom the wetwell spray function by c circulating water from the suppression pool, through the RHR
- heat exchangers, and returning it to the common wetwell spray l-sparger.- A wetwell spray subsystem is OPERABLE when the pump, heat exchanger and associated piping and valves are OPERABLE.
l-1 The wetwell spray is sufficient to condense the steam from small bypass leaks from the drywell to the wetwell airspace. l ReactorBuildingCoolingWater(RCW),circulatingthroughthe l shell side of the heat exchangers, exchanges heat with the suppression pool water and discharg6s this heat to the external l heat sink, via the associated Reactor Building Service Water ! -(R$W) loop. APPLICABLE Reference 1 contains the results of analyses used to predict SAFETY containment pressure and temperature following large and small ANALYSES break Loss of Coolant Accid 4nts (LOCAs). The intent of the analysis is to demonstrate that the heat removal capacity of the RHR heat exchangers is adequate to maintain the containment structure within its design limits. The time history for containment pressure is followed to demonstrate that the containment pressure remains less than the design limit following the accident event. The RHR Wetwell Spray system satisfies the requirements of Selection Criterion 3 of the NRC Interim policy Statement on Technical Specification leprovements as documented in Reference 2. 1 (continued) ABWR B 3.6-38 5/31/89 i O Anandment 9 16A.9-38 4 x ~- . , - , _ , , - - . . , - . , - - . ..-.a .,aw.. --,.-n. w. .-__.n, . , , . . . , - . , ,
t 23461N Standard Plant - m.9 .; i RHR Wetwell Spray B 3.6.2.4 RAlft feontinued) LCO Two subsystems of the RHR Wetwell Spray systes must be > OPERABLE. Each subsystem consists of one pump, one heat exchanger, and the necessary piping and valves to circulate water from the suppression pool to the wetwell spray sparger to reduce wetwell airspace pressure. The LCO assures that redundant RHR Wetwell Spray subsystems are available to perform the heat removal function. APPLICABILITY The RHR Wetwell Spray function is only requi-ed to be available in M00ES 1. I and 3 consistent with the requirement for when the Primary Containment must be OPERABLE. In MODES 4 and 5 there is insufficient energy in the reactor to overheat the - wetwell even if suppression pool cooling and wetwell spray were not available. ACTIONS Ad When one RHR Wetwell Spray subsystem is inoperable the remaining OPERABLE subsystem is adequate to perfom the wetwell
- spray function. However, the overall reliability is reduced
,'d because a single failure in the remaining subsystem could result in r.0 wetwell spray capability. As noted, the provisions of LCO 3.0.4 are not applicable. Continued opera-tion is permitted for a limited time due to the low probability of an event which would require wetwell sprays be OPERABLE.
R.I. R.2 With two RHR Wetwell Spray subsystems inoperable, no wetwell spraycapabilityremainstoyndleeitheraccidentortransient events. Continued operation is . justified because of the low probability of an event which would require wetwell spray. However, continued operation of the plant is pomitted for only a limited period of time. The initial inoperable subsystem is also required to be restored te OPERABLE status consistent with Required Action A.I. C.1. C.2 If the Required Actions and associated Completion Times of Condition A or B are not met, the reactor is required to be in MODE 3 and subsequently MODE 4 where the RHR Wetwell Spray system is not required. If the reactor cannot attain MODE 4 (continued) ABWR B 3.6 39 5/31/89 i O l l 16A.9-39 Amendment 9
MN 21A6100AP w
- Samadard Plant -
y ; O ,
- , RHR Wetwell Spray B 3.6.2.4 4 RASES fenntinued) i ACTIONS because of the inoperable RHR subsystems, the reactor coolant (continued) temperature should be maintained as low as practicable using an alternate decay heat removal method.
! Comolation Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available - systems and the time required to reasonably complete the Required Action. SR 3.6.2.4.1 SURVEILLANCE REQUIREMENTS Successful initiation of wetwell spray requires the operation of certain valves. Each manual, autocatic or power operated valve not locked, sealed, or otherwise secured in its correct position is checked to verify that it can be aligned to the position required to operate wetwell spray. SR 3.6.2.4.2 The components.cf RHR Suppression Pool Cooling System are designed to be tested during nomal plant operation. Each subsystem is tested to verify that the required flow of 4200 - gpm (Ref.1) through the associated heat exchanger (and common spray sparger) is achieved. The'B and C subsystems are tested in the combined Suppression Pool Cooling /Wetwell Spray Mode to i verify proper functioning with flow to the common wetwell spr6y t' sparger. Each subsystem is tested to ensure that both will provide the requisite flow. Surveillance Freauencies ~ In general, surveillance fre)quencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and a itkelihood of a change in the system / component status, i REFERENCES 1. A8WR SSAR, Section 6.2.
- 2. NEDO 31466, " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
l ABWR B 3.6 40 5/31/89 e 1 Amendment 9 16A.9 40 l
NN Remndard Piard
- m 61ooAP 1 d
,o 16A.93 B343 Hydrogen Control
,p 16M3.1 R343.1 Hydrogen Recombloer system l Nydrogen Receabiner System B 3.6.3.1 l1 B 3.6 CONTAINMENT SYSTEMS B 3.6.3.1 Hydroenn Racambiner 1rsten
'RASfs (Later)-
1 s
'r
)
t 3 ABWR B 3.6 41 5/31/89 Amendment 9 16A.941
l MM 21A6100AP i 2 Remndard Plant a- i 16A.93.2 B34.23 Pdsamry Costalanwat Oxygen Concentration O-Primary Containment Oxygen Concentration B 3.6.3.2 8 3.6 CONTAINMENT SYSTEMS B 3.6.3.2 primary Containment Orveen Concentration . Ratti , BACKGROUND All nuclear reactors must be des 10ned to withstand events which generate hydrpgen either due to the zirconium metal water reaction in the core or due to radiolysis. The primary method , to control hydrogen is to inert the primary containment. With
,a the primary containment combustible mixture cannot inert be(oxygen present concentration < 4%)for in the containment any hydrogen concentration. The capability to inert the primary works togethercontainment with the and Hyd maintain n Recombiner oxygenSystembelowLCO41 by(volume 3.6.3.1) to mitigate events ich produce hydrogen. For example, en event which rapidly generates Nrdrogen from zir.
conium metal water reaction will result in an increase in the concentration of hydrogen in the containment but oxygen concen- , tration will remain below 45 and no combustion can occur. Long term generation of both hydrogen and oxygen from radiolytic decomposition of tester may eventually result in a combustible
- mixtur6 in containment, except that the hydrogen recombiners remove hydrogen and oxygen gases faster than they can be produced from radiolysis and again no combustion can occur. .
APPLICABLE The Reference 1 calculations assume that the primary contain. SAFETY ment is inerted when the design basis LOCA occurs. Thus, the ANALYS[5 hydrogen assumed to be released to the containment as a result of metal water reaction in the reactor core will not produce combustible gas mixtures in the containment. Oxygen which is subsequently generated by radiolytte decomposition of water is recombined by the Hydrogen Recombiners (LCO 3.6.3.1) more rapidlythanitisproduced) The Oxygen Concentration that ensures.the containment is inert satisfies the requirements of Selection Criteria 2 of the NRC Interim Policy Statement on Technical Specification Improve-ments as documented in Reference 2. LCO The primary containment oxygen concentration is maintained belcw 4% by volume to ensure that an event which produces any amount of hydrogen does not result in a combustible mixture inside primary containment. (continued) i
- ABWR B 3.6 42 5/31/89 16A,942 Anwedment 9
u MM standard Plant - 23A6100AP ha A U , Primary Containment Oxygen Concentration B 3.6.3.2 BASES ftentinued)
-AFPLICABILITY The primary containment must be inert only in MODE 1 since during these conditions the highest probability for an event which could produce hydrogen exists.
Inerting primary containment is an operational problem because it constrains containment access. Therefore, the primary containneent is inerted as late as possible in the plant startup and is deinerted as soon as possible in the plant shutdown. As long as reactor power is below 15% the potential'for an event which generates significant hydrogen is low and the containment need not be inert. Furthermore, the probability of an event - which generates hydrogen occurring within the first 72 hours of-a startup or within the last 72 hours before a shutdown is low enough that these ' windows' with the primary containment not inerted are also justified. ACTIONS A.1. R.1 If oxygen concentration exceeds 4% by volume at any time while operating in MODE 1 with the exception of the relaxations
.[s). allowed curing startup and shutdown, oxygen concentration must be restored to below 4%. A short time (24 hours) is allowed 4 U with oxygen concentration above 4% due to other hydrogen mitigating systems (e.g. hydrogen recombiners) and due to the low probability and long duration of an event which would generate significant amounts of hydrogen. If it cannot be restored within the required Completion Time, reactor power is required to be reduced to s 15% of RTP where the potential for the generation of hydrogen in the primary containment is
! reduced, f; comelation Times ) All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. 1' SURVEILLANCE SR 3.6.3.2.1 REQUIREMENTS The primary containment must be determined to be inert by verifying oxygen concentration is below 4% by volume. (continued) L ABWR B 3.6 43 5/31/89 16A.9-43 Amensment 9
- sanoorr ABWR '
m,a Standard Plant P l Primary Containment Oxygen Concentration B 3.6.3.2 RAtt$ (continued) 4 $URVEILLANCE turve111ance Freauencies ' '~ . REQUIPEMENTS F (continued) In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to oerform the test, the ease of
. perfoming the test and a itkelihood of a change in the systeW +
component status. l' REFERENCES 1. ABWR $$AR, Section 6.2.
- 2. NED0 31466,
- Technical Specification Screening Criteria Application and Risk Assessment " November 1987. 1 l
t r; I
~
l ! O 4 i S I } I l c 4 !~ i 4 ABWR B 3.6 44 S/31/89 l O 16A.941 Amendment 9 i
ABWR 33462.arr na Se==dard Plant - 16A.9A B34A Secondary Comtalement Systemis i 16A.9A.1 B34A.1 Secondary Containment Secondary Containment 8 3.6.4.1 B 3.6 CONTAINMENT $YSTEMS 8 3.6.'4.1 secondary containment RA$t$ BACKGROUND The plant design and analyses postulate that a non tsolatable ' l release of fission products can occur only inside the primary containment. Since the primary containsant is a high pressure leak tight barrier, the bulk of these fission products are expected to remain entrapped within the primary containment and should pose no appreciable risk to the general population as required by Reference 1. As an extra margin of safety to accommodate small quantities of fission products which are postulated to escape from primary containment, a secondary containment has been provided. The secondary containment acts as an additional barrier to the release of fission products through and around primary containment penetrations or through components containing primary system fluid which may be located outside primary l containment. Additionall ..the secondary containment is the < required fission product arrier for plant operations that i
=D occur outside primary enntainment (e.g. handling irredtated t, fuel) or occur when the primary containment need nct be 1 OPERABLE.
-The secondary containment is a structure with a known leakage rate which completely encieses the primary containment and those components which may be postulated to contain primary system fluid. This structuro forms a control volume which serves to dilute the fission products. 11 is possible for the pressure in the control volume to rise relative to the environmental pressure (e.g. due to pump / motor heat load additions). To prevent groupd level exfiltration while allowing the secondary containment to be designed as a conventional structure, the secondary containment requires support systems to maintain secondary containment pressure less than the external pressure. Requiraatnts for these systems are separately specified in LCO 3.6.4.2,. Secondary Containment Isolation Valves, and LCO 3.6.4.3, Standby Gas Treatment System (SGTS).
(continued) ABWR B 3.6 45 5/31/89 Amendment 9 16A.9 45 l l
ABWR m 6ioarr Seandard Plant - ma 0-Secondary Containment B 3.6.4.1 RASt$ fcontinued) BACKGROUND lt has been established (Ref. 3) that the negative pressure , (continued) differential to be unifomly maintained in the secondary containment should be no less than 0.25 inches water when compared to ad,jacent regions (the environment) under all wind conditions up to the wind speed at which diffusion becomes-great enough to assure site boundary exposures less than those of the LOCA even if exflitration were to occur. In this manner the in leakage of fresh air negates the tendency of the fission products to exflitrate through the non leek tight secondary ' containment barrier. APPLICABLE There are two principal events described in Reference I which SAFETY take credit for Secor.dary Containment OPERABILITY. These are: - ANALYSES o Loss of Coolant Accident inside Containment o Fuel Handling Accident The secondary containment performs no active function in response to either of these limiting events, however, its leak tightness function is required to limit the offsite radiation dose as required by Reference 1. Maintaining the Secondary ~ Containment OPERABLE assures that fission products entrapped within the secondary containment structure, will be treated prior to discharge to the environment. Secondary containment satisfies the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical L Specification leprovements as documented in Reference 4.
.(continued)
)
ABWR B 3.6 46 5/31/89 9 Aroendment 9 16A.9-46
. . ~ . _ .- - - . - ~ -
M 2M6100AP Riandard Plant ha 1 I
^ i U l Secondary Containment 5 3.6.4.1 BASES fcontinued)-
LCOs Secondary Containment provides a controlled volume into which fission products which escape primary containment or reactor I coolant pressure boundary components located in secondary l containment can be diluted and processed prior to release to l the environment. For the Secondary Containment to be considered OPERABLE, it must have adequate leak tightness to ensure that the reeutred vacuum can be established and maintained. This OPERABILITY is defined by the Surveillance Reovirements for this LCO. APPLICABILITY Secondary Containment is required for all MODES which could j involve significant releases of fission products into the secondary containment. This includes MODES 1, 7 and 3 at which time the reactor primary coolant may be above 200'F and significant releases of radioactive steam and gas from postulated pipe ruptures or other events is possible. Fission products could escape primary containment and/or escape from those components outside primary containment which contain primary system fiuld. Secondary Containment is also required for other situations under which significant radioactive releases can be postulated such as handling of irradiated fuel
.h' in the Secondary Containment, during CORE ALTERATIONS, and V during operations with a potential for draining the reactor vessel.
ACTIONS L1 The secondary containment is the final barrier to the release of radioactive products from the primary system to the environment. If the Secondary Containment is inoperable in MODE 1, 2, or 3, radioactive releases could be larger than calculated for design basis events. A short time is allowed to
- restore Secondary Containment to OPERABLE status due to the low probability of an event occurring in this time period that would require the Secondary Containment function.
(continued) ABWR B 3.6 47 5/31/89
/unendinent 9 16A.9-47
- .w . . - . - . - - . . . . . . . _ . - . _ - _ . _ _ _ _ . - -- - _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ .
21A6100AP . Standard Plant m- A e O~ i-L Secondary Containment B 3.6.4.1 RASES fenntinued) ACTIONS R.1. B.2 l ' (continued) If the Required Action and associated Completion Time of Condition A are not met (in MODE 1, 2, or 3), then the plant
- conditions are not consistent with those assumed for Itcensing ,
analysis, or plant capability is threatened if a radioactive ' release were to occur. Therefore, the reactor is required to be in MODE 3 and subsequently in MODE 4 to minimite the i-potential consequences of potential events that could occur, c.1. c.f. c.3 l
- Handling irradiated fuel in the secondary containment, CCRE ALTERATIONS and operations with the y tential to drain the reactor vessel can be postulated to cause fission product release to the secondary containment. During refueling operations the secondary containment is the only barrier to release of fission products to the environment. Handling irradiated fuel in the secondary containment and CORE ALTERA-TIONS must be immediately suspended if the secondary contain.
ment is inoperable. Suspension of these activities shall not , preclude the completion of movement of components to a safe, conservative position. Operations that have the potential for draining the reactor ves:e1 must be suspended as soon as practicable to minimize the probability of a vessel draindown and sui, sequent potential.for fission oroduct release. Comolation Times All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. I SURVEILLANCE SR 3.6.4.1.1 REQUIREMENTS Assurance of of Secondary Containment OPERABILITY is provided by verifying that secondary containment vacuum is k 0.25 inches of vacuum water gauge. The secondary containment boundary must be intact or the negative pressure could not be maintained. (continued) ( ABWR B 3.6 48 5/31/89 i. l- Anwadment 9 16A.9-48
-l 1
MM 23A6100AP Remndard Plant ao 1 b i l L Secondary Containeent
- B 3.6.4.1-l RAKES fcontinued)
SURVEILLANCE 1R 3.E.4.1.2. SR 3.8.4.1.3 l REQUIREMENTS ' (continued) Opening of secondary containment hatches, panels, process openings and access ways could permit the infiltration of l outside air of such a magnitude as to prevent maintaining the desired negative pressure. Exterior environmental disturbances (e g. wind gusts) may then permit portions of the secondary containment to expertence a positive pressure relative to the environment. In this case fission product enfiltration could be postulated thus negating the secondary containment safety-function. Permanently closing or monttoring frequently all such openings provides adequate assurance that exftltration from the secondary containment will not occur. 1R 3.6.4.1.4. SR 3.6.4.1.E The SGTS exhausts the secondary containment free volume through appropriate treatment equipment to the environment. To assure that all untreated fission products are treated, these tests ' demonstrate that one SGTS will rapidly establish and maintain a pressure in the secondary containment less than the lowest postulated pressure external to the secondary containment f' boundary. This cannot be tecompitshed rapidly if the seconde y i containment boundary is not intact. Therefore, these tests are used to ensure secondary containment boundary integrity. Since these SRs are Secondary Containment tests, they need not be
- performed with each SGTS subsystems. However, the SGTS subsystems are tested on a STAGGERED TEST BA$l$ to ensure that in addition to the requirements of LCO 3.6.4.3 either SGTS will perfom this test.
Surveillance Freauencies Except as noted, survctllancq frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the ease of performing the test and Itkelihood of a change in the system / component status. (continued) ABWR B 3.6 49 5/31/89 O Amendment 9 16A.9-49
4 MM 23A6100AP .. .Standmed Plant nav. A L- , O: Secondary Containment B 3.6.4.1 Ralfi icontinued) REFERENCES 1. 10 CFR 100. 1 2. ABWR 55AR, Section 6 and 15. 1 1 3. Standard Review Plan 6.2.3.
- 4. NED0 31466, ' Technical Specificattun Screening Criteria Application and Risk Assessment,' November 1987.
0
)
ABWR B 3.6 50 5/31/89 O Amendment 9 16A.9-50
-.M M 23A6100AP Remndard Plant a. a 16A.9.4.2 B34.4.2 Secondary Costalament !aolation Valves Secondary Containment Isolation Valves 1 B 3.6.4.2 I B 3.6 C0hTAINMENT SYSTEMS I
B 3.6.4.2 secondarr cantainment Isolation valves 1 S&!iLL BACKGROUND The function of the Secondary Containment Isoletion Yalves. in combination with other accident mitigation systems. is to Itatt fission product release during and followirg a postulated 1' design basis accident to values less than leakage rates which would result in offstte doses greater than those set forth in Reference 1.- This includes ventilation system automatic isolation dampers as well as other valves designated in the Surveillance Requirements. Secondary Containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumotions used in Reference 2. l' APPLICABLE .The analytical methods and assumptions used in evaluating the SAFETY radiological consequences of design basis accidents are ANALYSES presented in Reference 2. The maximum isolation times for g . Secondary Containment Isolation Valves are the times used in
'- the accident analyses.
The Secondary Containment Isolation valves satisfy the requirements of Selection Criterion 3 of the NRC Interim Policy Statement on Technical Specification leprovements as documented in Reference 4. LCO The Secondary Containment Is914 tion Valves shall be OPERABLE with isolation times less than or equal to the times in l Reference 2. The analytical valve closure times were chosen to (- assure that 10 CFR 100 exposure limits are not exceeded for ! postulated events considered in Reference 2. Closure times are l selected to minimize the release of containment atmosphere to , the environs, thereby mitigating the offstte radiological l consequences. (continued) ABWR B 3.6 51 6/31/89 , . t3 V
- Amendment 9 16A.9-51 l
MM 23A6190AP Rtandard Plant m-2
- O ,
4 Secondary Containment Isolation Valves 7 lc 5 3.6.4.2 t RAEf1 (continued) APPLICABILITf The secondary Containment Isolation Valve function is required for all MODES and other specified conditions which could i involve significant releases of fission products into the secondary containment. This includes MODES 1, I and 3 at which time the reactor coolant may be above 200'F and significant releases of radioactive steam and gas from postulated pipe ruptures or other events are possible, fission gasses could bypass primary containment and/or escape from those components outs ee primary containment which contain primary system fluid. Secondary Containment Isolation Valves are also required to be OPERABLE for other situations where radioactive gas release can be postulated including handling of irradiated fuel in the . L ~ Secondary Containment CORE ALTERATIONS, and during operations with a potential for draining the reactor vessel. ACTIONS M i With one or more isolation valves inoperable, a remaining OPERABLE isolation valve in an affected penetration is adequate to perform the intended isolation function. However, the overall reliability is reduced because a single failure in the - remaining OPERABLE isolation valve could result in the loss of
- the isolation function for that penetration. If the affected penetration has two isolation valves, it must be verified that the other valve is OPERABLE. This action is satisfied by examining logs or other information, to determine if the valve is out of service for maintenance or other reasons. It does not mean to pert' ors the surveillance requirements needed to ~
demonstrate OPERABILITY of the valve. If the effected penetration is designed with only one isolation valve, the penetration is such that plant safety is not unduly threater.ed byatemporaryinabilitytojsolatethepenetration. A.2.1 A.2.2.1. A.2.2.2 in addition to Required Action A.), the inoperable valve (s) must be restored to OPERABLE status, or the line must be manually isolated. Acceptable isolation methods are stated in the Required Actions. Finally, the penetration must be periodically verified to be iso'ested. l (continued) ABWR B 3.6-52 5/31/89 O Amendown: 9 16A.9-52
-ABWR -
w ioarr Shndard PInnt ui
-g 1
I l Secondary Containment Isolation Valves B 3.6.4.2 RA$ct fenntinued) ACTIONS LL,11 (continued) If the Required Actions and associated Completion Times'of l Condition A are not met in MODE 1, 2 or 3. the isolation valves may not be capable of performing their intended function and the reactor is required to be in MODE 3 and subsequently in MODE 4. c.L c.f. c.3 If the Required Actions and associated Completion Times of Condition A are not met when handling irradiated fuel in the Secondary Containment or during CORE ALTERATIONS these opera. tions must be ismediately suspended, suspension of these activities shall not preclude the completion of the movement of a component to a safe, conservative position.- If fuel is being handled in the secondary containment while the reactor is in MODE 1. 2. or 3 then Conditions B anc C are concurrently applicable. If the reactor is shutdown and operations are being performed which have a potential for draining the reactor vessel, they must be suspended as soon as practicable to minimize the probability of a vessel draindown and subsequent I potential for fission product release. Camoletion Times I All Completion Times are based on industry accepted practice l and erfnsering judgement considering the number of available syster. 4M the time required to reasonably complete the Required Action. SURVEILLANCE SR 3.6.4.2.1 l ) REQUIREMENTS Not all secondary containment penetrations are designed to be closed by automatic isolation dampers / valves even though they must be closed during accident conditions. Therefore, these penetrations must be closed at all times during normal operation. This surveillance verifies that those penetrations are closed by blind flanges or rupture disks. (continued) 1 ABWR B 3.6 53 5/31/89 O _ 16A.9-53 Amendment 9 1 - .- - - . - - - _ _ - - - - ...
21A6100AP
- had=rd Plant , , ,
( Secondary Containment isolation Valves B 3.6.4.2 RicfE fenntinued) suRVEILLAhtE - sk 2.s.d.f.2 REQUIREMENTS (continued) This Surveillance Requirement demonstrates each Secondary p Containment Isolation Velve is OPERABLE by verifying isolation time to be within limits when tested pursuant to SR 3.0.5. SR 3.0.5 requires periodic inservice testing of valves in accord-ance with Reference 3. Section IW 3413 of Reference 3 requires that stroke times be specified and periodically esasured. This Surveillance Requirement provides assurance , l l that the closure 11ee for isolation valves will not escoed that . used in Chapter 15 analyses so that predicted radiological
- f. consequences will not be exceeded during the postulated events.
1R 2.6.4.1.3 , This Surveillance Requirement provides assurance that each Secondary Containment isolation valve is OPERABLE and will assume the proper position following containment isolation signals. M ance Frenuancies In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perform the test, the esse of performing the test and a likelihood of a change in the syste# component status. REFERENCES 1. 10 CFR 100 l
- 2. ABWR $$AR, Section 6 and 15.
- 3. )ASNE Section 11, IW 3000
- 4. NEDO-31466, ' Technical Specification Screening Criteria Application and Risk Assessment,' November 1987.
I' I ( ABWR B 3.6 54 5/31/89 \ \ O Amendment 9 16A.9-54
-----aw- m g- i--- , w.---+- 9w-.--w - . -
. --.-----,w.e - , - + y.7 - + . ,,, e. ,..,e- me -%., w. g g us, a wmm.3- wa
l M 23A6100AP
- m. a
.Stasadard Plant i 16A.9.4.3 B34.43 Standby Gas Trentinent System
)
L/] V J Standby Gas Treatment - l B 3.6.4.3 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.3 Standby ans Treatment system BASfS . , BACKGROUND The function of the Standby Gas Treatment System (SGTS) is to maintain the secondary containment at a negative pressure with respect to the environment following the design basis accident and process gaseous releases to limit the thyroid dose and the whole body dose at the plant site boundary within the guideltnes of Reference 1. The SGTS consists of the following components:
'A. One,100 percent capacity charcoal filter train, consist-ing of (components listed in order of air flow direction):
- 1. Prefilter
- 2. High efficiency particulate air (HEPA) filter
- 3. Charcoal adsorber 4 HEPA filter B. Two fully redundant subsystems, each with its own ductwork, dampers, and controls, and consisting of:
I'
- l. Demister
- 2. Electric heater
- 3. Centrifugal fan with inlet flow control vanes.
The sizing of the SGTS equipment and components is based on the design of the reactor building structure and the assumed leakage of the secondary containment. The internal pressure of the secondary containment is maintained at a negative pressure of 0.25 inches water gauge (per Ref. 5) when the system is in operation, which represents the internal pressure required to ensure zero exfiltration of air from the building when a single nonisolated line as large as four inches or all nonseismically qualified lines two inches and smaller are open. The demister is provided to remove entrained water in the air, while the electric heater reduces the relative humidity of the air stream to less than 70 percent (Ref. 3). The prefilter removes large particulate matter, while the HEPA filter is provided to remove fine particulate matter and protect the charcoal from fouling. The charcoal adsorber removes elemental todine and organic todides, and the final HEPA filter is provided to collect any carbon fines exhausted from the charcoal adsorber. (continued) ABWR B 3.6 55 5/31/89 O Amendment 9 16A.9-55
MM 21A6100AP j Standard Plant , w O1 Standby Gas Treatment B 3.6.4.3 BASES fcentinued) BACKGROUND The SGTS automatically starts and operates in response to (continued) actuation signals indicative of conditions that could require operation of the s; stem. Following initiation, both SGTS subsystem fans start. SGTS flows are controlled by modulating inlet vanes installed on the fans and two position volume control dampers installed in branch ducts to individual regions of the secondary contain. ment. The system operates at full fan capacity until the SGTS boundary region is drawn down to the design negative pressure of 2 0.25 inch vacuum water gauge. At this time, motor-operated dampers automatically throttle the exhaust flow so that the volume of air being exhausted is equal to the volume of air infiltrated at the design negative pressure. i APPLICABLE The design basis for the SGTS as presented in FSAR Chapter 15, SAFETY Accident Analysis, is to mitigate the consequences of the ANALYSIS Loss of Coolant Accident (LOCA) and the Fuel Handling Acci. dents. For all events analyzed the SGTS is shown to be auto-matically initiated to limit, via filtration and adsorption, the cite boundary radioactivity dose level to well within the Reference 1 guideliries. The SGTS satisfies the requirements of Selection Criterion 3 of ' the NRC Interim Policy Statement on Technical Specification Improvements as documented in Reference 6. j I LCO Both SGTS subsystems and the common, passive filter train must l be OPERABLE to assure at least one subsystem is available in ! the event of a single active failure, to meet its safety function following a LOCA or fuel handling accident. A description of what is required for the SGTS to be considered i OPERABLE is provided in the Background section. (continued) r ABWR B 3.6-56 5/31/89 O Amendment 9 16A.9-56
MM 23A6100AP i Remndard Plant n.u , A k) J Standby Gas Treatment B 3.6.4.3 i RAtri (continued) APPLICABILITY .Both SGTS subsystems and the filter train shall be OPERABLE during MODES 1. 2. 3 and when irradiated fuel is being handled, during CORE ALTERATIONS and during operations with a potential for draining the reactor vessel to prevent a potential source of redioactit.ity from escaping from the secondary containment. ? resulting in exceeding the offsite dose limits. This applic-ability is consistent with the requiremsnt for when the second. ary containment must be OPERABLE. ACTIONS M With one SGTS subsystem inoperable, the remaining OPERABLE subsystem is adequate to perform the required radioactivity release control function. However, the overall plant reliabil-ity is reduced because a single failure in the remaining subsystem could result in a total loss of the SGTS function. Therefore, the inoperable SGTS subsystem must be restored to OPERABLE status and continued operation with one subsystem inoperable is only allowed for a short time.
.A B.1. B.2. c.1 If both SGTS subsystems, or the common filter train, are inoperable in MODE 1. 2 or 3 the SGTS will not be capable of performing its intended function. The common filter train or
, one inoperable subsystem (which includes the connon filter train) must be restored to OPERABLE status within 4 hours. Continued operation is justified because of the low probability of an event which would require SGTS operation. The initial , inoperable subsystem must also be restored within 7 days from its initial discovery consistent with Required Action A.I. D.1. D.2 If the Required Actions and Associated Completion Times of Condition A or B are not met in MODE 1. 2 or 3. the reactor is required to be in MODE 3 and subsequently in MODE 4 where the probability of requiring operation of the-SGTS is reduced. (continued) { ABWR B 3.6 57 5/31/89 O Amendment 9 16A.9-57 p l
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23A6100AP Senadard Plant m O' F Standby Gas Treatment. B 3.6.4.3. RASES fcontinued) . ACTIONS f.1. E.2. E.3 8 (continued) If the Required Actions and associated Completion Times of Condition A or B are not met when handling irradiated fuel in the Secondary Containment, dering CORE ALTERATIONS, or durin operations with a potential for draining the reactor vessel,g these actions are suspended to minimize the possibility of a radioactive release as a consequence of a fuel handling acci-dont or inadvertent vessel draindown which uncovers the core. Suspension of these activities shall not preclude completion of the component to a safe, conservative position. if fuel is being handled while the reactor is in MODE 1, 2 or 3, then Conditions D and E are concurrently applicable. Comoletion Timeg All Completion Times are based on industry accepted practice and engineering judgement considering the number of available systems and the time required to reasonably complete the Required Action. SURVEILLANCE REQUIREMENTS. SR 3.6.4.3.1 Standby systems should be checked periodically to ensure they G. , will start and function properly. This surveillance require- ! ment ensures each subsystem will start on demand and continue [ to operate. Operation with the heaters on (automatic heater cycling to maintain temperature fulfills this requirement) for ! greater than or equal to 10 hours every 31 days reduces the l buildup of moisture on the adsorbers and HEPA filters. (continued) i i L ABWR B 3.6-58 5/31/89 l O Amendment 9 16A.9-58
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23A6100AP . Standard Plant ui-Standby Gas Treatment B 3.6.4.3 RAit$ (continued) SURVEILLANCE SR 3.6.4.3.2. SR 3.6.4.3.3. SR 3.6.4.3.4. SR 3.6.4.3.5 REQUIREMENTS-(continued) These surveillance requirements demonstrate that the designed ,* filtration capability of the system is maintained by verifying that the system flow HEPA filters and charcoal adsorbers satisfy the in place testing acceptance criteria, surveillance. - intervals and procedures or Reference 3 and ANSI N510 1975. . The laboratory analysis of a representative carbon sample (SR 3.6.4.3.4) must be performed in accordance with the testing criteria of Regulatory Position C.6.a of Reference 3. ' The i carbon sample to be used in this test must be obtained in accordance with Regulatory Position C.6.b of Reference 3. The in place acceptence criteria for SR 3.6.4.3.2 and SR 3.6.4.3.3 are defined in Regulatory Position C.5 of Reference 3. The system flow rate is verified during subsystem operation, for SR 3.6.4.3.2, SR 3.6.4.3.3 or SR 3.6.4.3.5 when tested in accordance with ANSI N510 1975 (Ref. 2). In addition to the 18 month surveillance intervals, SR 3.6.4.3.2 SR 3.6.4.3.3 and SR 3.6.4.3.4 are required whenever modifications or events which may affect the integrity of the HEPA filters or charcoal adsorbers have occurred. For the purpose of this specification, ' filter service area' means the area where painting, fire or chemical release occurred and from
] which an operating fan is taking suction, i
SR 3.6.4.3.6 The SGTS filter train and isolation damper OPERABILITY must be demonstrated to ensure that for the LOCA and fuel handling accident events, the licensing basis assumptions remain valid. A test actuation signal is used to test each SGTS subsystem and to verify that these system components will function when required. SR 3.6.4.3.7 The electric heaters in each filter train are provided and sized to reduce the humidity of the treated air from 100% to 70%. The incoming 1200 cfm of air at a temperature of [ J'F (Ref. 4) and 100% relative humidity requires [ ] zw (Ref. 5) of heat to lower its relative humidity to 70%. The test will be performed in accordance with ANSI N510 1975. (continued) ABWR B 3.6 59 5/31/89 1 Amendment 9 16A.9-59
' MM 23A6100AP Standard Plant Rm.A ; O Standby Gas Treatment 8 3.6.4.3 4 BA$fS feontinued) SURVEILLANCE Surveillance Frecuencies REQUIREMENTS (continued) In general, surveillance frequencies are based on industry accepted practice and engineering judgement considering the l unit conditions required to perform the test, the ease of- 'l performing the test and a Itkelihood of a change in the j system / component status. l REFERENCES 1. 10 CFR 100. Reactor Site Criteria.
- 2. ANSI /ASME N510 1975, Testing of Nuclear Air Cleaning Systems.
- 3. USNRC Regulatory Guide 1.52, Rev. 2.
- 4. ANSI /ASME N509 1976, Nuclear Power Plant Air Cleaning Units and Components.
- 5. Standard Review Plan, 6.5.3.
- 6. NEDO-31466, ' Technical Specification Screening Criteria Application and Risk Assessment," November 1987. -
ABWR B 3.6 60 5/31/89 O Amendment 9 16A.9-60
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MM 23A6100AP Reamda d Plaat Rev.A APPENDIX 16A.10 CONTENTS Section Dt]t Ragt i 16A.10.1 R311 Ranciar hallding Camilag Watar/Ri ndet I Rainding Servier Water Swatam Operati.4 16A.101 , i 16A.10.2 R312 Raneter Building Camilag Water (RCW1/Ranctor ! Baunat.Estdst B' star (REW1 Svatam Shutdown 16A.10-8 16A.10.3 R3.7.3 Central Raam HVAC Ermarpean Racirculation f
$3alast 16A.1011 ,
i 16A.10,4 R3.7.4 Main Condemner Oftpas 16A.1016 j i I l I h ( i t 6 I 16A.10-li t ( Ar.iendment 9
MM R*= 2M610MP
'and Plaut m. a l 164.10 B3.7 PIANT SYSTEMS l o
- 164.10.1 R3.7.1 Remeter Buildlag Coollag Water /Renesor Rallding Sevice Water SysteenWrating i
i N I RCW/R$W $ystem
- Operatina B3.7.1 B.3.7 PLANT $Y$i[M$
$ 3.7.1 kaarter kuildine toeline Wathr/kaattor Buildina Servite Water igg em Onoratino BA$tt BACKGRDUND The Reactor Building Cooling Water (RCW) $ystem and Reactor ,
Building Service Water (RCW) System together are designed to l provide cooling water for the removal of heat from plant 1 auxiliaries, such as Residual Heat Removal (RHR) system i mo;or-end pump seal coolers, heat exchangers, standby diesel 1 enerators, room coolers for Emergency Core Cooling System : [CCS) equipment etc., required for a safe reactor shutdown , c110 wine a design basis transient or accident. The RCW/R$W : system also provides cooling to plant components, as 1 required, during normal eperation and reactor shutdown < modes. During an accident, most equipment required for riormal operation only is isolated from the RCW system; cooling is directed only to safety related equipment, the control rod drive pumps, and the instrument air compressors. The operator may, at his discretion, isolate the control rod drive pump and the instrument air compressor coolers, for the purpose of this technical specification the combined ( RCW/R$W system consists of the [ ultimate heat sink (UH5)), three independent RCW and R$W cooling water headers (Divi-sions 1. II. Jnd 111) and their associated pumps, piping, valves and instrumentation. These divisions cool equipment - in Division 1. 11. and 111 respectively. These divisions are mechanically tnd electrically separated from each other so that failure of one division will not affect the - OPERABILITY of the other divisions. The RCW and R$W systems operate tcyther to provide the cooling water required by the RCH system loads. Operability of each RCW division is dependent uptn the operability of its corresponding R5W division, Ther4fo e an inoperable R$W subsystem renders the as!.ociated RCW/R$W division inoperable, and ultia.ately all components coeled by that division must te declared inoperable. Coolingwaterispumpedfromthe[ UHS)bythethreeReactor Building Service Water (R$W) Divisions to the RCW heat exchangers through the three main redundant suppl After removing heat from the p.CW heat exchangers,y headers. the water is discharged to the [UH$).
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(continued) ABWR B 3.7 1 5/31/89
/~
( Amendment 9 16A.1051
MM 23A6100AP a-a RemmA=md Plant e i RCW/R$W $ystem . Operating 8 3.7.1 BA5(5 f BACKGROUND The three RCW divisions supply cooling water to redundant (continued) equipeent required for a safe reactor shutdown. The spr. ] cific equipment for which the RCW/R$W system supplies t cooling water is listed in Reference 2. Each RCW/R$W division has two RCW pumps, two heat exchangers, and two R$W pumps. One RCW pump, one R$W pump, and one heat exchanger 1 in each division are adequate to support normal operation. ! and are in operation at all times. The standby pump in each division automatically starts and the standby heat exchanger is automatically valved in upon receipt of a LOCA signal. , The RCW/R$W system is designed to withstand a single active i or passive failure (pump seal or valve failure) coincident i with a loss of offsite power without losing the capability l to supply adequate cooling water to equipment required for , safe reactor shutdown. 4 APPLICABL[ following a design basis accident or transient, the RCW/R$W SAffiY system will realign itself and continue to operate auto-matically and without operator action. Mar.ual initiation of ANALYS($ i supported systems, e.g., RHR suppression pool cooling is however, performed for long term cooling operations, i The ability of the RCW/R$W system to support long term ' cooling cf the reactor or cortainment is evaluated in $5AR Chapters 6 (Engintered Safety features), p (Auxiliary Systems) and 16 (Accident Analyses). These analyses expitc-itly assume the RCW/R5W will provide adequate cooling support to the equipment required for safe reactor shutdown. These analyses include the evaluation of the long term containment response after a design basis Loss Of Coolant
- Accident (LOCA). The RCW/R$W system provides cooling water to the RHR heat exchangers for all modes of RHR to limit the suppression pool temperature and containment pressure following a LOCA. This ensures the containment can perform its intended function of limiting the release of radioactive materials to the environment following a LOCA. The RCW/R$W system also provides cooling to other components assumed to '
function duting a LOCA. The safety analyses for long term cooling were performed (Aeferences 4 and 5) for a LOCA concurrent with a loss of offsite power. and minimum available diesel generator power. The worst case single failure which would affect the perfor- ' mance of the RCW/R$W systems is the fatture of one of the ' three standby diesel generators which would affect one subsystem of the RCW/R5W system. (continued) ABWR B 3,7 2 $/31/89 O Amendment 9 M142
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2M61 UMP Se==dard Plant m t i RCW/R$W $ystem . Operating i
. B 3.7.1 ;
BA$t$ i APPLICABLE RCW/R$W $ystem . Operating satisfies the requirements of ;
$AF(TV Selection Criterion 3 of the NRC Interim Policy $tatement on ANALY$l$ Technical $pecification Improvements 45 documented in 7 (continued) Reference B.
( LCOs The OPERABILITY of the RCW/R$W system is required to ensure the effective operation of the RHR system in removing heat from the reactor or containment and the effective operation of other safety related equipment during a design basis
- accident or transient. The OPERABILITY of each independent ,
division of the RCW/R$W system is based on 1) having an - OPERABL[ [ UHS), 2) all pumps and heat exchangers in all i RCW/R$W divisions OPERABLE, 3) OPERABLE RCW flow paths to all components required to be OP[RABLE, 4 i OPERABLI R$W flow i pathscapableoftakingsuctionfromthe1.UH$)andtransfer. ' ring the water to the RCW heat exchangers, as required, and
- 5) a minimum water level in the pump well of the intake structure.
e Requiring all divisions OPIRABLE assures adequata capability .
/ to meet cooling requirements of the equipment required for !
( safe shutdown. i The OPERABILITY of the [ UHS) is based on having greater than l the minimum required water level and less than the maximum allowed water temperature. APPLICABIL11Y The requirements for OP[RABILITY of the RCW/R$W system in MODE $ 1, 2, and 3 are governed by the required OPERABILITY of the [CCS (LCO 3.5.1), the diesel generators (LCO 3.B.1) and the applicable Containment and Decay Heat removal modes of the RHR system (LCOs 3.4.5, 3.6.2.3 and 3.6.2.4). RCW/R$W system requirementr, for other operating modes are covered in LCO 3.7.2. ACTIONS A.I. A.2 With one RCW pump and/or one R$W pump and/or one RCW/R$W heat exchanger in any division inoperable, that division ($ still operable to provide cooling water to all its served components. One RCW pump, one R$W pump, and one heat exchanger in each division is adequate to meet all design criteria except for long ters containment cooling. As such, an extended time is allowed for a single RCW and/or R$W pump (continued)
- ABWR B 3.7 3 $/31/89 l
O Amendment 9 1M.10 3 L _ , _ _ _ _ _ __ _
~
M 21A6100AP te==tard Plant m .A I O' RCW/R$W System . Operating B 3.7.1 BASES e ACT]DNS and/or heat exchanger to be out of service, in any of the (continued) divisions, provided the RCW/R$W subsystems are otherwise OPIRABLE. As noted, each Division may be in this condition concurrently. : t if the inoperable component (s can not be restored to ' OPLRABLt status within the l al) owed Completion Time, then j Condition D applies. Alternatively, the affected system or ; compor*ents (i.e. those requiring all RCW/RSW pumps and beat ; exchangers in a division to be OPERABLt) may be declared r inoperable and the appropriate actions taken. i. If two RCW pumps, two R$W pumps or two heat exchangers are out of service in one or more divisions, then Action B, C or 0 applies. Secause of the level of redundancy in the RCW/R$W system ! design, the provisions of LCO 3.0.4 are not applicable, and ; Modes 1, 2, or 3 may be entered while in Condition A. ' B.l. B.f. C.I. C.f.1. C.f.2 ' The loss of both RCW pumps, both RSW pumps, or both heat ' exchangers in the same Division would render that division completely inoperable and the associated equipment cannot perform the intended function. With any RCW/R$W flow path ; inoperable (e.g. velve failure, flow blockage) the system (s) z served (cooled able; however,)the RCW/R$W division may still be able toby that RCW/RSW d supply cooling to all other associated systems. Other types of f ailures, such as the f ailure of an isolation valve (which would isolate non essential components in case of a LOCA) may reduce the capability of the division to less than its design value for the designated essential components. In th'sse cases, the division (s) is considered inoperable and cont,nued operation is only allowed for a short perto1. V,th one division inoperable, for reasons other than Cor.dt. sion A, there is still sufficient safety related cooling water available to assure safe shutdown capability, even with an additional single failure. However, because overall , system redundancy is reduced, continued operation is only allowed for seven days. With two divisions inoperable, the system is still capable of performing its intended function; however, system redundancy is severely degraded so that
- continued operation is only allowed for 24 hours at which time at least one of the inoperable Divisions must be restored to OPERABLE status. Additionally, the initial (continued) l ABWR B 3.7 4 $/31/89 Amendment 9 9
IfA.lM l
~- -- -
L ha-rd Plant 2M610MP m_ i !
- m. ,
RCW/R$W $ystem + Operating B 3.7.1 BA$[$ ' ACTION $ inoperable division must be restored to OP(RABLE status (continued) within seven days. Alternatively, the effected systee(s) or component (s), under any of these conditions, may be declared inoperable and the associated A:tions taken. D.1. D.2 If the Required Actions and associated completion times of Conditions A, B, or C cannot be met, or with all three . RCW/R$W Divisions inoperable, the reactor is required to be in MODE 3 in 12 hours and in MODE 4 in the following 24 hours. In MODE 4 the system requirements are reduced as specified in LCO 3.7.2. If MODE 4 cannot be achieved , becauseoftheinoperableRCW/R$Wdivision(s),thereactor coolant temperature should be maintained as low as practicable using an alternate heat removal method. - Comoletion Times All Completion Times are based on irdustry accepted practice and engineering judgement considering the number of avail. able systems and the time required to reasonably complete the Required Action.
$URV[1LLANC[ SR 3.7.1.1 RIQUIRIM[NTS This surveillance verifies the [ ultimate heat sinkl has sufficient cooling water to satisfy the design basis of 30 day cooling capability with no external makeup source. With the [ ultimate heat sink) inoperable, the affected RCW subsystems must be declared inoperable. [This SR is only
(- applicable when the UHS consists of a dedicated Dody of watersuchasacoolingtowerbasinorpond.) SR 3.7.1.2 Verificationofthe(VHS)temperatureensuresthattheheat removal capability of the RCW/RSW system is within the t i i assumptions of the design basis accides.t analysis, 50 3.7.1.3 ' This surveillance requirement verifies that the water level inside the pump wells of the intake structure is sufficient for proper operation of the R$W pumps (NPSH and pump vortex. ing are considered in deteretning this limit). If the water l (continued) ABWR B 3.7 5 5/31/89 l Amendment 9 IR145 l
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i MM 2M610MP
)
Senadard Plant w i O RCW/R$W Systes . Operatin 3.7. BA$ts
$URyt!LLANCE level is > feet, there is sufficient margin to the REQUIRtMINTS minimum leve[levery ] feet)sothatthesurveil.
re]quirement ([if14 days. (continued) lance can be cerformed However, the level i s[ ] feet, the surveillance should be performed more ' frequently (every 12 hours) to ensure adequate suction ! conditions for the R$W pumps. [This surveillance is not l spplicableforseawaterplants.) l 1R 3.7.1.4 Verification of the correct alignment of all valves is essential to ensure the proper flow paths servicing safety J related systems or components for the RCW/R$W subsystems. l SR 3.7.1,5 ThissurveillarceverifiestheOPERABILITYofthe[ ultimate heat sink active components). The 15 minute duration for i operation is sufficient to monitor the steady state perform. ! ance of the required active components. [This $R is only applicable for plants taking credit for UHS active , componentssuchascoolingtowerfans.) ' lE L I d dt This surveillance verifies the automatic isolation valves of the RCW/RSW system will automatically switch to the safety or emergency position to provide cooling water exclusively to designated equipment during an accident event. This sur-veillance also verifies the automatic start capability of the standby RCW/RSW pumps and [ ultimate heat sink) active components and the automatic inclusion of the standby heat exchanger in the coding loop on both the RCW and RSW sides. Surveillance Frecuencies in general, surveillance Frequencies are based on industry accepted practice and engineering judgement considering the unit conditions required to perfore the test, the ease of performing the test and a likelihood of a change in the systee/ component status. (continued) ABWR 8 3.7 6 $/01/89 O Amendment 9 1M.10 6
i 23A6100AP S*==dard Plant m l 0 i U s t i RCW/R$W System Operatin i
. 3.7. I BAST $ !,
REftR[NCES 1. Regulatory Guide 1.27 ' Ultimate Heat Sink for Nuclear Power Plants'. Para. C.l.
- 2. ABWR $$AR. Table 9.2 4. ,
- 3. Regulatory Guide 1.1 ' Net Positive $uction Head for Emergency Core Cooling and Containment Heat Removal System Pumps'.
- 4. ABWR $5AR. Section 6.2.1.1.3.3.1.4.
- 5. ABWR $$AR. $eciton 6.2.2.3.1.
- 6. NEDO.31466. ' Technical Specification Screening Criteria Application and Risk Assessment *, November 1987.
, k i i f i i i i ABWR B 3.7 7 $f33fgg O - Arnendment 9 16A.10 7 l
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t 21A61 M P R*==dard Plant _ i 1 16A.10.2 83.7.2 Reactor Bulldlag Coollag Water (RCW)/ Reactor Bulldlag Service Water (RSW) i Systesn Sheldews 5 i RCW/R$W System . Shutdown B 3.7.2 B.3.7 PLANT $Y$TEMS 8 3.7.2 Reactor Ru11dino Cae11hu Water fktW)/Reacter Ru11dino Service ' kater (RSW) System thutdown BASf5 ' BisCKGROUND The Reactor Building Cooling Water (RCW)/ Reactor Building Service Water (R5W) $ystem is described in the Bases for i LCO 3.7.1. - APPLICABLE The ability of the RCW/R$W system to support long ters
- SAFE 1Y cooling of the reactor or containment is evaluated in $5AR ANALYSE 5 Chapter 6 (Engineered Safety Features), 9 (Auxiliary Systems) and 15 (Accident Analyses). These snelysis expli.
citly assume that the S$W will provide adequate cooling support to the equipment required for safe rea:tc? shutdon These analyses include the evaluation of the long tem containment response a}}