Text

Proposed Tech Spec Sections 1,3,4,5 & 6,reflecting Numerous Editorial Corrections & Administrative Changes
	ML20155B459
Person / Time
Site:	Rancho Seco
Issue date:	09/30/1988
From:	; SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	;
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ML20155B429	List: ML20155B425; ML20155B459;
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	NUDOCS 8810060310
	Download: ML20155B459 (121)
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RANCHO SECO UNIT 1

, TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Sfta119n EA9ft 1 DEFINITIONS 1-1 1.1 RATED P0HE8 1-1 1.2 REACTOR _0PERATING CONDITIONS 1-1 1.2.1 Cqld Shutdown 1-1 1.2.2 Hot Shutdown 1-1 1

1.2.3 Reactor Critical 1-1 1.2.4 Hot Standby 1-1 j 1.2.5 P_ower Ooer.ation 1-1 l 1.2.6 Refueling Shutdown 1-1 l 1 !

! 1.2.7 Refueling Ooeration 1-2 {

j i j 1.2.8 Refuelina Interv d 1-2 i i 1.2.9 SIAttyp 1-2 1.2.10 Remain Critical 1-2 i 1.2.11 Tayg 1-2 l

1.2.12 Etatso - Coalitw.R 1-2 l f

1.2.13 Action 1-2 >

! i 1.2.14 Leakagt 1-2 l

l ;

l

! 1.3 OPERABLI 1-2a 1.4 PROTECI]ON INSTRUMENTATION LOGIC 1-2a l

1

! 1.4.1 Instrument Ch.annel 1-2a l

1.4.2 8tictor Protection Syltem 1-2a l 1.4.3 Ergitetion ChAnati 1-3 1.4.4 Etaetor Protertion_SylteL Logic 1-3 l 1.4.5 Safety featutti_Syltellogi.c 1-3 i 1.4.6 Et9tle of Redundancy 1-3 i Proposed ;

1 Amendment No. 28, 9/, 171 I

!i 0010060310 000930 l PDR ADOCK 03000.".1R ,

l P PNU l

i l

RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued)

Section PAgt 1.5 1Hf'RUMENTATION SURVEILLANCE 1-3 1.5.1 Trio Test 1-3 1.5.2 Channel Teit 1-3 1.5.3 Instrument Channel Check 1-3 1.5.4 Instrument Channel Calibratina 1-4

' l 1.5.5 Heat Balance Check 1-4 1.5.6 Heat Balance Calfbration 1-4 1.5.7 Source Cheth 1-4 l 1.6 OVADRANT POWER _H LI 1-4 1.6.1 Rentor_fower Imbalance 1-4 1.7 CONIAlhhERI_JRIEGRITY 1-4 1.8 LICENSEE EVENT REPQRIS 1-5 l 1.9 TIME PERIODS 1-5 1.9.1 Shift 1.- 5 1.9.2 Q1113 1-5 1.9.3 Meekly 1.5 1.9.4 Fortnight 1v 1-S 1.9.5 Monthly 1-5 1.9.6 Quar _ tar _13 1-5 l 1.9.7 Semiannually 1-5 1.9.8 anGUAlly 1-5 1.9.9 Deleted l

l 1.9.10 Refuel.ing_ Interval _ 1-5 1.10 SAFETY 1-5 1.11 FIRE SUPPRESS 10N_ SYSTEMS 1-6 1.12 SJAGGERED TEST BASIS 1-6 1.13 ER0CISS_CORIRQLEROGRAM 1-6 1.14 S011DIUCALLON 1-6 1.15 0FFSITE DOSE CALCUtAL10H BARVAL(00.CB1 1-7 Proposed Amendment No. 53, 81. 171 11

RANCHO SECO UNIT 1 i TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued)

Section Eagt Leatagg_3nd Leak Detection 3-12 !

3.1.6 3.1.7 Hoderator Temoerature Coefficient of Reactivity 3-15 3.1.8 Low Power Physics Testina Restrletions 3-15b .

t 3.1.9 Control Rod _Ooera1Lon 3-16 3.2 HIGH PRESSURE INJECTION. CHEMICAL ADDITION. AND LQH 3-17 i

l IBEERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEMS 3.3 EMERGERCLCORE C00LINGE10L3 MILD 1HG_ EMERGBC1 3-19 '

l COO.LIRG AND REACTOR BUILDING SPRAY . SYSTEMS 1 3.4 SJEAM AND POWER CONVERSION SYSTEM 3-23 l i

1 3.5 INSTRUMENTATION SYSIBS 3-25 3.5.1 OAer3.tional Safety _lnstrumentation 3-25 3.5.2 Control Rod Grouo and-Power Distr.1hy. tion Limits 3-31 !

3.5.3 Safety Features Actitation System Satooints 3-34 3.5.4 Incore Instrumentation 3-36 3.5.5 A c c i de n t Mon i t o r i ng .J n s t rumen t ation 3-38d j 3.5.6 Emeraency Fegdntef_ Initiation and Control Sets.oinh 3-38g l

i 3.5.7 Erne rg e n cy_ Shuid.own I n s t rume n t a t i on 3-381 ;

a j 3.6 BIAClQR BUILDING 3-39 :

1 3.7 AUXILIARY ELECTRICAL SYSTEMS 3-41 t

3.8 EUELLOAQIXG ANC REIRELING 3-44 l

l 3.9 SEENLfutLE00L 3-46a i i 3.10 3-47 SECOEDARY SYSTEM _ACTIVIlY

{

3.11 REeC101.3UILDING POLAR CRANE AND AUXLLIARY_HOISI 3-49 ;

{ 3.12 StiUMERS 3-51

!; 3,13 AIR FILTER _S151 BS 3-52 1

1

! Proposed j Amendment No. 28, 39, 80, 82, 84, 87, 93, 97, 98, 171

),

111

~ -,w., e , n pnm--.--r-,-~,-----,--------n .

vw--n-- -~-,r- - - - - , - - , ,m,-.-er,-wr-- w -,e r<---- - - - - - - , , -- ,m-re--

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS ;

4 TABLE OF CONTENTS (Continued)

Section Ea91 ,

3.14 FIRE SUPPRESSION 3-53

! i

. 3.14.1 Instrumentation 3-53 ;

3.14.2 Hater System 3-53 [

3.14.3 Soray and Sorinkler Systems 3-56 ,

3.14.4 CO2 System 3-56 3.14.5 Fire Hose Stations 3-56a -

3.14.6 Fire Rated Assemblies 3-58 t

3.15 RADI0 ACTIVE LIOUID EFFLUENT HONITORING INSTRUMENTATION 3-60 3.16 RADIOACUMf_iASEOUS EFFLUENT MONITORING INSTRU]iENTATION 3-63 !

1 3.17 (1QVID EFFLVI!ilS 3-70 I i

j 3.17.1 Concentration 3-70 i 3.17.2 0011 3-71 3.17.3 Ltquid Holdyo Tanks 3-72 i

3.17.4 Liquid Effluent Radwaste Tr" JLt 3-72a 3.18 GASEOUS EFFLUIBIS 3-13 l l

3.18.1 Dose Rait 3-73 l 3.18.2 Qoje-Nob 1_e Gtigi 3-74 t

3.18.3 Qqis_-Indine.131. Iodin.e-133. Tritium and RadioAgilyt 3-75 ;

L4aterials in Pa_rticulate Form 3.18.4 GAitoutRadyAltLIte11gte.1 3.78 3.18.5 CA1 S.tof_ age _Tankt 3-79 3.19 DILEIIQ 3.20 DELEIEQ 3.21 SOLID _3Ajll0 ACTIVE _ HASTES 3-80 3.22 RADIOLOGLCAL_ ENVIRONMENTAL _ M0!(IIORIBG 3-81 3.23 LNfQ_USE.4sb1!!S 3-87 3.24 EXELOSIYE_LAS_BlXIURE 3-89 Proposed Ameadment No. 28, 85, 98, 171 iv i

)

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued)

Section East 3.25 EUEL CYCLE DOSE 3-90 3.26 INTERLABORATORY COMPARISON PROGRAM 3-92 :

3.27 HUCLEAR SERVICE ELECTRICAL BUILDING EMERGENCY HEATIRG 3-93 l I

VENTILATION AND AIR CONDITIONING 3.28 TOI DIESEL GENERATOR CONTROL ROOM ESSENTIAL 3-93a VENTILATION SYSTEM l 3.29 HETEOROLOGICAL MONITORING INSTRUMENTATIQH 3-94 3.30 81DROGEN RECOMBINERS 3-96 4 SURVEILLANCE STANDARDS 4-0 l 4.0 GENERAL SURVEILLANCE REOUIREMENTS 4-0 l 4.1 OPERATIONAL SAFETY REVIEH 4-1 4.2 S])RVEILLANCE OF ASME CODE CLASS 1. 2. AND 3 SYSTEMS 4-10 I 4.2.1 Reactor Vessel Surveillance Soecimens 4-10 !

4.2.2 Inservice Insoection 4-11 4.2.3 Leakage Surveilla011 4-13 f

4.3 TESTING FOLLOHING OPENING Of_ SYSTEM 4-14 !

4.4 REACTOR BUILOJHG 4-15 4.4.1 Containment Leakast_Its11 4-15 4.4.2 Structural Integrity 4-21 4.4.3 Deleted 4.5 EMERGENCLCORE_CQ0 LING _AN D_REACIOJLBU I L DI NG 4-26 CQQLING SYSTEM PERIODIC TESTING 4.5.1 Imttgency Core Coolina Syltam 4-26 4.5.2 ReattoLBuilding Coolina Sy11km1 4-29 l

4.5.3 Dt. cay _1 tat Removal Syltam and Rotttnr_ Building _SarAy Sy11ta_Lt3kast 4-32 i

I Proposed l Amendment No. 28, 76, 85, 94, 95, 97, 98, 171 y

RAkCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued)

Section EAgt l 4.6 EMERGENCY POWER SYSTEM PERIODIC TESTING 4-34 ;

r 4.7 REACTOR CONTROL ROD SYSTEM TESTS 4-36 4.7.1 Control Rod Drive System Functional Tests 4-36 -

4.7.2 Control Rod Program Verification (Group vs. Core Positions) 4-37 i 4.8 AUXILIARY FEEDHATER PUMP PERIODIC TESTING 4-39 4.9 REACTIVITY ANOMALIES 4-40 ;

4.10 CGdTROL ROOM / TECHNICAL SUPPORT CENTER-EMERGENCY FILTERING SYSTEM 4-41 [

r 4.11 REACTOR BUILDING PURGE EXHAUST FILTERING SYSTEM 4-42 4.12 AUXILIARY AND SPENT FUEL BUILDING FILTER SYSTEMS 4-43 4.13 AUGMENTED INSERVICE _1RS E CIl0N PROGRAM FOR HIGH 4-44 ENERGY LINES OUTSIDE 0F CONTAINMENT 4.14 SNUBBERS 4-47 :

4.15 RADIOACTIVE MATERIALS SOURCES 4-48 4.16 Reserved 4-49 4.17 STEAM GENERATORS 4-51 4.17.1 Steam Generator Samn13_ Selection and Inspection 4-51 !

l 4.17.2 Sitam Generator Tube Samole Selection and_Insatt110D 4-51 ;

4.17.3 Inspection Frecuencies 4-52 f 4.17.4 Acceotante Criteria 4-53 t

4.17.5 Reports 4-54 1 4.17.6 OTSG Auxiliary Feedwater Header Surveillance 4-54 4.17.7 Inigerilon_AgCcotance Criteria and Corrective Actions 4-55 4.17.8 Reports 4-55 4.18 FIRE TUPPRESSION SYSTEM SURVEILLANCE 4-58 Proposed Amendment No. 28, 39, 66, 77, 97, 58, 171 vi

i v i l

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS l i TABLE OF CONTENTS (Continued) l Section EA91 ,

4.19 RADIOACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION 4-63 l

4.20 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 4-65 4.21 LIOUID EFFLUENTS 4-69 -

3 4.21.1 Concentration 4-69 4.21.2 D.olti 4-72 4.21.3 Liauid Holdup _ Tanks 4-73 l j 4.21.4 Li.quij Effluent Radwaste Treatment 4-73a 4.22 GASEOUS EFFLUENTS 4-74 4.22.1 Dose Rate 4-74

. 4.22.2 Dose-Noble Gases 4-77 f

} 4.22.3 Dose-Iodine-131. Iodine-133. Tritium. and Radioactive 4-78 Materials in Particulate Form i

4.22.4 Gaseous Radwaste Treatment 4-79 ,

1 4.22.5 Gas Storage Tanks 4-80

)

j 4.23 DELETED l 4.24 DELETEQ l

1 4.25 SOLID RADIOACTIVE WASTES 4-81 ,

4.26 RADIOLOGICAL L4VIRONMENTAL M04110 RING 4-83

, 4.27 LAND USE CENSUS 4-86 l 4.28 EXPLOSIVE GAS MIXTURE 4-87 f I

! 4.29 EUEL CYCLE DOSE 4-89 f 4,30 INTERLABORATORY COMPARISION PROGRAM SURVEILLANCE REQlIREMENT 4-90 f 4.31 }(UCLEAR SERVICE ELECIRECAL BUILDING EMERGENCY HEATING 4-91

'l VENTILATION AND AIR CONDITIONING l Proposed :'

1 Amendment No. (8, 97, 98, 171

! v11 l !

i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued)

SAction EAgt 4.32 TOI DIESEL GENERATOR CONTROL ROOM ESSENTIAL VENTILATION 4-92

!: SYSTEM j 4.34 METEOROLOGICAL MONITORING INSTRUMENTATION 4-93 3

4.35 HYOROGEN RECOMBINERS 4-95 5 DESIGN FEATURES 5-1 5.1 1111 5-1 5.1.1 Exclusion Area 5-1 I

i 5.1.2 Low Poculation Zone 5-1 j 5.1.3 Site Boundarv for Gaseous and Liauid Effluents (10 CFR 50 5-1 >

App. I. Guidelines) ,

J 1

! 5.1.4 Site Boundary for Liould Effluents (10 CFR 20 Compliance) 5-1

i' 5.2 CONTAINMENT 5-2 j i ;

5.2.1 Reactor Buildina 5-2 5.2.2 Reactor Building Isolation System 5-3 !

i 5.3 REACTOB 5-4 !

5.3.1 Reactor Core 5-4 l l

! 3.3.2 Reacter_ Coolant System 5-4 5.4 NEH AND SPENT FUEL STORAGE FACILITIES 5-6 :

]

5.4.1 Kew Fuel Insoection and Temocrary Storaae Rack 5-6 i i I 5.4.2 New and Soent Fuel Storage Racks and Failed Fuel 5-6 i Storage Container Rack l

1 5.4.3 New and Spent Fuel Temocrary Storage 5-6 l 5.4.4 Scent Fuel Pool and Storage Rack Design 5-6 :

1 Proposed

) Amendment No. 28, 88, 94, 95, 98, 171 l vita

-- -._y*, . . , , g :s, _ , _ . .- _ _ . __--____.,_ _._ _ - - - - - - -

RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued)

Section Eage 6.0 ADMINISTRATIVE CONTROLS 6-1 6.1 RESPONSIBILITY 6-1 6.2 ORGANIZATION 6-1 6.3 FACILITY STAFF OUALIFICATIONS 6-3 6.4 TRAINING 6-3

, 6.5 REVIEW AND AUDIT 6-3 6.5.1 Plant Review Committpe (PRC) 6-3 6.5.2 Hanagement Safety Review Committee (HSRC) 6-6 6.5.3 Technical Review and Control 6-8 6.5.4 Audits 6-9 6.6 LICENSEE EVENT REPORT ACTION 6-10a l 1 6.7 SAFETY LIMIT VIOLATION 6-11

. 6.8 PROCEQURES 6-11 i

l 6.9 REPORTING REOUIREMENTS 6-12 6.10 RECORD RETENTION 6-13 6.11 RADIATION PROTECTION PROGRAM 6-14 6.12 DELETED 6.13 HI/iH RADIATION AREA 6-15 !

6.14 ENVIRONMENTAL OUALIFICATION 6-16 6.15 PROCESS CONTROL PROGRAM (PCP) 6-17 ii i

6.16 QEESilE DOSE CALCULATION AND RADIOLOGICALHONITORING 6-18

PROGRAM MANUALS !

6.17 BA10R CHANGES TO RADIOACTIVE WASTE TREATHENT SyEUMS C-19 t (LIOUID. GA.SEOUS. AND SOLIDI 6.18 POSTACCIDENT SAMPLING 6-22 Proposed Amendment No. 28, 80, 96, 98, 171

viii i

b

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS LIST OF TABLES IAbit E&Et 1.2-1 OPERATIONAL H00ES 1-2c 1.9-1 FREQUENCY NOTATION 1-10 2.3-1 REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS 2-9 3.0-1 APPLICABILITY OF SPECIFICATIONS 3.0.3 AND 3.0.4 3-Oa 3.3-1 PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES 3-22b 3.5.1-1 INSTRUMENTS OPERATING CONDITIONS 3-27 I

3.5.5-1 ACCIDENT HONITORING INSTRUMENTATION OPERABILITY 3-38e REQUIREMENTS ;

3.5.7-1 EMERGENCY SHUTDOWN INSTRUMErlTATION 3-38j 3.6-1 SAFETY FEATURES CONTAINHENT ISOLATION VALVES 3-40 3.7-1 VOLTAGE PROTECTION SYSTEM RELAY TRIP VALUES 3-43a l 3.7-2 VOLTAGE PROTECTION SYSTEM LIMITING CONDITIONS 3-43b 3.12-1 SAFETY RELATED HYDRAULIC SNUBBERS 3-51a !

3.14-1 FIRE DETECTION INSTRUMENTS FOR SAFETY SYSTEMS 3-55 1 3.14-2 INSIDE BUILDING FIRE HOSE STATIONS 3-56b ,

] 3.14-3 CARBON DIOXIDE SUPPRESSION ZONES 3-56d d

I 3.14-4 FIRE HOSE STATIONS 3-57a l t

3.15-1 RADIOACTIVE LIQUID EFFLUENT HONITORING INSTRUMENTATION 3-61 [

3.16-1 RADIOACTIVE GASEOUS EFFLUENT HollITORING INSTRUMENTATION 3-64 i i

. 3.22-1 RADIOLOGICAL ENVIRONMENTAL HONITORING PROGRAM 3-83 :

I 3.22-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS 3-86 l

IN ENVIRONMENTAL SAMPLES !

1' !

3.29-1 NETEOR0 LOGICAL HONITORING INSTRUMENTATION 3-95 i I

i 4.0-1 APPLICABILITY OF SPECIFICATIONS 4.0.2 AND 4.0.3 4-Oa ,

4.1-1 INSTRUMENT SURVEILLANCE REQUIREMENTS 4-3 l

l Proposed I Amendment No. Et, 80, 85, 87, 88, 9A, 97, 98, 171 '

i 1x i

i

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS LIST OF TABLES IAhlt EASA 4.1-2 HINIMUM EQUIPMENT TEST FREQUENCY 4-8 4.1-3 HINIMUM SAMPLING FREQUENCY 4-9 4.2-1 CAPSULE ASSEMBLY WITHDRAWAL SCHEDULE AT DAVIS-BESSE 1 4-12b 4.6-1 DIESEL GENERATOR TEST SCHEDULE 4-34j 4.6-2 ADDITIONAL RELIABILITY ACTIONS 4-34k 4.14-1 SNUBBERS ACCESSIBLE DURING POWER OPERATIONS 4-47d 4.17-1 HINIMUM NUHRER OF STEAM GENERATORS TO BE INSPECTED 4-56 DURING INSERVICE INSPECTION 4.17-2A STEAM GENERATOR TUBE INSPECTION 4-57 4.17-28 STEAM GENERATOR TUBE INSPECTION (SPECIFIC LIMITED AREA) 4-57a i

4.17-3 OTSG AUXILIARY FEEDWATER HEATER SURVEILLANCE 4-S7b 4.19-1 RADIOACTIVE LIQUID EFFLUENT HONITORING INSTRUMENTATION 4-64 SURVEILLANCE REQUIREMENTS L

4.20-1 RADIOACTIVE GASEOUS EFFLUENT HONITORING INSTRUMENTATION 4-66 l SURVEILLANCE REQUIREMENTS 4.21-1 RADI0 ACTIVE LIQUID HASTE SAMPLING AND ANALYSIS PROGRAM 4-70 j 4.21-2 RADIOACTIVE LIQUID HASTE SAMPLING AND ANALYSIS PROGRAM 4-72b 4.22 1 RADIOACTIVE GASEOUS HASTE SAMPLING AND ANALYSIS PROGRAM 4-75 '

4.26-1 MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD) 4-84 l 4.28-1 EXPLOSIVE GAS HIXTURE INSTRUMENTATION SURVEILLANCE 4-88 l REQUIREHENTS j 4.34-1 HETi.OR0 LOGICAL HONITORING INSTRUMENTATION 4-94 6.2-1 SHIFT CREN PERSONNEL AND LICENSE REQUIREMENTS 6-2 l Proposed Amendment No. 28, 66, 76, 88, 94, 96, 97, 98, 171 x

I

RANCHO SECO UNIT 1 !

TECHNICAL SPECIFICATIONS ,

l LIST OF FIGURES l l Fiaure i 1 2.1-1 Core Protection Safety Limit, Pressure vs. Temperature 2-3a l

2.1-2 Core Protection Safety Limits, Reactor Power Imbalance 2-3b l

4 2.1-3 Core Protective Safety Bases 2-3c i

i i 2.3-1 Protective System Maximum Allowable Setpoints, 2-10 i Pressure vs. Temperature !

2.3-2 Protective System Maximum Allowable Setpoints, 2-11 l f Reactor Power Imbalance l 3.1.2-1 Reactor Coolant System Pressure-Temperature Limits 3-5 l for Heatup for the First 8 EFPY j 3.1.2-2 Reactor Coolant System Pressure-Temperature Limits 3-5a l !

for Cooldown for the First 8 EFPY !

t j 3.1.2-3 Inservice Leak and Hydrostatic Test (8 EFPY) Heatup 3-5b l j and Cooldown ;

l 3.1.9-1 Limiting Pressure vs. Temperature for Control 3-16a l .

Rod Drive Operation l

^

3.5.2-1 Rod Index vs. Power Level for Four-Pump Operation, 3-33c l t 0 to 40 EFPD !

t 3.5.2-2 Rod Index vs. Power Level for Four-Pump Operation 3-33d i After 30 EFPD l

' i j 3.5.2-3 Rod Index vs. Power Level for Four-Pump Operation 3-33e l !

I After 300 EFPD with APSRs Hithdrawn j i l

3.5.2-4 Rod Index vs. Power Level for Three-Pump Operation, 3-33f l !

O to 40 EFPD j 3

! 3.5.2-5 Rod Index vs. Power Level for Three-Pump Operation 3-339 l l a After 30 EFPD '

l l

3.5.2-6 Rod Index vs. Power Level for Three-Pump Operation 3-33h l j After 300 EFPD with APSRs Withdrawn l l l 1

l i

l Proposed l Amendment No. 26, 28, 29, 48, 83, 69, 87, 171 xi

i l .'

RANCHO SECO UNIT 1 ;

j TECH'MCAL SPECIFICATIONS j LIST OF FIGURES (Continued) l ELSE.t [

3.5.2-7 Core Imbalance vs. Power Level O to 40 EFPD 3-331 l ,

3.5.2-8 Core Imbalance vs. Power Level 3-33j i After 30 EFPD j l 3.5.2-9 Core Imbalance vs. Power Level 3-33k l !

! After 300 EFPD with APSRs Hithdrawn t i

) 3.5.4-1 Incore Instrumentation Specification Axial 3-38a l l 1 Imbalance Indication l t

d 3.5.4-2 Incore Instrumentation Specification 3-38b l i Radial Flux Tilt Indication l t

I 3.5.4-3 Incore Instrumentation Specification 3-38c ;

A.13-1 Main Steam Inservice Inspection 4-464 f

4.13-2 Main Feedwater Inservice Inspection 4-46b (

l 4.13-3 Main Steam Dump Inservice Inspection 4-46c f 1 r j 5.1-1 Exclusion Area 5-la l 5.1-2 Low Population Zone (5 mile radius) 5-lb (

f

! 5.1-3 Site Boundary for Gaseous and Liquid Effluents 5-Ic (10 CFR 50 Appendix I Guidelines)

]

j 5.1-4 Site Boundary for Liquid Effluents for 5-Id l 1

10 CFR 20 Compliance :

6.2-1 SMUD Corporate Support of Nuclear Safety 6-2a 6.2-2 Nuclear Organization Chart 3-2b l l l 1

l i

l Proposed

' Amendment No. 26, 28, 29, 48, 69, 96, 98, 171 xil l

l

RANCHO SECO UllIT 1 i TECHNICAL SPECIFICATIONS l

< Defiritions 1.2.7 Refuelino Ooeration l

! An operation involving a change in core geometry by manipulation of fuel or ;

i control rods when the reactor vessel head is removed. ;

i ,

l.2.8 Refuelina Interval * !

18 months.

l 1.2.9 Startup i

The reactor shall be considered in the startup mode when the shutdown margin 3

is reduced with the intent of going critical. :'

i

1.2.10 Remain Critical See Specification 3.0.3.

2 1.2.11 T gg ,

i At operating conditions T is defined as the arithmetic average of the coolanttemperaturesint$ghot and cold legs of the loop with the greater !

{ number of reactor coolant pum operating, if such a distinction of loops i 2

can be made.

1.2.12 Reatuo - Cooldown j The reactor is in heatup-cooldown when the range of reactor coolant

temperature is greater than 200'F and less than 525'F.

't l 1.2.13 &c.tlon s

Action including time requirements shall be that part of a specification

which prescribes remedial measures required under designated conditions.

I 1.2.14 LeAkun I

l A. IDENTIFIED LEAKAGE shall be' e

4 j a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as j pump seal or valve packing leaks that are captured and conducted i j to a sump or collecting tank, or [

j b. Leakage into the containment atmosphere from sources that are

both specifically located and known either not to interfere with ,

! the operation of leakage detection systems or not to be PRESSURE j BOUNDARY LEAKAGE, oi' l

'See Page 1-2b a

) Proposed Amendment No. 38, 67, 87, 93, 97, 171 1-2 i

-- ,-c., . - - - , - - - - , , , , - .---,-v-,,--,, - , - - - - n--n.- - -. , e-- - - , - -- , - , = - - - - , , - - - - - +

4 ;

RANCHO SECO UNIT 1 5 i TECHNICAL SPECIFICATIONS :

Definitions

! c. Reactor coolant system leakage through a steam generator to the i

secondary system. !

' 1 B. UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED !

j LEAKAGE or CONTROLLED LEAKAGE. l l C. PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator <

tube leakage) through a non-isolable fault in a Reactor Coolant System [

component body, pipe wall or vessel wall. ;

l D. CONTROLLED LEAKAGE shall be that leakage from the reactor coolant pump .

seals. l i

1.3 OPERABLE l 1

A component or system is operable when it is capable of performing its ;

! intended function within the requireo range. The component or system shall !

be considered to have this capability when: (1) it satisfies the limiting '

conditions for operation defined in Specification 3 (2) it has been tested periodically in accordance with Specification 4, and has met its performance ,

t requirements. (3) the system has available its normal and emergency sources i of power, and (4) its required auxiliaries are capable of performing their l intended function. When a system or component is determined to be l 4 inoperable solely because its normal power source is inoperable or its i l emergency power source is inoperable, it may be considered OPERABLE for the !

purpose of satisfying the requirements of its applicable Limiting Condition !

l for Operation provided its redundant system or component is OPERABLE with an r

OPERABLE normal and emergency power source. [

1.4 PROTECTION INSTRUMENTATION LOGIC J

1.4.1 Instrument Channel An instrument channel is the combination of sensor, wires, amplifiers and ouput devices which are connected for the purpose of measuring the value of i a process variable for the purpose of observation, control and/or j protection. An instrument channel may be either analog or digital. '

) 1.4.2 flgactor Protection System I

i The reactor protection system is shown in Figures 7.1-1 and 7.2-2 of the j i USAR. It is that combination of protective channels and associated

circuitry which forms the automatic system that protects the reactor by l control rod trip. It includes the four protection channels, their 1 associated instrument channel inputs, manual trip switch, all rod drive control protective trip breakers and activating relays or coils.

l t

J Proposed j Amendment No. 61, 87, 97, 171 1-2a L _

i l.'

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Definitions TABLE 1.2-1 i OPERATIONAL MODES I

i Operaticnal Reactivity Coolant Indicated Remarks i Mode Condition Temperature Neutron

, kerr Tavg*F Power (1 of Rated Power) ,

', Power >0.99 >525 12 l

Operation Hot >0.99 >525 <2 i Standby i

Startup >0.99 >525 <2

, <1.00 3

Hot 10.99 1525 0 I

Shutdown Heatup - 10.99 >200 0 j Cooldown <525 i Cold 10.99 1200 0 RCS Pressure as Shutdown defined in 3.1.2. :

i Refueling' 10.95 1140 at 0 RCS Pressure !

) Shutdown DHR pump as defined l

! suction in 3.1.2. l i !

i Refueling Reactor Vessel head removed and Refueling Shutdown conditions

operation i

)

Fuel in the reactor vessel with the vessel head closure bolts less than i j fully tensioned or with the head removed.

i J

l I

i Proposed j Amendment No. 97, 171 1-2c 1

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

' i

, RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS -

Definitions 1.4.3 Protection Channel A protection channel, as shown in Figure 7.1-1 of the USAR (one of three or l one of four independent channels, complete with sensors, sensor power supply units, amplifiers and bistable modules provided for every reactor protection safety parameter), is a combination of instrument channels forming a single

) digital output to the protection system's coincidence logic. Each protection channel includes two key-operated bypass switches, a protection channel bypass switch and a shutdown bypass switch.

1.4.4 Reactor Protection System Loaic !

This system utilizes reactor trip module relays (coils and contacts) in all l four of the protection channels as shown in Figure 7.1-1 of the USAR, to l 7 provide reactor trip signals for de-energizing the six control rod drive ,

trip breakers. The control rod drive trip breakers are arranged to provide r a one-out-of-two-times-two logic. Each element of the one-out-of-two- '

times-t.o Icgic is controlled by a separate two-out-of-four logic from the !

four reactor protection channels. With one channel bypassed and untripped. l 4

l the two-out-of-four logic functions as a two-out-of-three logic for the 4

three active channels.

1 j 1.4.5 Safety Features System Logic This system utilizes relay contact output from individual channels arranged '

, in three analog sub-systems and two two-out-of-three logic sub-systems as 1 shown in Figure 7.1-5 of the USAR. The logic sub-system is wired to provide l appropriate signals for the actuation of redundant safety features equipment i on a two-of-three basis for any given parameter.

I 1.4.6 Dearee of Redundancy

) The difference between the number of operable channels and the number of j !

1 channels which, when tripped, will cause an automatic system trip.

j l 1.5 INSTRUMENTATION SURVEILLANCE

! 1.5.1 Trio Test 1

A trip test is a test of logic elements in a protection channel to verify their associated trip action.

I) 1.5.2 Channel Test A channel test is the injection of an internal or external test signal into

] the channel to verify its proper response, including alarm and/or trip

, initiating action, wh:re applicable.

1.5.3 Instrument Channel Check I An instrument channel check is a verification of acceptable instrument i performance by observation of its behavior and/or state; this verification J

includes comparison of output and/or state of independent channels measuring j the same variable.

Proposed l Amendment No. 87, 171 l-3

i RANCHO SECO UNIT 1

TECHNICAL SPECIFICATIONS !

Definitions D. All automatic containment isolation valves are operable or closed ;

in the safety features position. :,

I E. The containment leakage satisfies Specification 4.4.1 and no known !

changes have occurred. ;

1.8 LICENSEE EVENT REPORTS !

t 2

Defined under Administrative Controls Section 6.9.4. !

] 1.9 TIME PERIODS

! May be extended to a maximum ot +25% to accommodate operations l ,

scheduling. The total maximum combined interval time for any three consecutive intervals shall not exceed 3.25 times a single specified !

surveillance interval. i 1.9.1 SHIFT j A time period covering at least once per twelve (12) hours. ;

i 1.9.2 DAILY !

I

A time period spaced to occur at least once per twenty-four (24) hours. !

1.9.3 WEEKLY j

A time period spaced to occur at least once per seven (7) days. ,

1.9.4 FORTNIGHTLY l

A time period spaced to occur once per fourteen (14) days.

1.9.5 HONTHLY J l 1 A time period spaced to occur at least once per thirty-one (31) days. ;

1.9.6 0UARTERLY j A time period spaced to occur at least once per ninety-two (92) days. (

, 1.9.7 SEMIANNUALLY !'

I A time period spaced to occur at least once per six (6) months.

) 1.9.8 ANNUALLY i A time period spaced to occur at least once per twelve (12) months.

l 1.9.9 DELETED l j 1.9.10 REFUELING INTERVAL

} A time period spaced to occur at least once per eighteen (18) months, i ,

1.10 SAFE 1Y i

Safety as used in these Technical Specifications shall mean nuclear safety and shall encompass all systems and components that have or may have an effect on the health and safety of the general public.

Proposed

Amendment No. 63, 87, 171 l 1-5

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Definitions 1.24 VENTILATION EXHAUST TREATHENT SYSTEMS The VENTILATION EXHAUST TREATHENT SYSTEMS are systems designed and installed to re h e gaseous radiolodine or radioactive material in particulate fsrm in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers Lnd/or HEPA filters for the purpose of removing lodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TP.EATHENT SYSTEMS components.

1.25 PjpCE - PURGING PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

1.26 YENTING VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during ,.

1 VENTING. Vent, used in system names, does not imply a VENTING '

process.

1.27 i ishallbetheaverage(weightedinproportiontothe j 1

concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of average beta and gamma energies l per disintegration (in HEV) for isotopes with half lives greater than 20 minutes, making up at least 95% of the total activity in the coolant (excluding todines). ;

1-i Proposed Amendment No. 98, 171 1-9

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Definitions TABLE 1.9-1 EREQUENCY NOTATION NOTATIQS FREOUENCY S At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

H At least once per 7 days.

H At least once per 31 days.

Q At least once per 92 days.

SY At least once per 184 days.

A At least once per 12 months R At least once per 18 months.

S/U Prior to each reactor startup.

I P Completed prior to each release.

NA Not appilcable, i

i I

I i

l i

l Proposed i Amendment No. 98, 171 1

1-10 l

, RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting i Safety System Settings 2.2 SAFETY LIMITS. REACTOR SYSTEM PRESSURE J

Apolicability i

Applies to the limit on reactor coolant system pressure. !

l Ob3ective 4

To maintain the integrity of the reactor coolant system and to prevent the l

) release of significant amounts of fission product activity. '

1 SAtti.f.lCA.tlon

2.2.1 The reactor coolant system pressure shall not exceed 2750 psig when [

l there are fuel assemblies in the reactor vessel, i 2.2.2 The nominal setpoint of the pressurizer code safety valves shall be f

less than or equal to 2500 psig. ;

f i Balti l The reactor coolant system (I) serves as a barrier to prevent radionuclides !

, in the reactor coolant from reaching the atmosphere. In the event of a fuel ,

4 cladding failure, the reactor coolant system is a barrier against the release i of fission products. Establishing a system pressure limit helps to assure l I the integrity of the reactor coolant system. The maximum transient pressure [

allowable in the reactor coolant system pressurt vessel under the ASME code, i Section III, is 110 percent of design pressure.(2) The maximum transient !

pressure allowable in the reactor coolant system piping, valves, and fittings t under ANSI Section B31.7 is 110 percent of design pressure. Thus, the safety i e limit of 275 psig (110 percent of the 2500 psi design pressure) has been !

) established.p2i Thesettingsforthereactorhfghpressur l l l psig) and the pressurizer code safety valves (2500 psig) (t)been 3 have trip (2355 '

established to assure that the reactor coolant system pressure safety limit s l is not exceeded. The initial hydrostatic test was conducted at 3125 psig 3

(125 percent of design pressure) to verify the integrity of the reactor l l coolant system. Additional assurance that the reactor coolant system I pressure does not exceed the safety limit is provided by setting the !

pressurizer electromatic relief valve at 2450 psig. This setpoint is above

normal transients limited by setting the reactor trip at 12355 psig and l

. sufficiently low to assure limited depenaence on safety valves operation.

REFERENCES l l

(1) USAR, section 4 j (2) USAR, paragraph 4.3.8.1 l

(3) USAR, paragraph 4.2.4 !

j Proposed !

Amendment No. 31, 83, 87, 171 l j 2-4 i j

l

\ .

. i RANCHO SECO UNIT 1 f TECHNICAL SPECIFICATIONS l

i Safety Limits and Limiting Safety System Settings of rated power. Adding to this the possible variatici in trip setpoints due ,

1 to calibration and instrument errors, the maximum actual power at which a !

I trip would b !

analysis.(4)e actuated could be 112 percent, which was used in the safety l, i A. Overpower trip based on flow and imbalance ,

J The power level trip setpoint produced by the reactor coolant system :

flow is based on a power-to-flow ratio which has been established to !

i accommodate the most severe thermal transient considered in the i

design, the loss-of-coolant flow accident from high power. The i J analysis in USAR Section 14 demonstrates the adequacy of the specified l t i

power to flow ratio.

The power level trip setpoint produced by the power-to-flow ratio :

! provides both high power level and low flow protection in the event t the reactor power level increases or the reactor coolant flow rate i j decreases. The power level trip setpoint produced by the power to i flow ratio provides overpower DNB protection for all modes of pump

. operation. For every flow rate there is a maximum permissible low I

flow rato. Typical power level and low flow rate combinations for the ;

} pump situations of Table 2.3-1 are as follows: [

I

]

! 1. Trip would occur when four reactor coolant pumps are operating if l l' power is 106 percent and reactor flow rate is 100 percent, or flow '

rate is 94.34 percent and power level is 100 percent, i

2. Trip would occur when three reactor coolant pumps are operating if i

- power is 78.8 percent and reactor flow rate is 74.4 percent or flow rate is 70.75 percent and power level is 75 percent.

l

3. Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 51.4 ,

percent and reactor flow rate is 48.5 percent or flow rate is !'

l 46.22 percent and the power level is 49 percent.

) For safety analysis calculations the maximum calibration and instrumentation i errors for the power level were used. !

\ l l The power-imbalance boundaries are established in order to prevent reactor '

I thermal limits from being exceeded. These thermal limits are either power i peaking kH/f t limits or DNBR limits. The reactor power imbalance (power in 1 the top half of core minus power in the bottom half of core) reduces the l power level trip produced by the power-to-flow ratio so that the boundaries 1

of Figure 2.3-2 are produced. The power-to-flow ratio reduces the power i level trip and associsted reatt .-power reactor-power-imbalance boundaries by

{ l.06 percent for a 1 percent flow reduction, l

i, i Proposed i Amendment No. 30, 33, 8/, 171 2-6

l i

RANCHO SECO UNIT 1 l TECHNICAL SPECIFICATIONS l

Safety Limits and Limiting j Safety System Settings ;

Table 2.3.1. Two conattions are imposed when the bypass is used: f By administrative control the nuclear overpower trip set point i

1. '

must be reduced to a value of 15.0 percent of rated power during reactor shutdown. j i

2. A high reactor coolant system pressure trip set point of 1820 psig is automatically imposed. f i

The purpose of the 1820 psig high pressure trip set point is to i prevent normal operation with part of the reactor protection system t bypassed. This high pressure trip set point is lowirr than the l normal low pressure trip set point so that the reactor must be I tripped before the bypass is initiated. The overpower tr'p set ;

point of 15.0 percent prevents any significant reactor power from i being produced when performing the physics tests. Sufficient !'

natural circulation (5) would be available to remove 5.0 percent of rated power if none of the reactor coolant pumps were .1perating.

G. Anticipatory Reactor Trip An Anticipatory Retctor Trip is nrovided to ensure a reactor trip {

Won loss of feedwater or turbine trip, in response to NUREG 0737 l ttem II.K.2.10. (6)

REFERENCES (1) USAR, paragraph 14.1.2.2 l

(2) USAR, paragraph 14.1.2.7 (3) USAR, paragraph 14.1.2.8 (4) USAR, paragraph 14.1.2.3 (5) USAR, paragraph 14.1.2.6 l

(6) SMUD to NRC Letter dated May 5, 1982 l

Proposed Amendment No. 49. 171 2-8

=_

l RANCHO SECO UNIT 1 l!

l TECHNICAL SPECIFICATIONS +

)

Limiting Conditions for Operation l

3. LIMITING CONDITIONS FOR OPERATION h 1

! 3.0 Ceneral Limiting Conditions For Operation .

l

[

3.0.1 Compliance with the Limiting Conditions for Operation contained [

4 in the succeeding Specifications is required during the ;

i OPERATIONAL MODES or other conditions specified therein; except i

that upon failure to meet the Limiting Conditions for Operation, i

^

the associated Action, including time requirements, shall be (

met.

i 1 3.0.2 Noncompliance with a Specification shall exist when the ,

i requirements of the Limiting Condition for Operation and j

- associated Action, including time requirements, are not met t I within the specified time intervals. If the Limiting Condition l for Operation is restored prior to expiration of the specified ;

i time Intervals, completion of the Action, including time requirements, is not required.

J ,

3.0.3 When a Limiting Condition for Operation is not met, except as ,

i provided in the associated Action including time requirements, ;

j within I hour action shall be initiated to place the unit in a l M00E in which the Specification does not apply to placing it, k as applicable, in [

1 !

1. At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , !

2. At least HOT SHUT 00HN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and :

3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . j l

Where corrective measures are completed that permit operation !

Under the Action including time requirements, the Action !

' including time requirements may be taken in accordance with the l l specified time limits as measured from the time of failure to !

meet the Limiting Condition for Operation. Exceptions to these I

] requirements are stated in the individual specifications and !

Table 3.0-1, 1 3.0.4 Entry into an OPERATIONAL MODE or other specified condition :

] shall not be made when the conditions for the Limiting i j

Conditions for Operation are not met and the associated Action, including time requirements, requires a shutdown if they are not met within a specified time interval. Entry into an j

OPERATIONAL MODE or specified condition may be made in accordance with Action including time requirements when

conformance to them permits continued operation of the facility j for an unlimited period of time. This provision shall not j prevent passage through or to OPERATIONAL MODES as required to comply with Action requirements. Exceptions to these 3

7 requirements are stated in the individual specifications and

, Table 3.0-1.

i

) Proposed j Amendment No. 91, 171 3-0 l

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! RANCHO SECC UNIT 1

! TECHNICAL SPECIFICATIONS 4

{ Limiting Conditions For Operation 1 TABLE 3.0-1 l

i Applicability of Specifications 3.0.3 and 3.0.4. (The "NA" indicates that l' the provisions of Specification (s) 3.0.3 and/or 3.0.4 are not applicable to

{ the sections identified.)

l 1

i Section Specification 3.0.3 Specification 3.0.4 l l 3.1 l I

j 3.2.1 I

3.2.2 N/A j

3.3 i l 3.4 3.5.1 i

, _ 3 _. 5 . 2 !

3.5.3 j

3.5.4 N/* N/A t l

j 3.5.5 N/A !

3.5.6 N/A N/A !

_ 3.5.7 N/A

(

3.6 N/A

{

) 3.7 j 3.0 N/A l 3.9 N/A N/A l j 3.10 l l

3.11 N/A N/A

3.12 l

) 3.13 #

l 3.14 N/A N/A I j

3.15 N/A N/A 3.16 N/A N/A l f j 3.17 N/A N/A l'

l 3.18 N/A N/A 7 3.19 N/A N/A l

i

The provisions of Specification 3.0.3 are not applicable when the reactor is in Refueling Shutdown or Refueling Operation.

l Proposed t

' Amendment No. 97, 171 l 3-Oa )

4 i l l

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions For Operation TABLE 3.0-1 (Continued)

Section Specification 3.0.3 Specification 3.0.4 3.20 N/A N/A 3.11___ N/A N/A

.._.3.22 N/A N/A 1 ___.3.23 N/A N/A 3.24 N/A N/A

___li25 N/A N/A

.3.26 N/A N/A 3.27

__._ld B 1129 N/A N/A 3.30

)

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.i j Proposed Amendment No. 97, 171 3-Ob

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS

Limiting Conditions For Operation Balti Spgeifications 3,0.1 through 3.0.4 establish the general requirements applicable to Limiting Conditions for Operation. These requirements are based on the requirements for Limiting Conditions for Operation stated in the Code of Federal Regulations. 10 CFR 50.36(c)(2)

"Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for

, operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specification until the condition can be met."

SotcificAtlolL1dd establishes the Applicability statement within each individual specification as the requirement for when (i.e., in which l OPERATIONAL H0 DES or other specified conditions) conformance to the Limiting '

Conditions for Operation is required for safe operation of the facility. The ACTION requirements establish those remedial measures that must be taken within specified time limits when the requirements of a Limiting Condition for Operation are not met.

There are two basic types of ACTION requirements. The first specifies the

remedial measures that permit continued operation of the facility which is not further restricted by the time limits of the ACTION requirements. In this case, conformance to the ACTION requirements provides an acceptable level of safety for unlimited continued operation as long as the ACTION i requirements continue to be met. The second type of ACTION requirement l J specifies a time limit in which conformance to the conditions of the Limiting l l Condition for Operation must be met. This time limit is the allowable outage ,

j time to restore an inoperable system or component to OPERABLE status or for j restoring parameters within specified limits. If these actions are not

completed within the allowable outage time limits, a shutdown is required to ,

! place the facility in a HODE or condition in which the specification no '

longer applies. It is not intended that the shutdown ACTION requirements be >

used as an operational convenience which permits (routine) voluntary removal !

of a system (s) or component (s) from service in lieu of other alternatives j

. that would not result in redundant systems or components being inoperable.

1 J

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i Proposed Amendment No. 171 3-Oc i

)

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions For Operation Halu (Continued)

The specified time limits of the ACTION requirements are applicable from the ,

time it is identified that a Limiting Condition for Operation is not met.

The time limits of the ACTION requirements are also applicable when a system or component is removed from service for surveillance testing or investigation of operational problems. Individual specifications may include a specified time limit for the completion of a Surveillance Requirement when equipment is removed from service. In this case, the allowable outage time limits of the ACTION requirements are anplicable when this limit expires if the surveillance has not been completed. When a shutdown is required to comply with ACTION requirements, the plant may have entered a H00E in which a new specification becomes appilcable. In this case, the time limits of the ACTION requirements would apply from the time that the new specification becomes appitcable if the requirements of the Limiting Condition for Operation are not met.

Specif.ication 3.0.2 establishes that noncompliance with a specification exists when the requirements of the Limiting Condition for Operation are not met and the associated ACTION requirements have not been implemented within the specified time interval. The purpose of this specification is to clarify that (1) implementation of the ACTION requirements within the specified time interval constitutes compliance with a specification and (2) completion of the remedial measures of the ACTION requirements is not required when compliance with a 1.imiting Condition for Operation is restored within the time interval specified in the associated ACTION requirements.

Specification 3.0.3 establishes the shutdown ACTION requirements that must be implemented when a Limiting Condition for Operation is not met and the condition is not specifically addressed by the associated ACTION requirements. The purpose of this specification is to delineate the time limits for placing the unit in a safe shutdown HODE when plant operation cannot be maintained within the limits for safe operation defined by the Limiting Conditions for Operation and its ACTION requirements. It is not intended to be used as an operational convenience which permits (routine) voluntary removal of redundant systems or components from service in lieu of other alternatives that would not result in redundant systems or components being inoperable. One hour is allowed to prepare for an orderly shutdown before 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 availability of the electrical grid.

The time limits specified to reach lower H0 DES of operation permit the shutdown to proceed in a controlled and orderly manner that is well within the specitled maximum cooldown rate and with'n the cooldown capabilities of the facility assuming only the minimum required equipment is OPERABLE. This reduces thermal stresses on components of thi primary coolant system and the potential for a plant upset that could challeige safety systems under conditions for which this specification applivs.

Proposed Amendment No. 171 3-Od

i i ..

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS l I

Limiting Conditions For Operation J

3 1

Balts (Continued) l j If remedial measures permitting limited continred operation of the facility i j under the provisions of the ACTION requirementt are completed, the shutdown !

may be terminated. The time limits of the ACTION requirements are applicable :

from the point in time there was a failure to meet a Limiting Condition for (

Operation. Therefore, the shutdown may be terminated if the ACTION t l requirements have been met or the time limits of the ACTION requirements have l not expired, thus providing an allowance for the completion of the required .

actions, f

] The time limits of Specification 3.0.3 allow 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months for the plant to be in !

the COLD SHUTDOHN H00E when a shutdown is required during the PCMER N00E of

~

operation. If the plant is in a lower H00E of operation when a shutdown is (

, required, the time limit for reaching the next lower H00E of operation i i applies. However, if a lower H0DE of operation is reached in I ns time than I allowed, the total allowable time to reach COLD SHUTDOHN or other applicable '

l H00E, is not reduced. For example, if HOT STANOBY is reached in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the

] time allowed to reach HOT SHUTDCHN is not reduced from the allowable limit of t 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months . Therefore, if remedial measures are completed that would permit a

~

i return to P0HER operation, a penalty is not incurred by having to reach a ,

S lower H00E of operation in less than the total time allowed. ;

I The saMe principle applies with regard to the allowable outage time limits of '

the ACTION requirements, if compliance with the ACTION requirements for one i specification results in entry into a H00E or condition of operation for l 1 another specification in which the requirements of the Limiting Condition for i j Operation are not met. If the new specification becomes applicable in less i J

time than specified, the difference may be added to the allowable outage time '

1 limits of the second specification. However, the allowable outage time limits of ACTION requirements for a higher HODE of operation may not be used to extend the allowable outage time that is applicable when a Limiting Condition for Operation is not met in a lowcr H00E of operation.

The shutdown requirements of Specification 3.0.3 do not apply when the reactor is in the COLD SHUTDOHN, REFUELING SHUTDOHN, or REFUELING OPERATION operational mode, becaute the ACTION requirements of individual specifications define the remedial measures to be taken.

I ~

f i i I

I i i

i Proposed Amendment No. 171 {

3-Oe j l l

! - - - - - - . - - - _ _ _ _ - . - - - _ -- - - I

l i

RANCHO SECO UNIT 1 l TECHNICAL SPECIFICATIONS i l

Limiting Conditions For Operation Bain (Continued)

Specification _3A4 establishes limitations on H00E changes when a Limiting Condition for Operation is not met. It precludes placing the facility in a higher H00E of operation when the requirements for a Limiting Condition for Operation are not met and continued noncompliance te these conditions would result in a shutdown to comply with the ACTION requirements if a change in HODES was permitted. The purpose of this specification is to ensure that facility operation is not initiated or that higher HOMS of operation are not entered when corrective action is being taken to obtain compliance with a specification by restoring equipment to OPERABLE status or parameters to specified limits. Compliance with ACTION requirements that permit continued operation of the facility for an unlimited period of time provides an acceptable level of safety for continued operation without regard to the status of the plant before or af ter a HODE change. Therefore, in this case, entry into an OPERATIONAL H00E or other specified condition may be made in accordance with the provisions of the ACTION requirements. The provisions of this specification should not, however, be interpreted as endorsing the failure to exercise good practice in restoring systems or components to OPERABLE status before plant startup.

When a shutdown is required to comply with ACTION requirements, the provisions of Specification 3.0.4 do not apply because they would delay placing the facility in a lower H00E of operation.

Proposed Amendment No. 171 3-Of i

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1.2 PRESSURIZATION, HEATUP, AND COOLDOWN LIMITATIONS SpecificAt10R 3.1.2.1 Inservice Leak and Hydrat1A11c_Ititti Pressure temperature limits for the first eight Effective Full Power Years (EFPY) of inservice leak and hydrostatic tests are given in Figure 3.1.2-3. Heatup and cooldosn rates shall be restricted according to the rates specified in Figure 3.1.2-3.

3.1.2.2 Rettyp_Caoldown:

for the first eight EFPY of operation, the reactor coolaat pressure l l and the system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with figure 3.1.2-1 !

and Figure 3.1.2-2, respectively. The Reactor Coolant System i temperature and pressurv shall be determined to be within the l limits at least once per 30 minutes during system heatup, cooldown, '

and inservice leak and hydrostatic testing operations. Heatup and cooldown rates shall not exceed the rates stat l on the associated figure.

M110D If hea'sup and cooldown rates are exceeded, stabilize the temperature and !

restore the temperature and/or pressure to within the limits within 30 minutes, and perform an engineering evaluttion to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System. Through this evaluation, determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT !

SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce RCS Tavg and pressure to lets i j

than 200'F and 500 psig, respectively, within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , this action applies to Specifications 3.1.2.4 and 3.1.2.5, below. !

3.1.2.3 The secondary side of the steam generator shall not be pressurized above 200 pstg if the temperature of the steam generator shell is below 130*F.

3.1.2.4 The pressurizer heatup and cooldown rates shall not exceed 100'F in any 1-hour period.

3.1.2.5 The spray shall not be used if the temperature difference between the pressurizer and spray fluid is greater than 410*F.

I 4 Proposed Amendment No. 22, 3!, 55, 87, 97, 171 3-3 1

I

! , RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS l ,

I l Limiting Conditions for Operation The maximum allowable pressure is taken to be the lowest pressure of the three :

calculated pressures. The pressure limit is adjusted for the pressure differential between the point of system pressure measurement and (Y. Ilmiting component for all reactor coolant pump combinations. The limit curves were prepared based upon the most limiting adjusted reference temperature of all l the beltline region materials at the end of the fifth effective full power j j year, j i

The actual shift in Rig i of the beltline region material will be !

l established periodicall during operations by removing and evaluating, in !

accordance with Appendix H to 10 CFR 50, reactor vessel material irradiation surveillance specimens installed near the inside wall of this or a similar reactor vessel in the core area. Because the neutron energy spectra at the ;

i specimen location and at the vessel inner wall location are essentially the !

2 same, the measured transition shif t for a sample can be applied with l 1 '

confidence to the adjacent section of the reactor vessel. The limit curves ;

i must be recalculated when the ARTND" determined from the surveillance

, capsule is different from the calcu'ated ARTNDT for the equivalent capsule I

radiation exposure. !

The unitradiated impact properties of the beltline region materials, required by Appendices G and H to 10 CFR 50, were determined for those :

, materials for which sufficient amounts of material were available. The l i adjusted reference temperatures are calculated by adding the radiation-induced I j ARigor and the untrradiated RTNOT. The predicted ARTNDT are l s calculated using the respective neutron fluence and copper and phosphorus ;

contents in accordance with Reg. Guide 1.99, j 3 1 j The assumed RTNDT of the closure head region is 60'F and the outlet nozzle i steel forgings is 60'F. j

, t i The limitations imposed on pressurizer heatup and cooldown and spray water i

! fnperature differential are provided to assure that the pressurizer is :

, operated within the design criteria assumed for the fatigue analysis performed !

j in accordance with the ASME code requirements. !

I 1 The spray temperature difference restriction based on a stress analysis of the !

i spray line nozzle is imposed to maintain the thermal stresses at tho l

pressurizer spray line nozzle below the design limit, Temperature .

l requirements for the steam generator correspord with the measured NDTT for the [

j shell. j REFERENCES {

f (1) USAR paragraph 4.1.2.4 (2) ASME Boller and Pressure Vessel Code, Section !!! l l (3) USAR paragraph 4.3.8.5 l l (4) USAR paragraph 4.3.3 (5) USAR paragraph 4.1.2.8 and 4.3.3 (6) Analysis of Capsule RSI-B from Sacramento Municipal Utility District .

i Unit 1 Reactor Vessel Materials Surveillance Program, BAW-1702, !

l February, 1982.

Proposed l Amendment No, I, 15, 22, 31, 55, 87, 171 t

) 3-4

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l . RANCHO SECO UNIT 1 !

TECHNICAL SPECIFICATIONS j ,,

Limiting Conditions for Operation 3.1.3 HINIMUM CONDITIONS FOR CRITICAi.ITY g

,! Specif1 rations !

^

3.1.3.1 The reactor coolant temperature shall be above 525'F except for !

portions of low power physics testing when the requirements of Specification 3.1.8 shall apply.

I

' 3.1.3.2 Reactor coolant temperature shall be above Ductility Transition .

Tamperature (DTT) + 10'F.

j 3.1.3.3 Whe'1 the reactor coolant temperature is below the minimum l l temperature specified in 3.1.3.1 above, except for portions of low ;

i power physics testing when the requirements of Specification 3.1.8 ;

shall apply, the reactor shall be suberitical by an amount equal to *

or greater than the calculated reactivity insertion due to

depressurization. l l

1 Atlinn .

I With the reactor subtritical by less than the required amount, immediately j initiate and continue boration until the required SHUTDOWN MARGIN is l

restored. i I 3.1.3.4 The reactor shall be maintained subcritical b at least 1 percent I Ak/k until a steam bubble is formed and an indicated water lev 61 ;

j between 10 and 316 inches is established in the pressurizer.

l Action l With the reactor subtritical by less than the required amount, immediately !

I initiate and continue boration until the required SHUTDOWN MARGIN is !

restored. 1 3.1.3.5 Except for physics tests and as limited by 3.5.2.1 and 3.5.2.5, l safety rod groups shall be fully withdrawn prior to any other !

reiuction in shutdown m\rgin by deboration or regulating rod

withdrawal during the aaproach to criticality. Following safety rod witiidrawal. the regulating rods shall be positioned within their position limits as defined by specification 3.5.2.5 prior ta deboration.

3 Balti 1

i At the beginning of life of the initial fuel cycle, the moderator tempera-j ture coefficiunt is expected to be slightly positive at operating tempera-

] tures with ine operating configuration of control rods. (1) Calculations j Show that above 525'r the positive moderator coefficient is acceptable.

1 Since the moderator temperature coefficient at lower temperatures will be j less negative or more positive than at operating te.tperature, startup and d operation of the reactor when reactor coolant temperature is less than 525'F is prohibited except where necessary for low power physics tuts.

)i j Proposed

. Amendment No. 3. 87, 97, 171 l 3-6 l

1

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation BA111 (Conf '"

The poten - , 'vity insertion due to the moderator pressure coefficient that coult '

om depressurizing the coolant from 2185 psis to saturatic 'n '

of 885 psia is approximately 0.1 percent Ak/k.

During physi .ests, special operating precautions will be taken. In addition, the strong negative Doppler coefficient (1) and the small integrated Ak/k would limit the magnitude of a power excursion resulting from a reduction of moderator density.

The requirement that the reactor is not to be made critical below DTT + 10'F provides increased assurances that the proper relationship between primary coolant pressure and temperatures will be maintained relative to the NDTT of che primary coolant system. Heatup to this temperature will be accomplished by operating the reactor coolant pumps. The DTT at Beginning of Life (BOL) for the most limiting component in the reactor coolant system is less than ;

+100'F. l l

If the shutdown margin required by Specification 3.5.2 is maintained, there i is no possibility of an accidental criticality as a result of a decrease of coolant pressure.

The requirement for pressurizer bubble formation and specified water level when the reactor is less than 1 percent subcritical will assure that the reactor coolant system cannot become solid in the event of a rod withdrawal accident or a start-up accident and that the water level is above the minimum detectable level.

The requirement that the safety rod groups be fully withdrawn before criticality ensures shutdown capability during startup. This does not prohibit rod latch confirmation, i.e., withdrawal by group to a maximum of 3 inches withdrawn of all seven groups prior to safety rod withdrawal.

REFERENCES (1) USAR, section 3

)

i Proposed A7endment No. 87, 97, 171 3-7

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1.5 CHEMISTRY Aeolicability Applies to the limiting conditions of reactor coolant chemistry for continuous l operation of the reactor.

Obiective To protect the reactor coolant system from the effects of impurities in the reactor coolant.

Soecification 3.1.5.1 The following limits shall not be exceeded for the listed reactor coolant conditions.

Contaminant Specification Reactor Coolant Conditions Oxygen as 02 0.10 ppm max above 250*F Chloride as Cl- 0.15 ppm max above cold shutdown conditions Fluoride as F- 0.15 ppm max above cold shutor 1 conditions 3.1.5.2 During operation above 250'F, if any of the specifications in 3.1.5.1 are exceeded, corrective action shall be initiated within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

If the concentration limit is not restored within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after initiation of corrective action, the reactor shall be placed in a cold shutdown condition using normal procedures.

3.1.5.3 During operations between 250'F and cold shutdown conditions, if the chloride or fluoride specifications in 3.1.5.1 are exceeded, corrective action shall be initiated within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to restore the normal operating limits. If the specifications are not restored within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months af ter initiation of corrective action, the reactor '

shall be placed in a cold shutdown condition using normal procedures.

3.1.5.4 If the oxygen concentration and either the chloride or fluoride corcentration of the primary coolant system exceed 1.0 ppm the reuctor shall be immediately brought to the hot shutdown condition using normal shutdown procedures, and action is to be taken immediately to return the system to within normal operation specifications. If specifications given in 3.1.5.1 have not been reached in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , the reactor shall be brought to a cold shutdown condition using normal procedures.

Proposed Amendment No. 87, 171 3-10

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1 RANCHO SECO UNIT 1 <

TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1.8 LOH P0HER PHYSICS TESTING RESTRICTIONS Soecification The following special limitations are placed on low power physics testing.

3.1.8.1 Reactor Protective System Requirements A. Below 1820 psig shutdown bypass trip setting limits shall apply in accordance with Table 2.3-1.

B. Above 1900 psig nuclear overpower trip shall be set at a maximum of 5.0 percent.

3.1.8.2 Startup rate rod withdrawal hold shall be in effect at all times.

3.1.8.3 During low power physics testing, the minimum reactor coolant temperature for criticality shall be 240*F. A minimum shutdown margin of 1 percent ak/k shall be maintained with the highest worth control rod fully withdrawn.

Bate.1 The above specification provides additional safety margins during low power physics tasting. The startup rate rod withdrawal hold is described in paragraph 7.2.2.1.3 of the USAR and applies to the source and intermediate power ranges.

l a

l l

Proposed Amendment No. 87, 171 3-15b ,

I i

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RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.2.2 The Low Temperature Overpressure Protection System will require the following conditions:

3.2.2.1 LTOP will be manually enabled prior to the RCS temperature dropping below 350*F during plant cooldown.

3.2.2.1.1 Hanual enabling of LTOP shall be accomplished by placing the Key Select Switch for the EH0V to the "Low" position and by placing HV-21505 in the "Open" position, t 3.2.2.2 All HPI Systems will be locked out whenever the RCS temperature is below 350*F. This shall be done by removing power from the four HPI Motor-Operated Valves (Loop A: i SFV-23809, SFV-23811; and Loop B: HV-23801, SFV-23812) '

with the valves in the closed position, except as specified below:

1P 3.2.2.2.1 During Hakeup, Letdown and Purification System startup, any of the four HPI HOVs specified in 3.2.2.2 above, which are performing the required isolation function, may be individually opened as required to fill and vent the respective injection line subject to the conditions stated in 3.2.2.2.2 and 3.2.2.2.3 below:

3.2.2.2.2 All power shall be removed from the HPI-Hakeup pumps l during the filling and venting specified in 3.2.2.2.1.

3.2.2.2.3 Only one of the four HPI HOVs specified in 3.2.2.2 above may be open at any one time, and it shall be closed and power removed from the motor operator upon l completion of its line's filling and venting.

3.2.2.3 The makeup tank water level is to be less than 86 inches.

3.2.2.4 The pressurizer water level will be maintained at or below 220 inches at system pressures above 100 psig and less than 275 inches for pressures less than or equal to 100 psig except during RCS filling and dra aing.

3.2.2.5 Whea the RCS temperature is below 350'F, the core flood tank discharge valves are closed and power is removed from f the motor operators before the RCS pressure is decreased to .

600 psig, except as specified below: I 1

1 1

Proposed f Amendment No. 82, 90, 171 3-17a l

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Bufti (Continued)

Concentration of boron in the concentrated boric acid storage tank may be higher than the concentration which would crystallize at ambient conditions.

For this reason and to ensure that a flow of boric acid is available when needed, this tank and its associated piping will be kept above 70*F (30'F above the crystallization temperature for the concentration present). Nnce in the high pressure injection system, the concentrate is sufficiently well mixed and diluted so that normal system temperatures ensure boric acid solubility. The value of 70*F is significantly above the temperature for a solution containing 12,200 ppm boron.(5) crystallization The Low Temperature Overpressure Protection (LTOP) System consists of both an active and a passive subsystem. The active subsystem utilizes the Electro-Matic Operated Valve (EMOV) which provides overpressure protection during normal plant operation. The EMOV actuation circuitry has been modified to provide a second setpoint (500 psig) that is used during low-temperature operations. The low setpoint is manually enabled at 350'F by positioning a key-operated switch in the reactor Control Room. An alarm will sound in the reactor Control Room if the reactor coolant pressure falls below 450 psig and the key-operated switch is not selected for low-tenperature operation. After selection of low temperature operation, additional alarms will occur if power is not removed from each HPI pump; if either Seal l Injection flow is greater than 42 gpm or makeup flow is greater than 135 gpm; if HPI valves are not closed; and if the EMOV block valve HV-21505 is not open. The passive subsystem is based on the plant design and operating philosophy that precludes the plant from being in a water solid condition (except for system hydrotests).

Injection line filling and venting is accomplished without an energized Hakeup or High Pressure Injection pump. Nitrogen overpressure in the Makeup tank provides the motive force for fluid flow during this evolution. This nitrogen overpressure is limited procedurally to 30 psig, and mechanically to 100 psig by the relief valve on the Makeup tank. This pressure clearly does not approach the EMOV low setpoint.

The exception to core flood discharge valves being closed and their motor operator deenergized is based on the Core Flood Tank pressure being reduced to less than the EMOV low setpoint of 500 psig. Hithout the potential for i RCS overpressurization, the valve need not be administrative 1y required to be closed. This provides for other surveillanco testing and maintenance on the Core Flood System.

l Proposed l Amendment No. 82, 87, 90, 171 3-18a .

l l

1

. l RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation BAlfti The requirements of Specification 3.3.1 assure that, before the reactor can be made critical, adequate safety features are operable. Two high pressure injection pumps and two decay heat removal pumps are specified. However, only one of each is necessary to supply emergency coolant to the reactor in the event of a loss-of-coolant accident. Both core flooding tanks are required the core.l 4}da single core flood tank has insufficient inventory to reflood The borated water storage tank is used for two purposes:

A. As a supply of borated water for accident conditions.

B. As a supply of borated water for flooding the fuel transfer canal during refueling operation.(2) 390,000 gallons of borated water are supplied for emergency core cooling and Reactor Building spray in the event of a loss of core coolant accident.

This amount fulfills requirements for emergency core cooling. The borated 1

water storage tank minimum volume of 390,000 gallons is based on refueling volume requirements. Heaters maintain the borated water supply at a temperature to prevent freezing. The boron concentration is set at the amount of boron required to maintain the core 1 percent subcritical at 70*F without any control rods in the core. This concentration is 1585 ppm boron while the minimum value specified in the tanks is 1,800 ppm boron.

The requirement that one BWST isolation valve shall be open assures a static head to the injection pump not lined up to the makeup tank.

The post accident Reactor Building cooling may be accomplished by two spray units or by a combination of two emerge :cy cooling units and one spray unit. The specified requirements assure that the required post accident components are available.

The spray system utilizes common suction lines with the decay heat removal system. If a single train of equipment is removed from either system, the other train must be assured to be operable in each system.

l l

Proposed Amendment No. 4, 0/dd//ddfdd/4/20/81, 61, 76, 87, 97, 171

. 3-22 1

- . - - - . . -.-_ - .,, - _ _ _ _ _ _ ._ .l

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RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.4 STEAM AND P0HER CONVERSION SYSTEM Acolicability Applies to the operability of the turbine cycle during normal operation and for the removal of decay heat.

Obiective To specify minimum conditions of the turbine cycle equipment necessary to assure the required steam relief capacity during normal operation and the capability to remove decay heat from the reactor core.

Soecification 3.4.1 The reactor coolant system shall not be brought or remain above 280*F through HOT SHUTDOHN when irradiated fuel is in the pressure vessel unless the following conditions are met:

r A. Capability to remove decay heat by use of two steam generators as specified in 3.1.1.2. A.

i

. B. One atmospheric dump valve per steam generator shall be operable.

C. A minimum of 250,000 gallons of water shall be available in the condensate storage tank.

D. Two main steam system safety valves are operable per steam generator.

E. Both auxiliary feedwater trains (i.e., pumps and their flow paths) are operable.

F. Both trains of main feedwater isolation on each main feedwater line are operable, i

G. Four independent backup instrument air bottle supply systems for ADVs and MFH, SFH, and AFH valves are operable.

) Hith less than the above required components operable, be on decay heat cooling within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Proposed Amendment No. 93, 171 j 3-23

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.4.2 The reactor shall not enter or remain in the STARTUP through POWER OPERATION mode unless the following conditions are met:

A. Capability to remove decay heat by v e of two steam generators as specified iri 3.1.1.2.

B. One atmospheric dump valve per steam generator shall be operable except that: (1) with only one atmospheric dump valve operable, restore an inoperable valve for the other steam generator within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in hot shutdown within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and on decay heat cooling within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; (2) with no atmospheric dump valves operable, restore at least one inoperable valve within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and on decay heat cooling within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months C. A minimum of 250,000 gallons of water shall be available in the condensate storage tank except that with less than the minimum volume, restore the minimum volume within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and on decay heat cooling within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D. Seventeen of the eighteen main steam safety valves are operable except that with less than the minimum number of valves, restore the inoperable valve (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and on decay heat cooling within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

E. Four turbine throttle stop valves are operable except that with less than the minimum number of valves, restore the inoperable valve (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and on decay heat cooling within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

F. Both auxiliary feedwater trains (i.e., pump and their flow ,

path) are operable except that: :

l (1) Hith one auxiliary feedwater train inoperable, restore i the train to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in hot :

shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and on decay heat '

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

(2) Hith both auxiliary feedwater trains inoperable, the reactor shall be made subcritical within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and the reactor shall be on decay heat cooling within the

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

j Proposed '

j Amendment No. 93, 171

! 3-23a I

l RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation The 250,000 gallons of water in the condensate storage tank is sufficient to remove decay heat (plus Reactor Coolant pump heat for two pumps) for approximately 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months . This volume provides sufficient water to remove the decay heat for approximately 5.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and to subsequently cool the plant to the DHR system pressure at a cooldown rate of 50*F/hr.(l)

The minimum relief gapacity of seventeen steam system safety valves is 13.329,163 lb/hr.(2> This is sufficient capacity to protect the steam system under the design overpower condition of 112 percent.(3)

The operability of the turbine throttle stop valves ensures that no more than one steam generator will blow down in the event of a main steam line break.

Both trains of main feedwater isolation on each main feedwater line are required to be operable. Train A of main feedwater isolation is comprised of main feedwater control valves, main feedwater block valves and startup control valves. Train B of main feedwater isolation is comprised of the main feedwater isolation valves.

Four independent Class I backup air supply systems are provided to assure power available to certain air operated valves in the event of the loss of normal air supply. One system supplies power for the MFH, Startup Feedwater (SFH) and AFH control valves feeding the "A" OTSG; another system supplies power for same valves feeding the "B" OTSG. Two systems supply power for ADVs with one for the ADVs on the "A" main steam line and one for the ADVs on the "B" main steam line. Each system is sized to provide at least 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of air supply.

REFERENCES (1) B and W Document 32-1141727-00, "Heat Removal Capability of SHUD j CST," Harch 1984.

(2) USAR paragraph 10.3.4 (3) USAR Appendix 3A, Answer to Question 3A.5 (4) B and H Calculation 86-1167930, "Rancho Seco: AFH Hinimum Flow Analysis" (SHUD Calculation No. Z-FHS 10150) l l

l I

l Proposed Amendment No. 31, 87, 93, 171 3-24a

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Li,miting Conditions for Operation Hain (Continued)

Each reactor protection channel key operated shutdown bypass switch is provided with alarm and lights to indicate when the shutdown bypass switch is being used. There are four shutdown bypass keys in the control room under the administrative control of the shift supervisor. The keys will not be used during reactor power operation.

There are fcur reactor protection channels. Normal trip logic is two out of four. Required trip logic for the power range instrumentation channels is two out of three. The EFIC trip logic is two times one-out-of-two taken twice. Minimum trip logic on other instrumentation channels is one out of two.

The EFIC system is designed to automatically initiate AFH when:

1. all four RC pumps are tripped,

2. RPS has tripped the reactor on anticipatory trip indicating loss of main feedwater,

3. the level of either steam generator is low,

4. either steam generator pressure is low, or

5. SFAS ECCS actuation (high RB pressure or low RCS pressure).

The EFIC system will isolate mqin feedwater to any steam generator when the pressure goes below 600 psig.(') l The EFIC system is also designed to isolate or feed AFH according to the following logic:

If both SGs are above 600 psig, supply AFH to both SGs

- If one SG is below 600 psig, supply AFH to the other SG

- If both SGs are below 600 ps thetwoSGsexceeds100psid{g)butthepressuredifferencebetween I

, supply AFH only to the SG with l the higher pressure If than both SGs 100 are)below psid(l , supply600 AFHpsig andSGs to both the pressure difference is less l

At cold shutdown conditions all EFIC initiate and isolate functions are manually or automatically bypassed. When pressure in both steam generators is greater than 750 psig, the following bypassed initiation signals will have been automatically reset: 1) Loss of 4 RC pumps, 2) low steam generator pressure, 3) low steam generator level.

Note (l): Pressure setpoint tolerances are 600 1 25 psig and 100 1 50 psid. l Proposed Amendment No. 31, 93, 171 3-26 t

RANCHO SECO UNIT 1

' TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.5.2 CQa:rol Rod Group and Power Distribution Limits Acolicabilitt This specification applies to power distribution and operation of control rods during power operation.

Obiective To assure an acceptable core power distribution during power operation, to set a limit on potential reactivity insertion from a hypothetical control rod ejection, and to assure core subcriticality after a reactor trip.

Soeci ficattQD 3.5.2.1 The available shutdown margin shall be not less than 1% ak/k with the highest worth control rod fully withdrawn. If the shutdown margin is less than 1% ak/k then, within one hour, initiate and continue boration until the required shutdown margin is established.

3.5.2.2 Operation with inoperable rods:

Operation with more than one inoperable rod as defined in Specification 4.7.1 and 4.7.2.3 in the safety or regulating rod banks shall not be permitted.

Action

a. If a control rod in the regulating and/or safety rod banks is declared inoperable in the withdrawn position as defined in ,

Specification paragraph 4.7.1.1 and 4.7.1.3, an evaluation shall be 1 initiated immediately to verify the existence of 1% ak/k hot shutdown margin. Boration may be initiated to increase the available rod worth either to compensate for the worth of the i inoperable rod or until the regulating banks are fully withdrawn, l whichever occurs first.

b. If within one hour of determination of an inoperable rod as defined in Specification 4.7.1, it is not determined that a 1% ak/k hot shutdown margin exists combining the worth of the inoperable rod with each of the other rods, the reactor shall be brought to the hot shutdown condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months unless this margin is established.

c. Following the determinatior of an inoperable rod as defined in l Specification 4.7.1, all rods shall be exercised by a movement i until indication is noted but not exceeding 2 inches within 24 I hours and exercised weekly until the rod problem is solved.

d. If a control rod in the regulating or safety rod groups is declared inoperable per 4.7.1.2, power shall be reduced within one hour to 60% of the thermal power allowable for the reactor coolant pump combination, and within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the Nuclear Power trip setpoint shall be reduced to less than or equal to 70% of the thermal power allcwable for the reactor coolant pump combination.

Proposed

Amendment No. 48, 97, 171 3-31

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation

e. If a control rod in the regulating or axial power shaping groups is l declared inoperable per Specification 4.7.1.2, operation above 60% of rated power may continue provided the rods in the group are positioned such that the rod that was declared inoperable is maintained within allowable group average position limits of Specification 4.7.1.2 and the withdrawal limits of Specification 3.5.2.5.C.

3.5.2.3 The worth of a single inserted control rod shall not exceed 0.65 percent ak/k at rated power or 1.0 percent ak/k at hot zero power except for physics testing when the requirement of Specification 3.1.8 shall apply.

3.5.2.4 Quadrant Power Tilt A. With the Quadrant Power Tilt determined to exceed 4.92% but less than or equal to 11.07% except for physics test.

1. Within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:

a) Either reduce the quadrant power tilt to 14.92%, or b) Reduce thermal power so as not to exceed thermal power, including power level cutoff, allowable for the reactor coolant pump combination, less at least 2% for each 1%, or fraction thereof, of quadrant power tilt in excess of 4.92%. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , take action to reduce the high flux trip and flux-a flux-flow trip setpoints at least 2% for each 1%, or fraction thereof, of quadrant power tilt in excess of 4.92%.

2. Verify that the Quadrant Power Tilt is 14.92% within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding that limit or reduce Thermal Power to less than 60% of Thermal Power allowable for the reactor coolant pump combination within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce J the High Flux Trip Setpoint to 165.5% of Thermal Power I allowable for the reactor coolant pump combination within {

the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3. Identify and correct the cause of the out of limit condition prior to increasing Thermal Power; subsequent Power Operation above 60% of Thermal Power allowable for the reactor coolant pump combination may proceed provided that the Quadrant Power Tilt is verified 14.92% at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95% or greater Rated Thermal Power.

1 Proposed I 4 Amendment No. 26, 87, 171 l 3-32 l l

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation The rod position limits are based on the most limiting of the following three criteria: ECCS power peaking, shutdown margin, and potential ejected rod worth. Therefore, compliance with the ECCS power peaking criterion is ensured by the rod position limits. The minimum available rod worth, consistent with the rod position limits, provides for achieving hot shutdown by reactor trip at in any the time, assuming full out position.the(highest

1) The worth controllimits rod position rod that is ensure also withdrawn thatremains inserted rod groups will not contain single rod worths greater than 0.65%

ok/k at rated power. These values have been sh9wn to be safe by the safety analysis of hypothetical rod ejection accident.s2) A maximum single inserted control rod worth of 1.0% ak/k is allowed by the rod position limits at hot zero power. A single inserted control rod worth of 1.0% ok/k at beginning-of-life, hot zero power would result in a lower transient peak thermal power and, therefore, less severe environmental consequences than an 0.65% ak/k ejected rod worth at rated power.

Control rod groups are withdrawn in numerical sequence beginning with Group

1. Groups 5, 6 and 7 are overlapped 25 percent. The normal position at power is for Group 7 to be partially inserted.

The Quadrant Power Tilt limits set forth in Specification 3.5.2.4 have been established to prevent the linear heat rate peaking increase associated with a positive quadrant power tilt during normal power operation from exceeding 7.36%. The limits in Specification 3.5.2.4 are measurement system independent. The actual operating limits, with the appropriate allowance for observability and instrumentation errors, for each measurement system are defined in the station operating procedures.

The Quadrant Tilt and axial imbalance monitoring in Specifications 3.5.2.4F and 3.5.2.6, respectively, normally will be parformed in the process computer. The two-hour frequency for monitoring these qualities will provide ,

adequate surveillance when the computer is out of service.

Allowance is provided for withdrawal limits and reactor power imbalance limits to be exceeded for a period of two hours without specification violation.

Acceptable rod positions and imbalance must be achieved within the two-hour time period or appropriate action such as a reduction of power taken.

Operating restrictions are included in Technical Specifications 3.5.2.50.1 and 3.5.2.50.2 to prevent excessive power peaking by transient xenon. The xenon reactivity must either be beyond the "undershoot" region and asymptotically approaching its equilibrium value at rated power or the reactor must be operated in the range of 87% to 92% of the maximum allowable power for a period exceeding two hours in the soluble poison control mode so that the transient peak is burned out at a lower power level.

REFERENCES (1) USAR, Section 3.2.2.1.3 (2) USAR, Section 14.2.2.4 (3) BAH-1850, October 1984, page 7-5 Proposed Amendment No. 26, 69, 87, 171 3-33b

.~

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Table 3.5.5-1 (Continued)

Action I. With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, either restore the inoperable Channel (s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or:

1) Initiate the pre-planned alternate method of monitoring, and

2) Prepare and submit a Special Report to the Commission pursuant l to Specification 6.9.5D within 30 days following the event, outlining the action taken, the cause of the inoperability, and the corrective action and schedule for implementation.

II. a. With the number of OPERABLE accident monitoring instrumentation channels less than the Total Number of Channels, restore the inoperable channel (s) to OPERABLE status within 30 days, or be in at least HOT SHUTDOHN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Number of Channels Operable, restore the inoperable channel (s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

III. With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Number of Channels Operable, restore the inoperable channel (s) to OPERABLE status within 30 days, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

1 Proposed Amendment 80, 93, 171 3-38f

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.5.6 (Continued) 831e.1 The EFIC system is designed to automatically initiate AFH when:

1. all four RC pumps are tripped, or

2. RPS has tripped the reactor on anticipatory trip indicating loss of main feedwater, or

3. the level of either steam generator is low, or

4. either steam generator pressure is low, or

5. SFAS ECCS actuation (high RB pressure or low RCS pressure).

The EFIC system will initiate main feedwater isolation to any steam generator as the pressure goes and stays below a minimum setpoint of 575 psig.

The EFIC system is also designed to isolate or feed AFH according to the following vect9[ logic. Setpoints are nominal and subject to instrument inaccuracles: tis

- If both SGs are above 600 psig, supply AFH to both SGs

- If one SG is below 600 psig, supply AFH to the other SG

- If both SGs are below 600 psig but the pressure difference between the two SGs exceeds 100 psid. supply AFH only to the SG with the l higher pres 59re l

- If both SGs tre below 600 psig and the pressure difference is less than 100 psid, supply AFH to both SGS l At cold shutdown conditions all EFIC automatic initiate and isolate functions are manually or automatically bypassed. Prior to a pressure of greater than 750 psig in both steam generators, the following bypassed initiation signals automatically reset: 1) Loss of 4 RC pumps, 2) low steam generator pressure, 3) low steam generator level.

Bypassing of automatic AFH initiation on Less of HFH Anticipatory Trip or SFAS actuation is controlled by bypass permissive logic within the RPS and SFAS, respectively.

Note (1): Pressure setpoint tolerances are 600 1 25 psig and 100 1 50 psid. l l

Proposed Amendment No. 93, 171 3-38h

- _ _ , g . _ _ _ _

l RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation E11Ri This specification provides a method to ensure that the Spent Fuel Pool bulk temperature does not reach the boiling point. The use of a train of the i Decay Heat Removal System (DHRS), as required by operating procedures, l provides immediate alternate cooling capability to ensure this. Either train of the DHRS can easily be lined up for Spent Fuel Pool cooling by opening two manual valves (DHS-032 and DHS-055 or 056), one motor operated valte (HV-26047 or 46), and starting the appropriate decay heat pump (P-261A or 8).

However, since use of the DHRS train for Spent Fuel Pool cooling effectively removes it from its normal service, an operating duration limit of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months per 12-month period is imposed.

The arrangement of spent fuel storage racks provides a minimum of 23 feet of water shielding over stored fuel assemblies to limit radiation at the surface of the water to no more than 2.5 mrem /hr during the storage period. 37 feet of water in the Spent Fuel Pool ensures that at least 23 feet of water is maintained over the top of the irradiated fuel assemblies (active fuel) seated in the storage racks.

REFERENCES (1) Licensing Report for High Density Spent Fuel Storage Racks for Rancho Seco.

[2] Time to Boil Calculation, Supplement No. 2 to Thermo-Hydraulic Calculations for Rancho Seco Nuclear Station; Report No. TH-661. ,

Proposed Amendment No. 84, 97, 171 3-46b

o RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation TABLE 3.14-1 FIRE DETECTION INSTRUMENTS FOR SAFETY SYSTEMS Detection Number of detectors in zones-Zone Instrument Location All Required to be OPERABLE i Heat Flame Smoke l 4 Control and Computer Room Cabinets 0 0 48 5 South East Uncontaminated Area - 0 0 3 r Turbine Deck, Auxiliary Building 7 Contaminated Area - Turbine Deck, 0 0 3 Auxiliary Building '

11 Hest Battery Room - Hezzanine Level. 0 0 8 Auxiliary Building i

12 Hest Battery Chargar Room - 0 0 4 Mezzanine Level, Auxiliary Building 13 Hest 480V Switchgear Room - 0 0 B Mezzanine Level, Auxiliary Building 14 West Cable Shaft - Auxiliary Building 0 0 6 15 East Cable Shaft - Auxiliary Building 0 0 6 16 East 480V Switchgear Room - 0 0 6 :

Mezzanine Level, Auxiliary Building l 17 East Battery Charger Room - 0 0 4 Mezzanine level, Auxiliary Building 19 Communication Room - Hezzanine Level. 0 0 8 Auxiliary Building 20 Electrical Penetration Room - 0 0 21 Mezzanine Level, Auxiliary Building 21 Reactor Building, 20 feet to 40 feet 0 0 3 i

Level I

l Proposed Amendment No. 35, 85, 171 1 3-55 l

~. ,

l RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation TABLE 3.14-1 i

. FIRE DETECTION INSTRUMENTS FOR SAFETY SYSTEMS (Continued)

Number of detectors in zones- !

Detection Zone Instrument Location All Required to be OPERABLE Heat Flame Smoke ;

44 Reactor Building, O feet to 20 feet 0 4 0 ;

i Level I 45 Electrical Penetration Area - 0 0 4 Reactor Building 46 Hut Area - Basement Level, 0 0 25 i Auxiliary Building 47 Pipe Penetration Area - Basement 0 0 18 Level Auxiliary Building 48 Decay Heat Room - Auxiliary Building 0 0 3 53-54 Turbine Area 8 0 0 75 NSEB Switchgear Room B A 0 4 l 76 NSEB Switchgear Room A 4 0 4 77 NSEB B Electrical Equipment 2 0 3 ;

78 NSEB A Electrical Equipment 2 0 3 P

79 NSEB Battery GB 1 0 1 80 NSEB Battery GA 1 0 1 81 NSEB B Cable Shaft 6 0 6 82 NSEB A Cable Shaft 6 0 6 j 83 NSEB Corridor, I foot Level 0 0 2 l 84 NSEB Corridor, 21 feet Level 0 0 4 l

85 NSEB Corridor, 40 feet Level 0 0 6 Proposed Amendment No. 85, 171 l 3-55b

)

i RANCHO SECO UNIT 1 i TECHNICAL SPECIFICATIONS i

Limiting Conditions for Operation 3.15 RADIOACTIVE LIQUID EFFLUENT HONITORING INSTRUMENTATION The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.15-1 shall be OPERABLE with their alarm / trip setpoir.ts set to ensure that the limits of Specification 5.17.1 ara not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with the methodology contained in the OFFSITE DOSE CALCULATION MANUAL (00CH). j Aeolicability During releases via the retention basin effluent discharge. i Action a. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than a value which will ensure that the limits of Specification 3.17.1 are met, immediately suspend the release of radioactive liquid

- effluents monitored by the affected channel, or declare the i j channel inopc able, or change the setpoint so it is acceptably '

4 conservative. ,

i

b. With less than the minimum number of radioactive liquid ;

effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.15-1. '

RAlti ;

Ouring normal operations, all radioactive contaminated water from primary i system leaks and drains (except demineralized reactor coolant as noted below) ,

is processed in a 11guld radwaste system and recycled into the Reactor Coolant Hakeup System or otherwise reused in the controlled areas of the plant.

Secondary system water is normally released from the plant.

If the secondary system water contains radioactive material it is first processed through the 'A' or 'B' Regenerant Hold-Up Tanks (RHUTs) and then transferred to the North or South Retention Basin. The water in a Retention Basin is released offsite as a batch release. These releases are monitored by i the Retention Basin Effluent Discharge Monitor. During periods of primary to 4 secondary leakage, or when the sumps are contaminated, administrative controls require the liquid effluent in Turbine Building sumps shall be diverted to the

'A' 'B' Regenerant Hold-Up Tanks.

and Demineralized reactor coolant can be transferred from the Demineralized Reactor Coolant Storage Tank (DRCST) to the 'A' and 'B' Regenerant Hold-Up Tanks for sampling, processing, and eventual discharge offsite as required by operational constraints.

Under normal conditions, the once through steam generators have no blow down.

4 If a blow down is required during periods of primary to secondary leakage, all l water will be retained and processed in the radwaste system or diverted to the

'A' and 'B' Regenerant Hold-Up Tanks.

l l

Proposed Amendment No. 53, 98, 171 3-60

~.

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.16 RADI0 ACTIVE GASEOUS EFFLUENT HONITORING INSTRUMENTATION The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.16-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.18.la are not exceeded. The alarm / trip ,

setpoints of these channels shall be determined in accordance with the '

methodology contained in the 00CH. Continuous samples of the gaseous effluent l for radiotodines and radioactive particulate material shall be taken as '

indicated in Table 3.16-1.

Apolicability At all timos.

Action a. With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than a value

which will ensure that the limits of Specification 3.18.la are

met, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative,

b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.16-1. Exert best efforts to return the t

instrument to OPERABLE status within 30 days and, if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report submitted pursuant to Specification 6.9.2.3 why the inoperability was not corrected in a timely manner.

HAin The radioactive gaseous effluent monitoring instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of radioactive gaseous l effluents. The alarm / trip setpoints for these instruments shall be calculated :

in accordance with the methodology contained in the 00CM to ensure that the !

alarm / trip will occur prior to exceeding the limits of Specification 3.18.la. l The OPERABILITY and use of this instrumentation is consistent with the -

requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

The Auxiliary Building Stack is the effluent release point for the Haste Gas System. The Auxiliary Building Stack Noble Gas Activity monitor will perform !

the necessary Haste Gas System release termination. The monitor alarms and terminates a Haste Gas Decay Tank release automatically if the activity exceeds the setpoint limits, i

i Proposed Amendment No. 53, 98, 171

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.18.4 Gaseous Radwaste Treatment The Haste Gas System and the VENTILATION EXHAUST TREATHENT SYSTEM shall be OPERABLE. The appropriate portions of these systems shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected air doses due to gaseous effluent releases, averaged over 31 days, would exceed 0.2 mrad for gamma l ,

radiation and 0.4 mrad for beta radiation. The appropriate portions of the Ventilation Exhaust Treatment System shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to ge,seous effluent releases from the site (see Figure 5.1-3), when averaged over 31 days, would exceed 0.3 i mrem to any organ.

Apolicability I

When Gaseous Radwaste Treatment System and/or Ventilation Exhaust Treatment System are not being used. ;

Action I

a. With gaseous waste being discharged without treatment and in excess

, of the above limits, prepare and submit to the Commission within 30 days a Special Report pursuant to Specification 6.9.5 which includes the following information 1

l

1. Explanation of why gaseous radwaste was being discharged without treatment, and identification of the equipment or subsystems not OPERABLE and the reason for inoperability.

2. Action (s) taken to restore the inoperable equipment to OPERABLE status.

3. Summary description of action (s) taken to prevent a recurrence, j The OPERABILITY of the Waste Gas System and the VENTILATION EXHAUST TREATHENT i SYSTEM ensures that the systems are available for use whenever gaseous '

effluents require treatment prior to release to the environment. The j requirement that the appropriate portlons of these systems be used, when !

specified, provides reasonable assurance that the releases of radioactive ;

matorisis in gaseous effluents are maintained "as low as is reasonably i achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems are the dose design objectives set forth in Sections II.B and II.C of Appendix I.

10 CFR Part 50, for gaseous effluents.

Proposed Amendment No. 53, 98, 171 3-78 i

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.18.5 Gas Storage Tanks the quantity of radioactivity contained in each waste gas decay tank shall be limited to less than or equal to 135,000 curies of noble gases (considered as Xe-133).

Apolicability At all times Action

a. When the Reactor Coolant System activity reaches the limit of Specification 3.1.4, sample the on line waste gas decay tank daily to ensure that the 135,000 curie equivalent Xe-133 limit is not exceeded.

b. With the quantity of radioactive material in any waste gas decay tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit, and describe the events leading to this condition in the next Semlannual Radioactive Effluent Release Report oursuant to Specification 6.9.2.3.

Balti Restricting the quantity of radioactivity contained in each waste gas decay tank provides assurance that in the event of an uncontrolled release of the tanks contents, the resulting total body exposure to an individual at the exclusion area boundary (see Figure 5.1-1) will not exceed 500 mrem. This is consistent with Standard Review Plan 15.7.1, "Haste Gas System Failure."

Potential atmospheric releases from a waste gas decay tank are evaluated l assuming design coolant activities (see page 140-25 Vol. VIII USAR). Based on l primary coolant activity as shown in USAR Table 140-7, the decay tank is l assumed to hold the activity associated with the off-gas from one Reactor '

Coolant System degassing with no credit taken for decay.

Calculation of the limiting decay tank activity based on the coolant activity limit of Technical Specification 3.1.4 yields a maximum decay tank inventory of 98,414 curies (Ref. USAR Tablo 140-23). In order for the decay tank l inventory to reach the limiting condition for operation, coolant activity would have to exceed the Technical Specification 3.1.4 limit on coolant activity and this would require a reactor shutdown, thus pr9 venting a further increase in gaseous activity.

Therefore, it is conservative to require that the online waste gas decay tank be sampled _ daily upon reaching the Reactor Coolant System limiting activity value (43/E) to ensure the 135,000 curies equivalent Xe-133 is not eveeeded.

Once the coolant is below the limiting activity, there is no requirement to sample waste gas decay tanks except for discharging.

i l

Proposed I Amendmcnt No. 53, 98, 171 1 3-79

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS 1.imiting Conditions for Operation 3.27 Nuclear Service Electrical Building Emeraency Heating Venti 1411on and Ait Conditionina System Specification 3.27.1 Both Nuclear Service Electrical Building (NSEB) Emergency Heating Ventilation and Air Conditioning trains shall be OPERABLE.

3.27.2 The room temperatures in the NSEB shall not exceed 104'F.

Apolicability HEATUP through POWER OPERATIONS Action

a. Hith one NSEB Emergency Heating Ventilation and Air Conditioning train inoperable, restore the train to OPERABLE status within 7 days or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With both NSEB Emergency Heating Ventilation and Air Conditioning trains inoperable, restore the trains to OPERABLE l status within 3.5 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

c. If any room temperature in the NSEB exceeds 104'F, reduce the room temperature to less than 104'F within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT 00HN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

B1111 The NSEB Emergency Heating Vsntilation and Air Conditioning System is required to provide cooling to protect required electrical components in the NSEB.

Proposed Amendment No. 94, 171 3-93

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.29 Meteoroloaical Monitorina Instrumentation The meteorological monitoring instrumentation e.hannels shown in Table 3.29-1 shall be OPERABLE. ,

1 Aeolicability At all times. l Action !

With one or more required meteorological monitoring channels inoperable for more than 7 days, prepare and submit a Special Report to the Connission pursuant to Specification 6.9.5 within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status and include a description of the event in the next Semiannual Radioactivo Effluent Release Report. l 1 ELLt1 i

The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release i of radioactive materials to the atmosphere. This capability is required to ;

evaluate the need for initiating protective measures to protect the health and safety of the public. ;

Delta Temperature is obtained from calculated differences between measured

! air temperatures at 60 meters and 10 meters. With one or more air temperature monitoring channels inoperable for determining delta temperature, the standard deviation of wind direction (sigma theta) can be used to determine stability classifications.

There is only one meteorological monitoring instrument channel associated

with each parameter shown on Table 3.29-1. Each channel has two instruments (primary and secondary) associated with it. The primary and secondary instruments use common equipment for signal processing.

1 l

)

j I

Proposed Amendment No. 88, 171 3-94 i~

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.0-1 ,

t Applicability of Specifications 4.0.2 and 4.0.3. (The "NA" indicates that the provisions of Specification (s) 4.0.2 and/or 4.0.3 are not !

applicable to the sections identified.) i i

Section Specification 4.0.2 Specification 4.0.3 ;

4.1 ,_

]

) 4.2 4.3 l 4.4 N/A N/A i

1 4.5 !

.__ 4 . 6 [

1 4.7 ,

4.8 l t

4.9 l

4.10 {

4.11 !

I ,

4.12 l 4.13 4.14 # ,

4.15 4.16 l 4.17 4.18

) 4.19 N/A

__ 4.20 N/A 4.21 N/A 4.22 N/A 4.23

4.24 4.25 N/A 4.26 N/A j 1 4.27 N/A 4.28 j #The provisions of Specification 4.0.2 are not appitcable to the Subsequent Visual Inspection Period.

I J Proposed

{ Amendment No. 97, 171 .

! 4-Oa 1 1

e e ii i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.0-1 (Continued)

Section Specification 4.0.2 Specification 4.0.3 4.29 4.30 4.31

' 4.32 4.34 N/A _ N/A 4.35 a

Proposed Amendment No. 97. 171 4,

t

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards i

hs.fti Soecifications 4.0.1 through 4.0.3 establish the general reasirements '

applicable to Surveillance Requirements. These requirements are based on the Surveillance Requirements stated in the Code of Federal Regulations, 10 CFR 50.36(c)(3):

"Surveillance requirements are raquirements relating to test, calibration, or inspection to ensure that the necessary quality t of systems and components is maintained, that facility operation f will be within safety limits, and that the limiting conditions of

] operation will be met."

i Soecification 4.0.1 establishes the requirement that surveillanen must be l performed during the OPERATIONAL HODES or other conditions for which the ;

requirements of the Limiting Conditions for Operation apply unless otherwise stated in an individual Surveillance Requirement. The purpose of this j '

! specification is to ensure that surveillances are performed to verify the '

) operational status of systems and components and that parameters are within specified limits to ensure safe operation of the facility when the plant is in a PODE or other specified condition for which the associated Limiting

Conditions for Operation are applicable. Surveillance Requirements do not 4 have to be performed when the facility is in an OPERATIONAL H00E for which the requirements of the associated Limiting Condition for Operation do not apply unless otherwise specified. The Surveillance Requirements associated l with a Special Test Exception are only applicable when the Special Test Exception is used as an allowable exception to the requirements of a specification. ,

Soecification 4JL2 establishes the failure to perform a Surveillance ,

Requirement within the allowed surveillance interval, defined by the :

4 provisions of Specification 1.9, as a condition that constitutes a failure '

to meet th> OPERABILITY requirements for a Limiting Condition for i Operation. Under the provisions of this specification, systems and ,

components are assumed to be OPERABLE when Surveillance Requirements have >

been satisfactorily performed within the specified time interval. However, .

nothing in this provision is to be construed as implying that systems or '

components are OPERABLE when they are found or known to be inoperable ;

although still meeting the Surveillance Requirements. This specification also clarifles that the ACTION requirements are applicable when Surveillance j i Requirements have not been completed within the allowed surveillance '

1 interval and that the time limits of the ACTION requirements apply from the !

.. time it is identified that a surveillance has not been performed and not at i l the time that the allowed surveillance interval was exceeded. Completion of l the Surveillance Requirement within the allowable outage time limits of the ,

ACTION requirements restores complisnce with the requirements of j Specification 4.0.2. However, this does not negate the fact that the l failure to have performed the surveillance within the allowed surveillance i interval, defined by the provisions of Specification 1.9, was a violation of

l

. I i Proposed '

Amendment No. 171 4-Oc l

, . - - ,.---~,,,,_,--------~.,--g--,e. e_-

j 1

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards D31e.1 (Continued) the OPERABILITY requirements of a Limiting Condition for Operation that is subject to enforcement action. Further, the failure to perform a surveillance within the provisions of Specification 1.9 is a violation of a )

Technical Specification requirement ard is, therefore, a reportable event under the requirements of 10 CFR 30.73(a)(2)(1)(B) because it is a condition prohibited by the plant's Technical Specifications.

If the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or a shutdown is required to comply with ACTION requirements (e.g.,

Specification 3.0.3), a 24-hour allowance is provided to permit a delay in implementing the ACTIOd requirements. This provides an adequate time limit to complete Surveillance Requirements that have not been performed. The purpose of this allowanco is to permit the completion of a surveillance Lefore a shutdown is required to comply with ACTION requirements or before other remedial measures would be required that may preclude completion of a surveillance. The basis for this allowance includes consideration for plant conditions, adequate planning, availability of personnel, the time required to perform the surveillance, and the safety significance of the delay in complating the required surveillance. This provision also provides a time limit for the completion of Surveillance Requirements that becoce applicable as a consequence of H00E change imposed by ACTION requirements and for completing Surveillance Requirements that are applicable when an exception to the requirements of Specification 4.0.3 is allowed. If a surveillance is not completed within the 24-hour allowance, the time limits of the ACTION requirements are applicable at that time. When a surveillance i. performed within the 24-hour allowance and the Surveillance Requirem:ots are not met, the time limits of the ACTION requiremente are applicable at the time the surveillance is terminated.

Surveillance Requirements do not have to be performed on inoperable i

equipment because the ACTION requirements define the remedial measures that apply. However, the Surveillance Requirements have to be met to demonstrate that inoperable equipment has been restored to OPERABLE status.

SpAcification 4.0 3 establishes that all applicable surveillances must be met before entry into an OPERATIONAL H00E or other condition of operation specified in the Applicability statement. The purpose of this specification is to ensure that system and component OPERABILITY requirements or parameter i limits are mat before entry into a H0DE or condition for which these systems and components ensure safe operation of the facility. This provision applies to changes in OPERATIONAL HN ZS or other specified conditions associated with plant shutdown as well as startup.

l Proposed l / nenoaent No.171 4-Od l

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards HA111 (Continued)

Under the provisions of th!; specification, the applicable Surveillance Requirements must be perfvcmed within the specified surveillance interval to

, ensure that the Limiting Conditions for Operation are met during initial I plant startup or following a plant outage.

When a shutdown is required to comply with ACTION requirements, the provisions of Specification 4.0.3 do not apply because this would delay olacing the facility in a lower HODE of operation.

}

l Proposed Amandment No. 171 4-Oe

RANCHO SECO UNIT 1 !

TECHNICAL SPECIFICATIONS Surveillance Standards Calibration Calibration shall be performed to assure the presentation and acquisition of accurate information. The nuclear flux (power range) channels amplifiers shall be calibrated (during steady state operating conditions) against a heat balance standard whenever the Nuclear Instrumentation indication is 10 percent l or more above the core thermal power or 2 percent or more below the core i i thermal power. I Channels are subject only to "drift" errors induced within the instrumentation itself and consequently, can tolerate longer intervals between calibrations. l Process system instrumentation errors induced by drift can be expected to remain within acceptable tolera;W if recalibration is performed at the intervals of each refueling period.

Substantial calibration shifts within a channel (essentially a channel failure) will be revealed during routine checking and testing procedures. ,

l Thus, minimum calibration frequencies set forth are considered acceptable.

l Inting l The frequency of on-line testing of reactor protective channels as shown in l Table 4.1-1 will assure the required level of performance. l The equipment testing and system sampling frequencies specified in Table i.1-2 I

i andTable4.1-3areconsideregdequatetomaintaintheequipmentandsystems in a safe operational status, i Power Distribution Happing The incore instrumentation detector system will provide a means of assuring that axial and radial power peaks and the peak locations are being controlled by the provisions of the Technical Specifications within the limits employed in the safety analysis.

EfERENCES (1) USAR paragraph 1.4.12.

l Proposed l Amendment No. 78, 87, 171 4-2 l

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.1-1 (Coctinued)

INSTRUPTENT SURVEILLANCE REQUIREMENTS Channel Descriotion Check Test. Calibrate Remarks

27. Reactor Building spray NA R NA valves Channel 8 manual trip Process Instrumentation

28. Core flooding tanks

a. Pressure channels D NA R

b. Level r.hannels D NA R

29. Pressurizer level channels D NA R Equivalent to Item 49, but escludes SPDS. ,

30. Pressurizer temperature 5 NA R channels

31. Makeup tank level D NA R channels

32. High pressure injection NA NA R flow channels

33. Low pressure injection NA NA R flow channels

34. Borated water storage W NA R tank level indicator Proposed Amendment Ko. 171 47

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.1-1 (Continued)

INSTRUMENT SURVEILLANCE REQUIREMENTS Channel Description Check Test Calibrate Remarks

35. Spray additivo tank

a. Level channel W NA R

36. Concentrated boric acid storage tank

a. Level channel W NA R

b. Temperature channel M NA R

37. Steam generator water W NA R level

33. Control rod absolute 5(1) NA R(2) (1) Check with relative position position indicator (2) Calibrate rod misalignment channel

39. Control rod relative S(1) NA R(1) (1) Check with absolute posi-position tion indicator

40. Reactor Building NA NA R temperature

41. Reactor Building emergency NA M R sump level alarm Proposed Amendment No. 4J. 171 4-7a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS ,

Survallianco $ttadards .

Table 4.1-1 (Continued)

INSTRUMENT SURVEILLANCE REQUIREMENTS Channel Descriolion Check Test Calibrate Remarks

42. Reactor Building drain NA NA R accumulation tank level

43. Incore neutron detectors M(1) NA NA (1) Check functioning, including functioning of computer readout and/or recorder readout.

44. a. Process radiation monitoring system W Q R

b. Area radiation monitoring system W M Q

c. Containment Area Monitors W M 2

45. Deleted

46. a. Environmental air monitors M(1) NA R (1) Check functioning

b. Chlorine Detector W(1) M(1) R(1) (1) Only required if the total quantity of gaseous chlorine in the RESTRICTED AREA exceeds 100 pouris.

47. Strong motion accelerometer Q(1) NA R (1) Battery check

48. Deleted

49. Pressurizer Water Level M NA R

50. Auxiliary Feedwater Flow Rate M NA R

51. Spent Fuel Fool Level W(1) NA R (1) Daily during refueling when moving fuel or control rods.

52. EMOV Power Position Indicator (Frimary Detector) M NA R

53. EMOV Position Indicator (Backup Detector)

T/C or Acoustic M NA R

54. EMOV Block Valve Position Indicator M NA R

55. Safety Valve Position Indicator (Prim..ry Detecter)

T/C M NA R eroposed Amendment No. 7, 21, 54. f3, 92, 97, 92, 171 4-7b

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l i RANCKD SECO IMIT 1 -

TECHNICAL SPECIFICATIONS

] Surveillance Standards TABLE 4.1-1 (Continued)

J INSTRUMENT SURVEILLANCE REQUIREMENTS Channel Descriotion Check Test Calibrate Reshark s 1

56. Safety Valve Position In-i dicator (Backup Detector) q Acoustic M N/A R 3 57. Voltage Protection (1) Test using undervoltage relay's (Definite and Inverse "TRIP TEST." No quantitative Time Delay Relays) measurement *.

a. Undervoltage Relay M(1) N/A R Voltage Setpoint i b. Undervoltage Relay M(1) N/A R Time Delay J

a 58. Containment Area High 5 M(1) R (1) Test using installed source Range Monitor

59. Wide Range Contalemment M N/A R

, Water Level

! 60. Containment Hydrogen 5 M Q j Analyzer 4

61. Emergency Sune Level M N/A R i 62. Containment Wide Range M N/A R l Pressure Monitor / Recorder

) 63. High Range Noble Gas S M(1) R (1) Source check only l

i Effluent Monitors 1 - R8 Enhaust Stack ,

1 - Aun. Building Stack 1

- Radweste Vent i

64 Main Steam Line Radiation 5 M(1) R (1) Test us % installed source Monitors

! 65. Subcooling N rgin Monitors M N/A R j 66. Incore Thermocouples M N/A R l 67. Low Temperature Over N/A (1) R (1) Prior to cooldown

Pressure Protection l

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4 Proposed 1 Amend.ent no. a. u. sz. n. n. u. m 4-7c

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

RANOO SECO UNIT 1 .

TECHNICAL SPtCIFICATIONS .

Surve;11ance Standards TABLE 4.1-1 (Continued)

INSTRUMENT SURVEILLANCE REQUIREMENTS Chanael Descriotion Check Test Calibrate Remarks

68. AN Initiation

4. Manual NA M NA

b. Low Level SGA or B 5 M (1) R (1) (1) Include time delay module.

c. Low Pressure SGA or B 5 M (1) R (1) (1) Include time delay module. *

(2) (2) Calibrate pressure transmitters l annually. l

d. Loss of MN Anticipa-tory Reactor Trip 5 M NA

e. Loss of four RC Pumps 5 M NA

f. SFAS Actuation 5 R NA

g. Automatic Trip Logic 5 M NA

h. Bypasses 5 M R (1) (1) Calibrate pressure transmitters annually.

69. $rA Fain Feedwater Ll_ng 1191B1192

a. Manual NA M NA

b. Automatic Trip Logic 5 M NA

70. $G]LMain Feedwater Line Isolation

a. Manual NA M NA

b. Automatic Trip Logic 5 M NA Propcsed Amendment No. FJ, 171 4-7d

P>JiCHO SECO tK4IT 1 TECHNICAL SPECIFICATIONS ,

Survaillance Standards -

TAfstE 4.1-1 (Continued)

INSTRUMENT SURVEILLANCE REQUIREMENTS Channel Dessriotion Check Test Calibrate Remarks

71. 6FV Valve Conr#nds (Vector)

a. Vector Enable $ M NA

b. SGA Pressure Low 5 M R(1) (1) Calibrate pressure transmitters annually.

c. SGB Pressure Low 5 M R(1) (1) Calibrate pressure transmitters annually.

d. SG Pressure Difference SGA Pressure > S M (1) R(1)(2) (1) Include time delay module. l SGB Pressure (2) Calibrate pressure transmitters 3 annually.

SG8 Pressure > S M (1) R(1)(2) (1) Include time delay module. l SGA Pressure (2) Calibrate pressure transmitters 3 annually.

72. AFV_Contr31 Valve Control

a. Manual / Auto NA M NA in Manual

73. $G Level Control

a. Setpoint Selection NA M NA

b. Control Enable NA M NA C. Module Response NA M (1) R (1) Confirm enternal controller settings

74. ADV Control Valve Control

a. Manual / Auto NA M NA in Manual

75. SG Pressure Control

a. Module Response NA M (1) R (1) Confirm external controller settings

76. Besiva_ Instrument Air

$ypolv System

a. Pressure D NA NA Proposed Amendment No. PJ, 171 4-7e

a RANCHO SECO UNIT 1 .

TECHNICAL SPECIFICATIONS -

Surveillance Standards TABLE 4.1-1 (Continued)

INSTRUMENT SURVEILLANCE REQUIREMENTS Chamel Description Check Test Calibrate Rema rk s

, 82. Spray Pond Water Temperature D M R

83. Spray Pond Water Level D M R Emersency Shutdm Instrumentation

84. Wide Range OTSG Level M M R

85. Wide Range OTSG Pressure M M R

86. Pressurizer Level M M R

87. Wide Range Reactor Coolant Hot Leg Temperature M M R

88. Wide Range Reactor Coolant :

Cold Leg Temperature M M R I

89. Wide Range Reactor Coolant Pressure M M R l

90. Source Range Neutron Flus Indicator

M M R l

91. Makeup Tank Level M M R l

5 - Each Shift M - Monthly P - Prior to each startup if not done previous week D - Daily Q = Quarterly R - Once during the refueling interval W - Weekly SY - Semiannual

%, di zed and disconnected except when control room habitability is lost.

proposed h dment So. 77, 171 4-7g e

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Balti

! Irradiation surveillance provides the capability of deto mining the radiation-induced changes in the mechanical and impact properties in the region of the reactor vessel surrounding the core. Test specimens of base metal, deposited weld metal and the heat-affected zone are installed in capsule assemblies placed inside the vessel. In accordance with the schedules of Table 4.2-1, specimens will be removed and a series of drop weight tests, Charpy impact tests and tension tests will be conducted.

Threshold neutron flux detectors and maximum temperature detectors will be installed with the specimens. Changes in nil ductility transition

temperature will be determined, and appropriate alteration to plant operating parameters will be made.

l To assure the availability of adequate surveillance data for the Rancho Seco 4 No. I reactor vessel, a program has been developed to monitor the irradiation of the surveillance specimen capsules at the Davis Besse No. 1 ,

reactor and compare this to the irradiation of the Rancho Seco No I reactor i

vessel. Fluence estimates which are conservative in the appropriate

, direction are used for this comparison. The frequency of monitoring varies

depending on the known neutron fluence lead factor between the capsules and the reactor vessel. This provides ample time for anticipating problems and j initiating corrective action should operation of the host reactor be seriously delayed. For the purpose of Technical Specification 4.2.1.2, the l i definition of Regulatory Guide 1.16, Revision 4 (August 1975) applies for the term ' commercial operation'. Cumulative reactor utilization factor is defined as: ((Cumulative thermal megawatt hours since attainment of commercial operation at 100% power) x 100] + [(licensed thermal power) x

! (cumulative hours since attainment of commercial operation at 1001, power)).

j A preoperational examination was made which included all the items in ASME

{ Code Class I systems that would normally be completed throughout the '

inspection interval. This survey established initial system integrity and provided a baseline for future testing. .

I

Specification 4.2.2.1 ensures that inservice inspection of ASHE Code Class l

1, 2 and 3 components and inservice testing of ASME Code Class 1, 2 and 3

pumps and valves will be performed in accordance with a periodically updated version of Section XI of the ASME Boiler and Pressure Vessel Code and

Addenda as required by 10 CFR 50.55a. Relief from any of the above requirements has been provided in writing by the Commission and is not a i

part of these Technical Specifications, j

i 1

1 Proposed ;

i Amendment No. 76, 171 I 4-12 l l

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS i Surveillance Standards 4.4.1.1.5 Report of Test Results ,

Each integrated leak rate test will be the subject of a summary technical report in accordance with 10 CFR 50, Appendix J.Section V.B. and Specification 6.9.5M. The report will include any deterioration noted in the con-tainment surfaces inspection described in Specification 4.4.1.4, a description of test methods used and a summary of local leak detection tests. Sufficient data and analysis '

shall be included to show that a stabilized leak rate was attained and to identify all significant required correction factors such as those associated with humidity and barometric pressure, and all significant errors such as thosv associated l

with instrumentation sensitivities and data scatter. The report shall be submitted to the Director of Nuclear Reactor Regulation, NRC, approximately 3 months after the conduct of each test.

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i Proposed 1 Amendment No. 97, 171 1 4-16a I

e RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.4.1.3 Isolation Valve Functional Tests Remotely operated Reactor Building isolation valves shall be stroked to the position required to fulfill their safety function in accordance with requirements of Section XI of the ASHE Boller and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55(a)(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(1).

4.4.1.4 Containment Surfaces Insoection l

A visual examination of the accessible interior and exterior surfaces of the containment structure and its components shall be l performed prior to any integrated leak rate test, to uncover any l evidence of deterioration which may affect either the containment's structural integrity or leak-tightness. The

, discovery of any significant deterioration shall be accompanied by corrective actions in accordance with acceptable procedures, nondestructive tests and inspections, and local testing where practical, prior to the conduct of any integrated leak rate test.

Such repairs shall be reported as part of the test results per Specification 6.9.5H.

4.4.1.5 Reactor Buildina Hodifications Any major modification or replacement of components affecting the l Reactor Building integrity shall be followed by either an integrated leak rate test or a local leak test, as appropriate.

l and shall meet the acceptance criteria of 4.4.1.1 and 4.4.1.2.3, 1

respectively.

B1511 The Reactor Building is designed for an internal pressure of 59 psig and a steam-air mixture temperature of 286'F. Prior to initial operation, the l containment will be strength tested at 115 percent of design pressure. The containment will also be leak tested prior to initial operation at Pp and l Pt (52 psig and 26 psig, respectively). These tests will verify that the l leakage rate from Reactor Building pressur relationshipsgiveninthespecification.jptipnsatisfiesthe s i n2i The performance of a periodic integrated leakage rate test during plant life provides a current assessment of potential leakage from the containment in case of an accident that would pressurize the interior of the containment.

The minimum of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months was specified for the integrated leakage rate test l to help stabilize conditions and thus improve accuracy and to better evaluate data scatter. The frequency of the periodic integrated leakage rate test is keyed to the refueling schedule for the reactor, because these tests can best be performed during refueling shutdowns.

Proposed Amendment No. 76, 171 4-19 l

e RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards j l

The specified frequency of periodic integrated leakage rate tests is based on threemajorconsiderations. First is the low probability of leaks in the liner because of conformance of the complete containment to a 0.10 percent leakage rate at 52 psig during pre-operational testing and the absence of any ,

significant stresses in the liner during reactor operation. Second is the ;

more frequent testing at 52 psig of those portions of the containment envelope that are most likely to develop leaks during reactor operation (penetrations ,

and isolation valves) and the low value 0.06 percent of leakage that is j specificed as acceptable from penetrations and isolation valves. Third is the '

tendon stress surveillance program which provides assurance that an important part of the structural integrity of the containment is maintained. ,

I Hore frequent testing of various penetrations is specified as these locations !

are more susceptible to leakage than the Reactor Building liner due to the l mechanical closure involved. Particular attention is given to testing those l penetrations with resilient sealing materials, penetrations that vent directly j to the Reactor Building atmosphere, and penetrations that connect to the Reactor Coolant System pressure boundary. The basis for the specification of a total leakage rate of 0.075 percent from penetrations and isolation valves l

is that approximately three quarters of the allowable integrated leakage rate should be from those sources, in order to provide assurance that the integrated leakage rate would remain within the specified limits during the inter"ais between integrated leakage rate tests. Valve operability tests are I specified to assure proper closure or opening of the Reactor Building isolation valves to provide for isolation of functioning of safety features ,

systems. Valves will be stroked to the position required to fulfill their j safety function unless it is established that such testing is not practical '

during operations.

The airlock seals are tested at 10 psig because that is the manufacturer's recommended pressure for reverse flow through the seals. The extrapolation formula is derived assuming laminar, incompressible flow and provides conservative leak rates.

This specification complies with Appendix J to 10 CFR 50 as published in the 1 Federal Register on February 14, 1973, with the exemptions to Appendix J l granted July 13, 1977.

REFERENCES l

(1) USAR, paragraph 5.2.1.1.1 1 (2) USAR, Section 14 l l

l

, 1 Proposed I Amendment No. II, 87, 171 i 4-20

t RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards soecification (Continued)

B. The grease coverage will be noted along with the temperature to accumulate a record of grease variation versus temperature.

4.4.2.5 Reoorts A report covering the results of each inspection will be filed l with the plant quality assurance records. If any significant or critical deterioration is noted by this inspection, it will be reported to the NRC in accordance with Technical Specification 6.9.58. l Should this be necessary, the initial report may be made within 10 days of the completion of the tests and the detailed report may follow within 90 days of the completion of the tests.

Proposed Amendment No. 4, 171 4-23

.. RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS t

Surveillance Standards D. Nuclear Service Cooling and Raw Hater Systems

1. During each refueling interval, the safety features function of the nuclear service cooling water and raw water systems shall be tested. These tests may be in conjunction with other ECCS refueling interval tests i which require automatic actuation of these systems. i

2. The test will be considered satisfactory if control board indication verifies all components have responded to the actuation signal and all appropriate pump breakers shall have opened or closed, and all power actuated valves have completed their travel, t 4.5.1.2 Commonents Tests i

A. Testing 1

At least quarterly Inservice testing of ECCS and Nuclear Service Cooling and Raw Hater Pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure l Vessel Code and applicable Addenda as required by 10 CFR l 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(1).

j B. Flow Path Verification Following Inservice testing of pumps and valves as required by paragraph 4.5.1.2A, required flow paths shall be demonstrated operable by verifying that each valve (manual, power-actuated or automatic) in the flow path that is not locked in position is in l its normal operating position. ;

EA111 l The emergency core cooling systems are the principal reactor safeguards in !

l the event of a loss-of-coolant accident. The removal of heat from the core !

provided by these systems is designed to limit core damage. '

The decay heat removal pumps are tested singularly for operability by opening the bonted water storage tank outlet valves and the test line valves to the borated water storage tank. This allows water to be pumped from the borated water storage tank through each of the injection lines and back to the tank through a test line.

With the reactor shut down, the check valves in each core flooding line are checked for operability by reducing the reactor coolant system pressure until the indicated level in the core flood tanks verify the check valves have opened.

REFERENCES USAR subsection 6.2 Proposed Amendment No. 76, 87, 171

t RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards t 4.5.3 DECAY HEAT REMOVAL SYSTEM AND REACTOR BUILDING SPRAY SYSTEM LEAKAGE i

Acolicability

~

Applies to Decay Heat Removal System and the Reactor Building Spray System leakage. -

Objective To prevent significant offsite exposures by maintaining a preventive leakage rate for the Decay Heat Removal System and the Reactor Building Spray System.

1 i Soecification :

4.5.3.1 !

A nesttance Limit The maximum allowable leakage from the components (which include valve stems, flanges and pump seals) in the Decay Heat Removal 1 System and the Reactor Building Spray System shall not exceed a j sum total of 0.63 gallons per hour (gph) for both systems.

4.5.3.2. A leit - Decay Heat Removal System During each refueling interval, the following tests of the Decay i Heat Removal System shall be conducted to determine leakage: t i

l 1. The portion of the Decay Heat Removal System, except as i specified in (2), that is outside the containment shall be 4 tested either by use in normal operation or by hydrostatically testing at 450 psig.

3 j 2. Piping from the containment emergency sump to the decay heat

removal pump suction isolation valve shall be pressure tested i at no less than 52 psig as a containment local leak rate test

! under paragraph 4.4.1.2.

3. Visual inspection shall be made for excessive leakage from components of the system. Any excessive leakage shall be

. measured by collection and weighing or by another equivalent

! method.

4.5.3.2.B lait - Reactor Building Sorav System Ouring each refueling interval, the following tests of the Reactor j Building Spray System shall be conducted in order to determine i j leakage:

1 Proposed j Amendment No. 57, 171 4-32

I RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards ,

4.5.3.2.8 1. The section of the system that is downstream of the pump suction isolation valve shall be tested by use in normal operation or by hydrostatically testing at 180 psig.

2. The section of the system from the containment emergency sump isolation valve to the pump suction isolation valve l shall be tested at no less than 52 psig as a containment ,

local leak rate test under paragraph 4.4.1.2. ,

3. Visual inspection shall be made for excessive leakage from components of the system. Any excessive leakage shall be measured by collecting and weighing or by another equivalent method.

BAH 1 The leakage rate limit for the Decay Heat Removal System is a judgment value based on assuring that the components can be expected to operate without

. mechanical failure for a period on the order of 200 days after a LOCA. The test pressures achieved eithat by normal system operation or by hydrostati-cally testing give an adequate margin over the highest pressure within the

! system after a design basis accident. Similarly, the pressure tests for i the return lines from the containment to the Decay Heat Removal Sy tem are equivalent to the peak calculated pressure after a LOCA. A Decay Heat !

Removal System and Reactor Building Spray System sum total leakage rate of ;

4 0.63 gph will limit offsite exposures due to leakage to insignificant levels l relative to those calculated for leakage directly from the Reactor Building in the design basis accident. The dose to the thyroid calculated as a r ult of this leakage is 7.21 rem for a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months exposure at the site boundary.(p) '

j REFERENCES

(1) USAR, paragraph 14.3.9.3.

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I Proposed 1 Amendment No. 57, 68, 87, 171

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4-33 1

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.9 REACTIVITY ANOMALIES 6pp1_icability I Applies to potential reactivity anomalies. l Objective To require the evaluation of reactivity anomalies of a specified magnitude occurring during the operation of the unit.

Soecification l Following a normalization of the computed boron concentration as a function ,

of burnup, the actual boron concentration of the coolant shall be compared monthly with the predicted value. If the difference between the observed and predicted steady-state concentrations reaches the equivalent of l 1 percent in reactivity, an evaluation will be made to determine the cause of the discrepancy and reported to the Nuclear Regulatory Commission. !

EM To eliminate possible errors in the calculations of the initial reactivity i of the core and the reactivity depletion rate, the predicted relation i between fuel burnup and the boron concentration, necessary to maintain l adequate control characteristics, must be adjusted (normalized) to )

accurately reflect actual core conditions. When full power is reached initially, and with the control rod groups in the desired positions, the boron concentration is measured and the predicted curve is adjusted to this point. As power operation proceeds, the measured boron concentration is compared with the predicted concentration and the slope of the curve relating burnup and reactivity is compared with that predicted. This process of normalization should be completed after about 10 percent of the total core burnup. Thereafter, actual boron concentration can be compared with prediction, and the reactivity status of the core can be continuously evaluated. Any reactivity anomaly greater than 1 percent would be I

I unexpected, and its occurrence would be thoroughly investigated and evaluated.

The value of 1 percent is considered a safe limit since a shutdown margin of at least 1 percent with the most reactive rod in the fully withdrawn I position is always maintained. I i

Proposed l Amendment No. 87, 171 4-40

i RANCHO SECO UNIT 1 ,

TECHNICAL SPECIFICATIONS Surveillance Standards Specification (Continued) 4.11.1 B. 3. (a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly and obtaining samples at l least 2 inches in diameter with a length equal to the L I thickness of the bed, or !

l l (b) Emptying a longitudinal sample from an adsorber tray.

l mixing the adsorber thoroughly, and obtaining samples at least 2 inches in diameter with a length equal to the thickness of the bed.

4. Verifying system flow rate nf 66,700 cfm a 10% during system operation when tested in accordance with ANSI N510.

C, Verified by determining that the air distribution across the ,

l filter banks is uniform per ANSI N510 following original !

i installation, modification or repair, l

! D. Demonstrated operable after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber i operation by: '

l. Verifying within 31 days after removal that a laboratory ;

analysis of a carbon sample from at least one test canister '

or carbon sample removed from one of the charcoal adsorbers demonstrates a removal efficiency of 195% for radioactive methyl iodide when the sample is tested as used charcoal in i i accordance with ASTM D3803 (30*C, 95% R.H.). The carbon '

samples not obtained from test canisters shall be prepared ,

from either: ;

1 (a) Emptying one entire bed from a removed adsorber tray. [

mixing the adsorbent thoroughly, and obtaining samples l at least 2 inches in diameter and with a length equal to !

the thickness of the bed, or j 1 (b) Emptying a longitudinal sample from an adsorber tray. l mixing the adsorber thoroughly, and obtaining samples at i least 2 inches in diameter with a length equal to the thickness of the bed. !

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2. If an adsorber tray is removed in obtaining the charcoal I sample per Specification 4.11.10.1 above: l (a) Verify that the charcoal adsorbers remove 199.5% of a halogenated hydrocarbon refrigerant test gas when they i are tested in-place in accordance with ANSI N510 while l operating the filter train at a flow rate of 66,700 cfm i 10%.

Proposed Amendment No. 39, 171 4-42a l

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.12 AUXILIARY AND SPENT FUEL BUILDING FILTER SYSTEMS Apolicability Applies to the Auxiliary Building exhaust filter system and to the Spent Fuel Pool Building when irradiated fuel which has decayed less than 90 days is being moved or is stored in it.

Obarlt11e To verify that the Auxiliary Building exhaust filter system and components will be able to perform their design functions, i Specification 4.12.1 When irradiated fuel which has decayed less than 90 days is in the spent fuel storage pool:

A. The spent fuel storage pool building exhaust ventilation system shall be verified to be operating with all spent fuel building doors clossd (excepting intermittent personnel use) prior to fuel movement and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months during either fuel movement within the spent fuel storage pool or crane operation with loads over the spent fuel storage pool.

4.12.2 Proper operation of the ventilation system shall be:

A. Verified at least once per 31 days by observing flow through the operating HEPA filter and charcoal adsorber train and verifying that the train operates with <6 inches Water Gauge pressure drop across the combined HEPA and Charcoal filter banks and verifying system operation for at least 15 minutes.

B. Verified at least once per refueling interval, or once every 18 l months, whichever occurs first, or after each partial or complete replacement of the HEPA filter bank or charcoal adsorber bank, or following painting, fire, or chemical release in the operating air makeup system, or after any structural maintenance on the HEPA filter or charcoal adsorber housings, by: 1

1. Verifying that the charcoal adsorbers remove 199.5 percent l of a halogenated hydrocarbon refrigerant test gas when they are tested in-plice in accordance with ANSI N510 while operating the filter train at a flow rate of 43,400 cfm = l 10 percent.

2. Verifying that the HEPA filter banks removed 199.97, of the DOP when they are tested in-place in accordance with ANSI N510 while operating the filter traim at a flow rate of l 43,400 cfm : 10 percent.

Proposed Amendment 39, 171 4-43

,, RANCHO SECO UNIT 1 i TECHNICAL SPECIFICATIONS i Survalliance Standards ,

Specification (Continued) ,

4.12.2 B. 3. Verifying that a negative pressure of 10.01 inches H.G. is ;

maintained in the Spent fuel Building, with the Auxiliary ;

~

Building exhaust system operating at a flow rate

1) of 43,400 cfm a 10 percent and exhausting through the l HEPA filters and charcoal adsorbers to the facility ver.t. l and 2) of 10,800 cfm : 10 percent exhaust from the spent l t fuel pool area. ,

t

4. Verifying within 31 days after removal that a laboratory analysis of a carbon sample from at least one test canister i or carbon sample removed from one of the charcoal adsorbers -

demonstrates a removal efficiency of 195% for radioactive ,

methyl iodide when the sample is tested in accordance with ASTM D3803 (30'C, 95% R.H.). The carbon samples not obtained from test canisters shall be prepared from either:

I (a) Emptying one entire bed from a removed adsorber tray.

I mixing the adsorbent thoroughly, and obtaining samples ,

at least two inches in diameter and with a length equal l' to the thickness of the bed, or (b) Emptying a longitudinal sample from an adsorber tray. l l mixing the adsorbent thoroughly, and obtaining samples I of at least two inches in diameter and with a length

equal to the thickness of the bed, i 1

5. Verifying a system flow rate of 43,400 cfm a 10 percent during system operation when tested in accordance with ANSI !

N510.

l l C. Verified by determinin that the air distribution across the l adsorber section is un form, per ANSI N510, following original j installation, modification or major repair.

I D. Demonstrated operable af ter every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber :

operation by: l

1. Verifying within 30 days after removal that a laboratory l analysis of a carbon sample from at least one test canister i or carbon sample removed from one of the charcoal adsorbers demonstrates a removal efficiency of 2951 for radioactive ,

methyl iodide when the sample is tested in accordance with ASTM 03803 (30*C, 95% R.H.), The carbon samples not obtained from test canisters shall be prepared by either: '

(a) Emptying one entire bed from a removed adsorber tray, j mixing the adsorbent thoroughly, and obtaining samples l .

of at least two inches in diameter and with a length equal to the thickness of the bed.

(b) Emptying a longitudinal semple from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples l at least two inches in diameter and with a length equal to the thickness of the bed.

j Proposed i Amendment No If, 171 l 4-43a

.. RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS

' Surveillance Standards Soecification (Continuad) 4.12.2 0. 2. After reinstallation of the sampled adsorber tray per Specification 4.12.20.1:

(a) Verify that the charcoal adsorbers remove 199.5% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510 while operating the filter train at a flow rate of 43,400 cfm z 10 percent.

(b) Verify that the HEPA filter bank removes 199.97. of the DOP when tested in-place in accordance with ANSI N510 i while operating the filter train at a flow rate of l 43,400 cfm a 101..

E. Started on a manual signal and operated for 15 minutes in each 31-day period, if not slready operating.

i 111111 The Auxiliary Building exhaust system consists of two full capacity units arranged to take suction from a corxnon plenum, d.aw the air through HEPA and charcoal filter banks and discharge it into the plant vent. Only one unit is operated at a time, allowing the other to be serviced or held in reserve. '

This system draws all of the potentially radioactively contaminated air in the plant, external to the Reactor Building, through it. The following major areas are served:

1) Spent Fuel Building

2) Radio-Chemical Lab Hoods and Service Area

3) Radwaste Area

4) Haste Gas Discharge

5) Condenser Air Ejector Exhaust ,

6) Various Instrumentation and Sampling Discharges While providing service to these areas the filters are credited with a minimum DF of 10 for radioactive iodine and a DF of 100 for particulate matter bhich may be released in the following:

1) Letdown Line Rupture outside the Reactor Building

2) Post LOCA Decay Heat Removal Leakage

3) Dropped fuel Assembly in Spent Fuel Pool l

4) OTSG Tube Rupture l

5) Makeup Tank Rupture Releases of radioactive materials, and the resulting dosage from these accidents, are based on the maximum flow rate from the plant vent. Reduced flow rates are conservative as to the effect of plant releases. Shutdown of l

the entire system in response to a specific occurrence is likewise allowable at the operator's discretion. The negative pressure requirement for the Spent Fuel Building is to ensure that all potential releases following a dropped fuel assembly accident are drawn into the exhaust system, filtered, and monitored prior to release.

Proposed Amendment No. H , 171 4-43b i

i

! RANCHO SECO UNIT 1 l TECHNICAL SPECIFICATIONS [

Surveillance Standards Successive Insoection Intitrvals Every 10 years thereafter (or Volumetric inspection of 1/3 of nearest refueling outage) the welds at the expiration of i

each 1/3 of the inspection ;

interval with a cumulative 100

) percent coverage of all welds.

Matt - The welds selected during each inspection period shall be t

distributed among the total number to be examined to l provide a representative sampling of the conditions of 4

the welds. ;

3. E.taminations that reveal unacceptable structural defects in a :

weld during an inspection under 4.13 A.2 shall be extended to [

] require an additional inspection of another 1/3 of the welds, i i If farther unacceptable defects are detected in the second l I sampling, the remainder of the welds shall be inspected. l 4

! 4. In the event repairs of any welds are required following any ,

l examination during successive inspection intervals, the inspection schedule for the repaired welds will revert back to ,

the first 10 year inspection program.

B. For all welds in critical areas other than those identified as

! postulated break location on Figures 4.13-1, 2 and 3: i I 1. Inservice inspection shall be performed in accordance with the I provisions of paragraph 4.2 of these Technical Specifications. :

1 !

l C. For all welds in the critical areas as identified on Figures 4.13-1, !

j 2 and 3:

1. A visual inspection of the surface of the insulation at all weld >

locations shall be performed on a weekly basis for detection of i leaks. Any detected leaks shall be investigated and evaluated.

l If the leakage is caused by a through-wall flaw, either the

plant shall be shut down, or the leaking piping isolated.

l Repairs shall be performed prior to return of this line to j service.

4 i 2. Repairs, re-examination and piping pressure tests shall be conducted in accordance with the rules of ASME Code,Section XI. l

]

i J

Proposed

Amendment No. 76, 87, 171 4-45 1

i RANCHO SECO UNTT 1 TECHNICAL SPFCIrkA s j

' h ting Conditions for Operation Table 4.'t, 5 SNUBBERS ACCESSIBLE DURING P0HER OPERATIONS Snubber System Snubber Installed ,

List N L I.D. No. QN. Location and Elevation P-5 1-SH-10-8 H-5 RCS, RBISS, 2' P-6 1-SH-10-8 H-6 RCS, RBISS, 2' P-7 1-SH-10-B H-7 RCS, Rt11SS, 2' ;

53 1-SH-26524-4A CFS, RBISS, -15' 55 1-SH-22000-9 PLS, RBOSS, -14' 111 7-SH-30800-12 HSS, RBY, 38' l 116 4-SH-60014-4A RCO TB, -30' !

117 1-SH-22000-12 Pi,$, RBOSS, -19' 122 1-SH-21006-6A S!H RBISS, 3' 125 1-SH-21021-3 RCS, RBISS, 15' 126 1-SH-23823-7A S!H RBISS, - S' t 130 1-SH-21028-2 RCS, RBISS, 16' l 137 1-SH-21501-28 SIH, RBISS, 7' '

SS-1 SS-1 RCS, RBISS, 31' SS-2 SS-2 RCS, RBISS, 31' ;

SS-3 SS-3 RCS, RBISS, 31' 55-4 SS-4 RCS, RBISS, 31'

)

SS-5 SS-5 RCS, RBISS, 31' -

SS-6 SS-6 RCS, RBISS, 31' SS-7 SS-7 RCS, RBISS, 31' SS-8 SS-8 RCS, RBISS, 31' SS-9 SS-9 RCS, RBISS, 31' 55-10 SS-10 RCS, RBISS, 31' SS-11 SS-11 RCS, RBISS, 31' SS-12 SS-12 RCS, RBISS, 31' SS-13 SS-)3 RCS, RBISS, 6' 55-14 SS-14 RCS, RBISS, 6' SS-15 SS-15 RCS, RBISS, 6' SS-16 SS-16 RCS, RBISS, 6' SS-17 55-17 RCS, RBISS, 6' 55-18 SS-18 RCS, RBISS, 6' SS-19 55-19 RCS, RBISS, 6' SS-20 55-20 RCS, RBISS, 6' SS-21 SS-21 RCS, RBISS, 6' 55-22 55-22 RCS, RBISS, 6' 55-23 SS-23 RCS, RBISS, 6' l 55-24 SS-24 RCS, RBISS, 6' l

l Proposed Amendment No. 77, 171 4-47d i

i t

RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS Surveillance Standards 4.17 STEAM GENERATORS Apolicability Applies to inservice inspection of the steam generator tubes. >

Okintire To verify the operability and structural integrity of the tubing as part of the reactor coolant boundary.

Spardtteation l Each steam generator shall be demonstrated OPERABLE by performance of the i following augmented inservice inspection program and the requirements of Specification 1.3. ;

4.17.1 Steam _ Generator fample Selection and Inspection Steam generator tubing shall be demonstrated OPERABLE by selecting l and inspecting steam generators as specified in Table 4.17-1.

l l 4.17.2 Steam Generator Tube Sample Selection and InspRtion The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as i i specified in Table 4.17-2A. The inspection result classification and the corresponding action required for inspection of "specific l limited areas" (see paragraph 4.17 Re) shall be as specified in l Table 4.17-28. The inservice inspection of steam generator tubes '

shall be performed at the frequencies specified in Specification 4.17.3 and the inspected tubes shall be verified acceptable per the l acceptance criteria of Specification 4.17.4. The tubes selected for these inspections shall include at least 3% of the total number of tubes in both steam generators and be selected on a random basis except:

a. If experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas.

l Proposed Amendment No. 76, 171 4-51

,. RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS i

4.17.2 (continued) Surveillance Standards

b. The first sampic inspection during inservice inspection l (subsequent to the first inservice inspection) of each steam ;

generator shall include: 1

1. All nonplugged tubes that previously had detectable wa?1 l penetrations (>20%), and

2. Tubes in those areas where experience has indicated potential problems,

c. The second and third sample inspections during each inservice insptction may be less than a full tube inspection by concentrating (selecting at least 50% of the tubes to be inspected) the inspection on those areas of the tube sheet array and on those portions of the tubes where tubes with imperfections were previously found,

d. A tube inspection (pursuant to Specification 4.17.4a.5) shall be l parformed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, tnis shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection. ("Adjacent"is interpreted to mean the nearest tube capable of being inspected.)

Tubes which do not permit passage of the 9ddy current probe will be considered as degraded tubes when classifying inspection results.

e. Tubes in specific limited areas whien are distinguished by unique

} operating conditions and/or physical construction (for example, tubes adjacent to the open inspection lane or tubes whose 15th l tube support plate hole is not broached but drilled) may be excluded from random samples if all such tubes in the specific area of a steam generator are inspected. No credit will be taken for these tubes in meeting minimum sample size requirements.

The re.ults of each sample inspection shall be classified into one of the following three categories:

Caltson Inspection _ Resulti C-1 Less than 5% of the total tubes inspected are de raded tubes and none of the inspected tubes are defect ve.

C-2 One or more tubes, but nct more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes insp)cted are degraded tubes, i

C-3 Hore than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.

Note: In all inspections, previously degraded tubes must exhibit significant (>10%) further wall penetrations to be included in the above percentage calculations.

Proposed i Amendment No. 76, 171 1

4-52

i

' l RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.17-3 OTSG AUXILIARY FEEDRATER HEADER SURVEILLANCE OlSG A Special_laterest.Tubel

Row Tube No 5 1, 46

) 6 2, 3, 49, 50, 51 7 1, 2, 53, 54 I 8 1, 2, 56, 57 44 1, 119 45 1, 120 46 1, 119

47 1, 122

} 48 1, 123 49 1, 124 103 1, 124 j 105 1 106 1, 119 i

107 1, 120

! 103 1, 119 t 144 1, 2, 56, 57 145 1, 2, 53, 54 j 146 2, 3, 49, 50, 51 i 147 1, 46 i

1 l

I l

i

? ,

1 1

Proposed Amendment No. 76, 171 1 4-57b

i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.17-3 (Continued)

OTSG AUXILIARY FEE 0 HATER HEADER SURVEILLANCE DIS.(ULSPEtLLinterest Tubet Row Tube No 5 1, 46 6 2, 3, 49, 50 7 1, 2, 53, 54 8 1, 2, 56, 57 44 1, 119 45 1, 120 46 1, 119 47 122 48 1, 123 49 1, 124 103 1,124 104 123 105 122 106 1, 119 107 1, 120 108 1, 119 144 1, 2, 56, 57 145 1, 2, 53, 54 146 2, 3, 49, 50, 51 147 1, 46 Proposed Amendmer.t No. 76, 171 4-57c

i 3 ..

t RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS Survelliance Standards 4.18 FIRE SUPPRESdiun SYSTEM SURVEILLANCE 4.1C,1 Instrumentation 4.18.1.1 Except for fire detection instruments inaccessible during power operation, ea:h of the fire detection instruments in Table 3.14-1 shall be demonstrated OPERABLE at least semiannually by a CHANNEL FUNCTIONAL TEST. Fire detectors which are not accessible during plant operation shall be demonstrated OPERABLE by the performance of a CHANNEL FUNCTIONAL TEST during each COLD SHUTDOWN exceeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless performed in the previous 6 months.

4.18.1.2 The NFPA Standard 72D supervised circuits associated with the l detector alarms for each of the fire instruments in Table 3.14-1 shall be demonstrated OPERABLE at least semiannually.

4.18.1.3 Fon-supervised c'~cuits associated with detector alarms, between the instrument and the ccatrol room shall be demonstrated OPERABLE at least once per 31 days.

4.18.2 Hater System 4.18.2.1 The fire suppression water system shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying the contained water supply volume,

b. At least once per 31 days on a STAGGERED TEST BASIS by starting each electric motor driven pump and operating it for at least 15 minutes on recirculation flow,

c. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path is in its correct position as indicated by position instrumentation,

d. At least once per 6 months by performance of a system flush to each test fixture.

e. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel,

f. At least once per 18 months by performing a system functional test which includes simulated automatic actuation of the system throughout its operating sequence, and:

1. Verifying that each pump develops at least 2000 gpm at a l minimum pressure of 125 psig.

Proposad Amendernt No. 35, 171 4-58

~~ ,

k RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS urveillance Standards 4.18.2.1 (Continued)

2. Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel, and

3. Verifying that each fire suppression pump starts (sequentially) to maintain the fire suppression water system pressure greater than or equal to 80 psig.

g. At least once per 3 years by performing a flow test of the system in accordance with Chapter 5. Section 11 of the Fire Protection Handbook, 14th Edition, published by the National Fire Protection Association.

l l

\

l l

i Proposed Amendment No. 35, 171 4-59 l

i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.18.2.1 (Continued)

h. The fire pump diesel engine shall be demonstrated OPERABLE:

1. At least once per 31 days by verifying:

a. The fuel storage tank contains at least 250 gallons of fuel, and

b. The diesel starts from ambient conditions and operates for at least 30 minutes on recirculation flow.

2. At least once per 92 days by verifying that a sample of diesel from the fuel storage tank, obtained in accordance with ASTH-D270-65, is within the acceptable limits specified in Table 1 of ASTH-0975-78 with respect to l viscosity, water content and sediment.

3. At least once per 18 months, by subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for the class of service.

1. The fire pump diesel starting 24-volt battery bank and charger shall be demonstrated OPERABLE:

1. At least once per 7 days by verifying that:

a) The electrolyte level of each battery is above the plates, and b) The overall battery voltage is greater than or equal to 24 volts.

2. At least once per 92 days by verifying that the specific gravity is appropriate for continued service of the battery.

3. At least once per 18 months by verifying that:

a) The batteries, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration, and b) The battery-to-battery and terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material. ,

Proposed Amendment No. 35, 171 4-60

i O

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Table 4.20-1 (Continued)

Instrument Instrument Channel Source Channel Channel Instrument Check Chark Calibration 1111

3. Auxiliary Building '

Grade Level Vent

a. Noble Gas D H R(3) Q(2)

Activity Monitor

b. Iodine Sampler H NA NA NA

c. Particulate H NA NA NA Sampler

d. System Effluent D NA R Q(6)

Flow Rate Device

e. Sampler Monitor D NA R Q '

Flow Rate Measurement Device

' l I

i a

j Proposed Amendment No. 53 73, 87, 98, 171 4-67 .

i f

s RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.22.4 Gaseous Radwaste Treatment Surveillance Reauirement Doses due to gaseous releases to areas at and beyond the Site Boundary For Gaseous Effluents (see Figure 5.1-3) shall be projected at least once per 31 days in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCH) when Gaseous Radwaste Treatment Systems are not being fully utilized.

The installed VENTILATION EXHAUST TREATHENT SYSTEli and Haste Gas System shall be considered OPERABLE by meeting Specifications 3.18.1, 3.18.2 and 3.18.3.

Bale.1 The operability of the Haste Gas System and the VENTILATION EXHAUST TREATHENT SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used, when l specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as the dose design objectives set forth in Sections II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.

i i

Propar,ed Amendment No. 53, 98, 171 4-79

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS ,

Design Features j

5. DESIGN FEATURES 5.1 SJJE The Rancho Seco reactor is located on the 2,480 acres owned by Sacramento Municipal Utility District, 26 miles north-northeast of Stockton and 25 miles southeast of the City of Sacramento, California. USAR Figure 1.1-2 shows the plan of the site. The minimum in 10 CFRdistance to thebe 100.3, shall boundary of th9)9xplusion 2,100 feet.(i (21 area, as defined 5.1.1 Exclusion Area The EXCLUSION AREA is shown in Figure 5.1-1. !

5.1.2 Low Pooulation Zone The LOH POPULATION ZONE is shown in Figure 5.1-2.

5.1.3 Site Boundarv For Gaseous and Liauid Effluents The SITE BOUNDARY FOR GASEOUS AND LIQUID EFFLUENTS for meeting 10 CFR 50, Appendix I gulielines is shown in Figure 5.1-3.

5.1.4 Site Boundary For Liquid Effluents The SITE BOUNDARY FOR LIQUID EFFLUENTS for 10 CFR 20 compliance is shown in Figure 5.1-4.

REFERENCES (1) USAR paragraph 1.2.1 (2) USAR paragraph 2.2.1 i

e l

l i

l Proposed Amendment No. 87, 98, 171 5-1

-..e - , - - , ,,- - . _ , ,v.-n.-- - -

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Chief Executive Officer (CEbi, Nuclear shall be responsible for the management of the overall facility, and the Assistant General Manager (AGM), Nuclear Power Production shall be responsible to him for the operation and maintenance of the plant. They shall delegate in writing the succession to their responsibility during their absences.

6.2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for the facility management and technical support shall be as shown on Figure 6.2-1.

EACILITY STAFF 6.2.2 The facility organization shall be as shown on Figure 6.2-2 and: ,

a. Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.

b. At least one licensed Operator shall be in the Control Room when fuel is in the reactor.

c. At least two licensed Operators, one of whom shall be senior licensed, shall be present in the Control Room for all Reactor operational modes from HEATUP-C0CLDOWN through POWER OPERATION.

d. An individual qualified in radiation protection practices and procedure *, shall be on site when fuel is in the reactor.

e. ALL COR'. ALTERATIONS after the initial fuel loading shall be direct'y supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator limited to Fuel Handling who has no other concurrent responsibilities during this operation.

f. A site Fire Brigade of at least 5 members shall be maintained onsite at all times.* The Fire Brigade shall not include 5 l members of the minimum shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency. I

Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected 4 absence provided immediate action is taken to fill the required positions.

Proposed Amendment No. 18, 24, 25, 87, 96, 171 i 6-1

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.2 ORGANIZATION FACILITY STAFF

g. Administrative procedures shall be developed and implemented to limit the working hours of facility staff who perform safety-related functions; e.g., senior reacto* operators, reactor operators, key radiation protection piirsonnel, auxiliary l operators, and key maintenance personnel.

Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall he to have operating personnel work a normal shift schedule whi1e the plant is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance or major plant modifications, on a temporary basis, the following guidelines shall be followed:

a. An individual should not be permitted to work more than 16 hcurs straight, excluding shift terrcVer time,

b. An individual should not ce permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24-hour period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48-hour period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any seven day period, all excluding shift turnover time.

c. A break of at least eight hours should be allowed between work periods, including shift turnover time.

d. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.

Any deviation from the above guidelines shall be authorized by the AGM, Nuclear Power Production or designee, in accordance with established procedures and with documentation of the basis for granting the deviation. Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the AGM, Nuclear Power Production or designee to assure that excessive hours have not been assigned. Routine deviation from the above guidelines is not authorized.

Proposed Amendment No. 49, 96, 171 6-la

s RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS l Administrative Controls 6.3 FACILITY STAFF OUALIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the Manager, Radiation Protection who shall meet or exceed the qualifications of Regulatory Guide 1.8, Revision 2, and (2) the l Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline. The STA shall receive specific training in plant design, and response and analysis of the plant for transients and accidents.

6.4 TRAINING 6.4.1 A retraining and replacement training program for the operating staff shall be maintained under the direction of the Manager, Training and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 or ANS 3.1-1981 as endorsed by Regulatory Guide 1.8, Revision 2, and 10 CFR Part 55.

6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Manager, Training and shall meet or exceed the requirements of Section 27 of the NFRA Code - 1975, except refresher classroom training shall be on a quarterly schedule.

6.5 REVIEH AND AUDIT 6.5.1 PLANT REVIEW COMMITTEE (PRC)

FUNCTIOR 6.5.1.1 The Plant Review Committee shall function to advise the AGM, Nuclear Power Production and Hanagement Safety Review Committee on all matters related to nuclear safety.

l COMPOSITION 6.5.1.2 The Plant Review Committee shall be composed of a chairman, and a minimum of six members.

l

. Proposed 1

Amendment No. 14, 18, 24, 27, 62, 96, 171 6-3

,, RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS RESPONSIBILITIES (Continued)

g. All events requiring a Licensee Event Report as defined by 10 CFR 50.73 and NUREG 1022 to determine adequacy of corrective action and to detect any degrading trend.

h. Special investigations and reports thereon as requested by the AGH, 1

Nuclear Power Production.

1. The Plant Security Plan and changes thereto.

j. The Emergency Plan and changes thereto.

k. Changes to the PROCESS CONTROL PROGRAM, the OFFSITE DOSE CALCULATION MANUAL and the RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM MANUAL. (See Specifications 6.15 and 6.16.)

1. Major changes to the Radioactive Haste Treatment Systems (Liquid, Gaseous and Solid), and all information required by Specification 6.17.

m. Review of any accidental, unplanned, or uncontrolled release of radioact've material to the environs including the preparation and forwarding of reports covering evaluation, recommendations, and disposition of the corrective action to prevent recurrence, and the forwarding of these reports to the AGH, Nuclear Power Production and l to the HSRC.

AUTHORITY 6.5.1.7 The Plant Review Committee shall:

a. Recommend in writing to the AGH, Nuclear Power Production approval or disapproval of items considered under 6.5.1.6(a) through (m) above.

b. Render determinations in writing with regard to whether or not each item considered under 6.5.1.6(a) through (e), and (1) above constitutes an unreviewed safety question,

c. Provide immediate written notification t? the Chairman of the Hanagement Safety Review Committee of disagreement between the PRC and the AGH, Nuclear Power Production; however, the AGH, Nuclear Power Production shall have responsibility for resolution of such disagreements pursuant to 6.5.1.1 above.

RECORDS 6.5.1.8 Records of the PRC activities shall be prepared, approved and distributed as indicated below:

a. Minutes of each PRC meeting, including appropriate documentation f of reviews encompassed by Specification 6.5.1.6e and g, shall be prepared, approved, and forwarded to the AGH, Nuclear Power Production and to the Chairman, Management Safety Review Committee within 14 days following each meeting.

Proposed hendment No. 24, 62, 63, 96, 98, 171 6-5

s RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls

a. (Continued)

The AGH, Nuclear Power Production will approve plant administrative procedures, Security Plan Implementing Procedures and Emergency Plan Implementing Procedures.

Approval of temporary procedure changes which clearly do not change the intent of the approved procedures can be made by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License. The change shall be documented, reviewed and approved by the procedures approval authority within 14 days of implementation,

b. Proposed changes or modifications to plant systems or equipment that affect nuclear safety shall be reviewed by an individual (s) other than the individual (s) who designed the modification, but who may be from the same organization as the individual (s) who designed the modifications. Such modifications shall be approved by the AGH, Nuclear Power Production or designee.

c. Proposed tests and experiments which affect plant nuclear safety and are not addressed in the Updated Safety Analysis Report shall be reviewed by an individual (s) other than the individual (s) who prepared the proposed test or experiment.

Such tests or experiments shall be approved by the AGH, Nuclear Power Production or designee.

d. Events reportable pursuant to the Technical Specification 6.9 and violations of Technical Specifications shall be .

investigated and a report prepared which evaluates the event and which provides recommendations to prevent recurrence. Such reports shall be reviewed by the PRC and forwarded to the AGH, M lear Power Production, and to the Chairman, MSRC.

e. Individuals responsible for reviews performed in accordance with 6.5.3a, 6.5.3b and 6.5.3c shall meet or exceed the 4 qualification requirements of Section 4.4 of ANSI 18.1. 1971.

Each such review shall include a determination of whether or not additional, cross-disciplinary review is necessary. If deemed necessary, such review shall be performed by the review personnel of the appropriate discipline. A list of qualified reviewers for the independent reviews described in 6.5.3a., b.,

l c., above shall be established by the AGH, Nuclear Power Production. l l

l i

1 Proposed !

Amendment No. 96, 171 6-9

RANCHO SECO UNIT 1

- TECHNICAL SPECIFICATIONS !

Administrative Controls AUDITS 6.5.4 Audits of facility activities shall be performed under the cognizance of the Director, Quality Assurance, Nuclear. These audits shall encompass:

a. The conformance of facility operation to all provisions contained within the Technical Specifications and applicable license conditions at least once per year.

b. The performance, training and qualifications of the District's entire facility technical staff at least once per year.

c. The result of all actions taken to correct deficiencies occurring in facility equipment, structures, systems or methods of operation that affect nuclear safety at least once per 6 months for those changes not previously audited,

d. The performance of all activities required by the Quality Assurance Program to meet the criteria of Appendix 8,10 CFR 50, at least once per 2 years.

e. The facility Emergency Plan and implementing procedures at least once per 2 years,

f. The Facility Security Plan and implementing procedures at least once per 2 years,

g. Any other area of facility operation considered appropriate by the Chief Executive Officer, Nuclear.

h. Compliance with fire protection requirements and implementing procedures at least once per 2 years.

1. An independent fire protection and loss prevention inspection and audit shall be performed annually utilizing either qualified offsite licensee personnel or an outside fire protection firm.

j. An inspection and audit of the fire protection and loss prevention program shall be performed by an outside qualified fire consultant at intervais no greater than 3 yaars.

k. The radiological environmental monitoring program and the results ,

thereof at least once per 12 months.

1. The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at r least once oer 24 months. ;

m. The PROCE! "00L PROGRAM and implementing procedures for processing . \ aging of radioactive wastes from liquid systems at least once p, 44 months.

n. The performance of activities required by the Quality Assurance Program for Effluent Control and Environmental Monitoring at least once per 12 months.

Proposed Amendment No. 18, 53, 96, 98, 171 6-10

w RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The provisions of 10 CFR 50.36(c)(1)(i) and 10 CFR 50.72 shall be complied with,

b. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within I hour. The Director, Nuclear Operations and Maintenance, the AGM, Nuclear Power Production, and the Chairman of the HSRC shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. A Safety Limit Violation Report shall b'e prepared. The report shall be reviewed by the PRC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence,

d. The Safety Limit Violation Report shall be submitted to the Commission, the MSRC, and the AGM, Nuclear Power Production, within 14 days of the violation, r

i 6.8 PROCEDURES 6.8.1 Hritten procedures shall be established, implemented and maintained covering the activities referenced below:

a. The applicable procedures recommended in Appendix A of Safety Guide 33, November 1972,

b. Refueling operations.

c. Surveillance and test activities of safety related equipment.

d. Security Plan implementation. l

e. Emergency Plan implementation.

f. Fire Protection Procedures implementation,

g. PROCESS CONTROL PROGRAM implementation,

h. OFFSITE DOSE CALCULATION MANUAL implementation,

i. RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM MANUAL implementation.

j. Quality Assurance Program for Effluent Control and Environmental Monitoring using the guidance of Regulatory Guide 4.15 Revision '

l. February 1979. l 6.8.2 Each procedure of 6.8.1 above and changes thereto shall be '

reviewed and approved as set forth in Specification 6.5. j Proposed Amendment No. 18, 24, 53, 87, 96, 98, 171 6-11

s RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.9.2 Radiological Reoorts 6.9.2.1 Annual Radioloaical Reoorts Annual reports covering the activities of the unit, as described below, for the previous calendar year shall be submitted as follows:

6.9.2.1.1 Annual Occuoational Radiation Exoosure Reoort The Annual Occupational Radiation Exposure Report shall be submitted to the Commission within the first calendar quarter of each calendar year in compliance with 10 CFR 70.407.

6.9.2.1.2 Annual Exoosure Reoort The Annual Exposure Report shall be submitted to the Commission within the first calendar quarter of each calendar year in accordance with the guidance contained in Regulatory Guide 1.16.

6.9.2.2 Annual Radiological Environmental Ooeratina Reoort

6.9.2.2.1 Routine Annual Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year.

6.9.2.2.2 The Annual Radiological Environmental Operating Report shall include summaries and statistical evaluation of the results of the radiological environmental surveillance activities for the report period, including a comparison with operational controls (as appropriate). The reports shall also include the results of the Land Use Census required by Specification 3.23 and 4.27. In l l the event a radionuclide concentration should be confirmed in ,

excess of the reporting level in Table 3.22-2 by environmental l measurements, the report shall describe a planned course of !

corrective action.

Proposed Amendment No. 17, 53, 98, 171 6-12a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.9.2.2.2 (Continued)

The Annual Radiological Environmental Operating Report shall include summarized and tabulated results of all radiological environmental samples taken during the report period. In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.

The report shall also include the following: a summary description of the Radiological Environmental Monitoring Program; including a map of all sampling locations keyed to a table giving distances and directions from one reactor, and the results of licenseo participation in the Interlaboratory Comparison Program (Specification 4.30). The annual report shall also include information related to Specification 4.29, Fuel Cycle Dose.

6.9.2.3 Semiannual Radioactive Effluent Release Reoort Routine Semiannual Radioactive Effluent Release Reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year.

6.9.2.3.1 The Semiannual Radioactive Effluent Release Report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit with data summarized on a quarterly basis.

The Semlannual Radioactive Effluent Release Report shall include a summary of meteorological data collected over the report period. In lieu of submitting all meteorological data with the af ter July 1 report, the information will be retained in a file onsite and shall be submitted to the NRC upon request.

t I

f i

Proposed ,

Amendment No. 53, 87, 96, 98, 171 6-12b

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.9.2.3.1 (Continued)

The Semiannual Radioactive Effluent Release Report shall include an assessment of the radiation doses from radioactive gaseous and liquid effluents to individuals due to their activities outside the site boundary (Figures 5.1-3 and 5.1-4) during the report period.

The Semiannual Radioactive Effluent Release Report shall include the following information for all unplanned releases to unrestricted areas of radioactive materials in gaseous and liquid effluents:

a. A description of the event and equipment involved,

b. Cause(s) for the unplanned release,

c. Actions taken to prevent recurrence.

d. Consequences of the unplanned release.

The Semiannual Radioactive Effluent Release Report shall include l an assessment of radiation doses from the radioactive liquid and gaseous effluents released from the unit during each calendar quarter. The assessment of radiation doses shall be performed in accordance with the OFFSITE DOSE CALCULATION MANUAL (0DCH).

The Semiannual Radioactive Effluent Release Report shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP), RADIOLOGICAL ENVIRONMENTAL HONITORING PROGRAM MANUA'. and 0FFSITE DOSE CALCULATION MANUAL (ODCH) pursuant to Specifications 6.15 and 6.16 as well as any major changes to Liquid, Gaseous or Solid Radwaste Treatment Systems l pursuant to Specification 6.17.

The Semiannual Radioactive Effluent Release Report shall include tables for comparison with Specifications 3.17.2, 3.18.2, and 3.18.3. The July-December report shall include a summary table for comparison with the annual values in Specifications 3.17.2, 3.18.2, and 3.18.3.

The Semiannual Radioactive Effluent Release Report shall also ,

include events described in Specifications 3.17.1, 3.17.3, l 3.18.1 and 3.18.5. l l

l l

l 1

l Proposed Amendment No. 17, 49, 53, 63, 96, 98, 171 6-12c l

I

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls Soecial Reoorts 6.9.5 Special reports shall be submitted to the Regional Administrator, NRC Region V office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:

A. A one-time only, "Narrative Summary of Operating Experience" will be submitted to cover the transition period (calendar year 1977).

B. A Reactor Building Structural Integrity Report shall be submitted within 90 days of completion of the following tests per the requirements of Technical Specifications 4.4.2.1 and 4.4.2.5.

1. Lif t-Off Measurements (4.4.2.2)

2. Strand Surveillance (4.4.2.3)

3. Anchorage Surveillance (4.4.2.4)

C. Inservice Inspection Program (4.2.2)

D. Inoperable Accident Monitoring Instrumentation 30 days (3.5.5) i E. Status of Inoperable Fire Protection Equipment 30 days (3.14.1.2, l l

3.14.2.2, 3.14.3.2, 3.14.4.2, 3.14.5.2, 3.14.6.2) 14 days (3.14.2.3) l F. Inoperable Emergency Control Room /TSC Ventilation Room Filter System 30 days (3.13.3) l G. Radioactive Liquid Effluent Dose 30 days (3.17.2)

H. Noble Gas Limits 30 days (3.18.2)

1. Radioiodine and Particulates 30 days (3.18.3) l l

1

, 1 Proposed l Amendment No. 17, 18, 53, 70, 75, 85, 97, 98, 171 1 4

6-129 l 1

4 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS.

Administrative Controls Soecial Reoorts (Continued)

J. Gaseous and Liquid Radwaste Treatment 30 days (3.18.4 and 3.17.4)

K. Radiological Environmental Monitoring Program 30 days (3.22)

L. Meteorological Instrumentation 10 days (3.29)

H. Integrated Leak Rate Test 90 days (4.4.1.1.5)

N. Fuel Cycle Dose 30 days (3.25)

O. Land Use Census (and REMP Monitoring 30 days (3.23)

Point Locations)

P. Steam Generator Tube Inspection 30 days (4.17.5)

Proposed Amendment No. 85, 98, 171 6-12h

4 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.10 RECORD RETENTION l 6.10.1 The following records shall be retained for at least 5 years:

Records and logs of facility operation covering time interval at

a. ,

each power level.

l

b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of l equipment related to nuclear safety.

c. Licensee Event Reports. ;

d. Records of surveillance 3ctivities, inspections and calibrations required by these Technical Specifications. !

e. Records of reactor tests and experiments. !

I

f. Records of changes made to Operating Procedures. !

g. Records of radioactive shipments. !

h. Records of sealed source leak tests and results.

1. Records of annual physical inventory of all sealed source i material of record. j 6.10.2 The following records shall be retained for the duration of the !

Facility Operating License:

a. Record and drawing changes reflecting facility design modifications made tc systems and equipment described in the Updated Safety Analysis Report.

Proposed Amendment No. 63, 171 6-13

- Facility Change Safety Analysis Log No. 1076 Proposed Amendment No. 171 Page 1 of 13 ATTACHMENT III DESCRIPTION OF CHANGES, REASON FOR CHANGES, AND N0 SIGNIFICANT HAZARDS CONSIDERATION Proposed Amendment No.171, Revision 1 consists of various editorial and technical changes to the Rancho Seco Technical Specifications. A description of and reason for each change is presented below:

1. Change _Deitdplion: Index, pr.ge 1: Delete "Mode" from title of section 1.2.12; add section 1.2.14 "Leakage"; correct page numbers for various section titles.

Renon: Correct index to accurately reflect section numbers, titles and page numbers.

2. Change _

Description:

Index, page 11: Add section 1.5.7; change title of section 1.8 to "LICENSEE EVENT REPORTS"; delete title of section 1.9.9.

Reason: Correct index to accurately reflect section numbers and titles.

3. Change _Qescdp.tlon: Index, page 111: Add "and Leak Detection" to title of sectior 3.1.6; delete "$ HOCK SUPPRESSORS" from title of section 3.12.

Reason: Account for the expansion of section 3.1.6 and to correct title to be consistent with actual title of section 3.12.

4. Change Descriplion: Index, page iv: Change page number for section 3.14.5; change title of section 3.14.6 to "Fire Rated Assemblies."

Reason: Correct title and page number to accurately reflect actual text.

5. ChaDge_Deicdplion: Index, page v: Add "VENTILATION" to title of section 3.27.

Reason: Correct index title to match title on actual specification.

6. Changu_D_ciedp.t ton: Index, page vi: Change page number for section 4.9; change title of section 4.10 to "CONTROL ROOM / TECHNICAL SUPPORT CENTER EMERGENCY FILTERING SYSTEH" and correct page number; delete "SHOCK SUPPRESSORS" from title of section 4.14.

Reason: Correct titles and page numbers to accurately reflect actual text.

7. Change _Descdption: Index, page vii and vita: Add titles and page numbers for sections 4.22.4 and 4.22.5. Some index sections shifted to page vila.

Reason: Inadvertently omitted when issued as Amendment No. 98.

Page vila changed due to shifting of the index.

8. Change _Qescdpilon: Index, page viit: Add appropriate acronyms to sections 6.5.1 and 6.5.2; change title of section 6.6 to "LICENSEE EVENT REPORT ACTION" and correct page number; delete title of section 6.12.

Reason: Correct index to accurately reflect actual text.

9. Change _.Destdption: Index, page tx: Change page number for Table 3.7-1.

Reason: Correct page number.

~

1

- Facility Change Safety Analysis Log No. 1076 !

Proposed Amendment No. 171 Page 2 of 13

10. Change Descriotion: Index, page x: Change title of Table 6.2-1 to correspond to actual title of Table 6.2-1.

ReAion: Correct table title to match title on actual table. ,

11. Chanae_Descriotion: Index, page xi: Delete listing for Figure '

3.1.2-4; add page numbers to list of figures.

Reason: Provide clarification and additional information for increased accessibility. l

12. Change Descriotion: Index, page xil: Delete listing for Figures 3.5.2-10. -11. -12 and Figure 3.18-1; correct typographical error on title of Figure 5.1-1; add qualifying parenthetical to title of Figure

. 5.1-2; delete "to Rancho Seco" from title of Figure 6.2-1; add page numbers to all figure listings.

Heaton: Correct index to reflect actual text; provide additional :

information for increased accessibility.

13. Change _QelCdRilDn: Specification 1.2.10, page 1-2: Change definition to "See Specification 3.0.3."

Renon: Specification 3.0.3 was added by Amendment No. 97. With this i addition the definition for "Remain Critical" is redundant and in some cases confusing. ,

14. Change DescdRilon: Specification 1.4.2, page 1-2a: Change "FSAR" to '

"USAR."

Reaton: Pursuant to 10 CFR 50.71, the Updated Safety Analysis Report ;

(USAR) is the official updated version, issued annually, of the Final '

Safety Analysis Report (FSAR). Accordingly, the USAR is referenced in the Technical Specifications rather than the FSAR. [

I

15. Change _.Deatdption: Table 1.2-1, page 1-2c: Add header and section name to top of page.

Renon: Inadvertently omitted when issued as Amendment No. 97.

16. Chauge_Restdation: Specifications 1.4.3, 1.4.4 and 1.4.5, page 1-3:

Change "FSAR" to "USAR."

ReAion: See the reason given for Change Description No. 14. ;

i

17. Change _Descdpilon: Specification 1.9, page I-5: In first line, correct spelling of word "accomodate" to "accommodate"; delete Specification 1.9.9.

Reason: Typographical correction. There are no longer any 2 year time periods. Amendment No. 94 revised Specification 4.6.3 which had the only ,

i biannual surveillance. !

)a i

18. Change _Descdotion: Specification 1.27, page 1-9: Delete "-BAR" from the first line. _

l Reason: An editorial correction is made to the E definition.

i

19. Change _Descdoting: Table 1.9-1, page 1-10: Add header and section !

! name to top of page; delete definition for "BA."

Reason: Inadvertently omitted when issued as Amendment No. 98; see the reason given for Change Description No. 17. l v .. - _ _ _ _ _ _ _ _ _ _ _ __ _ . .

, Facility Change Safety Analysis Log No. 1076 Proposed Amendment No. 171 Page 3 of 13

20. Chanae

Description:

Bases for Specification 2.2, page 2-4: Change the designation of the setting for the reactor high pressure trip from "2300" psig to "2355 psig."

Rea. son: Amendment No. 83 changed 2300 to 2355. Amendment No. 87 changed the references for Specification 2.2 and inadvertently changed the bases back to the pre-Amendment 83 version.

21. Change _

Description:

Specification 2.3.1 Bases A, page 2-6: In first paragraph change "FSAR" to "USAR."

Renon: See the reason given for Change Description No. 14.

22. Change Descrip.tlon: References, page 2-8: Change "FSAR" to "USAR."

Reason: See the reason given for Change Description No. 14.

23. Change

Description:

Page 3-0: Correct spelling of the word "UNIT" in header.

Renon: Typographical correction.

24. Change Descriotion: Table 3.0-1 and 4.0-1, pages 3-0a, b and 4-0a, b:

Include recently added sections to the table; add header and section name.

Renon: Amendment No. 97 added these tables. Tech Spec sections added with Amendment No. 98 are addressed in Tables 3.0-1 and 4.0-1. Header and section name were inadvertently omitted when issued as Amendment No. 97.

25. Change Rescrlpilan: Bases for Specification 3.0, pages 3-Oc through 3-Of: Add Bases for Specifications 3.0.1 through 3.0.4.

Realon: Specifications 3.0.1 through 3.0.4 were added by Amendment No. 97 without any Bases. Bases are from Generic Letter 87-09.

1

26. Change _ Descrinilon: Specification 3.1.2.2, page 3-3: Change "EFP years" to "EFPY" and delete "power."

Reason: To make this specification consistent with Specification 3.1.2.1.

27. Change.Jescrjpt100: Bases for Specification 3.1.2, page 3-4: Change !

"7." to "A" in second paragraph; change "RTON7" to "RTNDT" in third I paragraph. In the References section correc" the ASME Code title by adding the word: "Vessel" in Reference (2); delete Reference (5), and renumber References (6) and (7) as References (5) and (6).

Reason: Typographical corrections. The correction to the ASME title provides its proper designation. The referenced FSAR paragraphs were changed to USAR in Proposed Amendment No. 139 (1/29/86). Further review has determined that paragraph 4.4.4 as shown for Reference (5) is nonexistent in the USAR and is therefore deleted. References (6) and (7) are renumbered accordingly.

28. Change _DescrJptbn: Specification 3.1.3 Bases, pages 3-6 and 3-7:

Delete reference designator "(2)" and delete Reference (2).

Reason: The reference was changed to the USAR by Amendment No. 87; however, further review has determined that paragraph 3.2.1.4 as shown I for Reference (2) is nonexistent in the USAR and is therefore deleted.

. Facility Change S6fety Analysis Log No. 1076 Proposed Amendment No. 171 Page 4 of 13

29. Change Descriotion: Specification 3.1.5, page 3-10: In the Applicability statement change "continous" to "continuous."

Ruson: Typographical correction.

30. Change Descriotion: Specification 3.1.8 Bases, page 3-15b: Add "of the USAR" after the number 7.2.2.1.3.

Realon: The proposed change provides more explicit identification.

31. Chanae Descriotion: Specifications 3.2.2.2, 3.2.2.2.2, 3.2.2.2.3, 3.2.2.5, 3.3.1.B.3, 3.3.1.B.5, and Bases for 3.2, pages 3-17a and 3-18a:

Delete "open and tagged" and "racked out" and add "remove power" or "power removed from."

Renoo: Hording is not consistent. "Removing power" is consistent with wording in Standard Tech Specs.

32. Change Descriotion: Specification 3.3 Bases, page 3-22: Delete subparagraph (a).

Reason: Subparagraph (a) states "Motor operated valves shall be placed in the closed position and power supplies de-energized." In reviewing the previous amendments affecting this Bases, there is no accounting for the appearance of this subparagraph. Since the wording is irrelevant to the borated water storage tank being discussed, the District proposes to delete subparagraph (a).

33. Changelestdp. tion: Specifications 3.4.1 and 3.4.2, pages 3-23 and 3-23a: Add "through HOT SHUTDOHN" to 3.4.1. Delete "critical" and add "in the STARTUP through POWER OPERATION mode" to 3.4.2.

Bea1on: The applicability of 3.4.1 and 3.4.2 was not clearly stated.

The intent clearly is that these specifications are applicable for the range of plant conditions now specified.

34. Changeleicdp11on: Bases for Specification 3.4 page 3-24a: Add the function of Turbine Stop Valves to the Bases.

Realon: There was no Bases for the function of these valves.

35. Change _Q11cdpuna: Specification 3.5.1 Bases, page 3-26: Add Note (1), "Pressure setpoint tolerances are 600 1 25 psig and 100 1 50 psid"; correct the differential pressure unit from 100 psig to 100 psid.

Reason: Note (1) was added for clarification and is in accord with the actual EFIC design as described in Proposed Amendment Nr. 152, approved as Amendment No. 93 (1/5/88). Changing the differential pressure unit to psid is an editorial correction.

36. ChangeJ)eicdotlon: Specification 3.5.2.2, page 3-31: 'Jelete letter "A." from subheading.

Reason: Unnecessary since there are no comparable succeeding lettered items.

37. Change _Desctlption: Specification 3.5.2.2, page 3-32: Change subheading letter "F" to "e."

Reason: Page 3-31 was retilsed by Amendment No. 97 and lettering on page 3-32 does not agree.

. Facility Change Safety Analysis Log No. 1076 Proposed Amendment No. 171 Page 5 of 13

38. Change

Description:

Specificat!on 3.5.2 Bases, page 3-33b: Correct specification designation from "3.5.2.5.0.1" to "3.5.2.50.1" and "3.5.2.5.D.2" to "3.5.2.5D.2." '

Reason: The correction properly designate' the specification.

39. Chanae Descriotion: Specification 3.5.2 References, page 3-33b:

Change "3.2.2.1.2" in Reference (1) to "3.2.2.1.3."

Reason: Review of the USAR has determined that Section 3.2.2.1.3 is a more appropriate refeience to the subject of hot shutdown reactivity as addressed in the Bases.

40. Change

Description:

Table 3.5.5-1 Action, page 3-38f: Correct spelling of "pursuant" in item 2).

Reason: Typographical correction.

41. Change

Description:

Specification 3.5.6 Bases, page 3-38h: Add Note (1), "Pressure setpoint tolerances are 600 1 25 psig and 100 1 50 psid"; correct the differential pressure unit from "100 psig" to "100 psid."

i Reason: Note (1) was added for clarification and is in accord with the actual EFIC design as described in Proposed Amendment No. 152, approved as Amendment No. 93 (1/5/88). Changing the differential pressure unit from "100 psig" to "100 psid" is an editorial correction.

42. Change

Description:

Specification 3.9 Bases, page 3-46b: Replace the reference to a specific operating procedure with the words "as required by operating procedures." 4 Reason: Replacement of the wording "per Operating Procedure A.21, l Section 7.3" with less specific wording allows future renumbering of the procedure or the section without necessitating a Technical Specification amendment. ,

1

43. Change _De.sritplion: Table 3.14-1, page 3-55: Increase the quantity of smoke detectors in Detection Zone 13 from 4 to 8, and in Detection Zone 16 from 4 to 6.

Reason: The additional smoke detectors were installed to comply with 10 CFR 50, Appendix R, Section III.G.2, and with the District's commitment made in LER 87-29.

44. Change _Re1IrlpijDn: Table 3.14-1, page 3-55b: Change number of heat i detectors in Zones 81 and 82 from 0 to 6.

Reason: Identify total number of fire detection instruments in these Zones.

! 45. Change _Desttjplion: Specification 3.15 Bases, page 3-60: Add "monitor" to end of third senience in second paragraph.

Reason: Add missing word.

46. Changt_Descripilon: Specification 3,16 Bases, page 3-63: Correct order of words "radioactive of" in first sentence to "of radioactive."

Season: Editorial correction, i

i

. Facility Change Safety Analysis Log No. 1076 Proposed Amendment No. 171 Page 6 of 13

47. Change Descriotion: Specification 3.18.4, page 3-78: Delete "(see figure 5.1-3)" in secord sentence.

ReAssa: This figure reference is not applicable to this sentence.

48. Change Descriotion: Specification 3.20 Bases, page 3-79: In second paragraph, change "Vol. VI FSAR" to "Vol. VIII, USAR." In the third paragraph, change the abbreviation "Ci" to "curies," and change "FSAR Table . . ." to "USAR Table . . ."

Reason: In reviewing the USAR, it was determined that the contents of page 140-25 of Volume VIII are eruivalent to page 140-25 of Volume VI of the FSAR. The changes from "FSAR" to "USAR" are as explained for Change Description No. 14. Spelling out "curies" is consistent with other uses of the unit in Specification 3.20.

49. Change _DAssdphon: Specification 3.27 Action, page 3-93: Change "non-OPERABLE" to "inoperable" in statements a. and b.

4 Reason: Editorial clarification.

i I 50. Change _ Des _CrJpMon: Specification 3.29 Action and Bases, page 3-94:

) Add word "Release" to title of Semiannual Radioactive Effluent Release Report; add a discussion of meteorological monitor channels.

Reason: Correct actual report title; clarify channel configuration.

I 51. Change _Descriphon: Table 4.0-1, page 4-0a, b: See Change Description and Reason for Item 24.

i 52. Change _DescIlpilon: Bases for 4.0, pages 4-Oc through 4-Oe: Add Bases

for Specifications 4.0.1 through 4.0.3.

Reason: 4.0.1 through 4.0.3 were added by Amendment No. 97 without any Bases. Bases are from Generic Letter 87-09.

1 53. Change _Q11CIlption: Specification 4.1 Bases, page 4-2: Add the word "are" to first line of second paragraph; add reference to the heat balance calibration requirement when Nuclear Instrumentation indicated power is 10*l,or more above core thermal power.

Reason: Editorial clarification. Amendment No. 78 appropriately changed the Bases to address the 101. calibration requirement. Amendment No. 87 inadvertently changed the Bases for Calibration back to the pre-Amendment No. 78 version. The Calibration Bases as approved in Amendment No. 78 is placed back into the Tech Specs.

, 54. Change _DestdpilDD: Table 4.1 ), Item 29, page 4-7: Add the wording "Equivalent to Item 49, but excludes SPDS" to the Remarks column.

i BeasoD: The added wording makes clear that the daily check is not j redundant with Item 49 which is checked monthly.

55. Change _01stdption: Table 4.1-1, Item 39, page 4-7a: Add calibration per refueling interval for control rod relative position.

Reason: Control rod absolute and relative positions are checked, one compared to the other, once per shif t. Since the control rod absolute

. position is calibrated each refueling interval, it is appropriate for control rod relative position to be calibrated at the same time.

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. Facility Change Safety Analysis Log No. 1076 Proposed Amendment No. 171 Page 7 of 13

56. Chanae

Description:

Table 4.1-1, page 4-7b: Delete Item 45, Emergency plant radiation instruments.

Reason: Item 45. Emergency plant radiation instruments, was deleted because this equipment (hand-held radiation meters) does not fall under the applicability of Specification 4.1, which states: ,

"Applies to items directly related to safety limits and limiting conditions for operation during power operation. During cold shutdown, systems and components required to maintain safe shutdown will be tested."

57. Change _Ducriotion: Table 4.1-1, page 4-7b: Hove 44.c to 46.b. Item 44.d is relabeled 44.c.

ReuQD: Chlorine detector is more appropriate with environmental monitors than radiation monitors.

58. Chance DelCdpilDn: Table 4.1-1, Item 63, page 4-7c: Add note "(1)

Source check only" to Remarks and "(1)" to Test column.

Reason: Clearly identify monthly testing for these monitors.

59. Change _

Description:

Table 4.1-1, pages 4-7d and 4-7e: Add remark for Items 68.c. 68.h, 71.b, 71.c, and 71.d that pressure transmitters are to be calibrated annually.

Reason: The results of design engineering calculations indicate that these EFIC pressure transmitters, located outside of containment, require annual calibration rather than on a refueling interval as specified for the remainder of each of these channels.

60. Change _0_eist.ipilon: Table 4.1-1, page 4-79: Change the Channel Description of Item 85 from "Hide Range OSTG Pressure Pressurizer Level" to "Hide Range OTSG Pressure." Insert new Item 86 with the Channel Description "Pressurizer Level," and renumber existing Items 87 through

90. The Check, Test and Calibrate notations are the same for both Item 85 and 86.

Reassa: Combining the two Channel Description items was a typographical error.

61. Change _ Duct _lplion: Specification 4.2.2 Bases, page 4-12: In second paragraph, change "4.2.8" to "4.2.1.2."

Reason: For editorial purposes, Specification 4.2.8 was moved to Specification 4.2.1 by Amendment No. 76 (9/30/85). Through oversight, the corresponding number change was not made in the Bases, and is herein proposed for correction.

62. Change _Descdption: Specification 4.4.1.1.5, page 4-16a: Add specific !

requirement for reporting any deterioration noted in the containment i surfaces inspection performed before each integrated leak rate test. l Reason: Clarification is provided that reporting is required by 10 CFR ,

50, Appendix J,Section V.B. and "Special Report" Specification 6.9.SH. i 1

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. Facility Change Safety Analysis Log No. 1076 :

Proposed Amendment No. 171 Page 8 of 13 I

63. Change _D11ntigtlon: Specification 4.4.1.4, page 4-19: Change title of specification from "Annual Inspection" to "Containment Surfaces ;

Inspection." Delete the requirement for an annual inspection. Also,

! change the term "integrated leak test" to "integrated Icak rate test,"

and reference "Special Report" Specification 6.9.5M. In addition, delete all but the first and last sentences from the second paragraph of the ,

! Bases on page 4-19, and move the remainder of second paragraph of Bases l from page 4-20 to page 4-19. <

. Benon: Only an inspection prior to an integrated leak rate test and not an annual inspection of the containment surfaces is required by 10 CFR 50, App. J Section V.A. The change in test terminology is editorial, and addition of the special report reference is for I clarification. The Bases statement that periodic tests are to be performed without preliminary leak detection surveys or leak repairs, etc., is deleted as an option allowed by IE Notice 85-71, and 10 CFR 50,

Appendix J. The remaining deleted sentences from the Bases concern

reduced pressure testing, an option which is not performed at Rancho Seco.

64. Change _Dessdation: Specification 4.4.1.4, page 4-20: Change date of Federo! Register reference in last line from "February 23" to "February 14." l Reason: Editorial correction. ,

65. Change _De_scdttion: Specification 4.4.2.5, page 4-23: Delete that

part of the first sentence calling for report preparation and review by 1 the District's Generation Engineering Department, and change "Specification 6.9-1" to "Specification 6.9.58." Also delete -

1 Specifications 4.4.2.6, 4.4.2.6.2, 4.4.2.6.3 and 4.4.2.6.4.

Realon: Reporting responsibilities are established by administrative procedures, and thus need not be made a Technical Specification requirement. The number 6.9-1 is an incorrect designation for a l specification. Specification 6.9.1 Jiscusses general reporting requirements. Specification 6.9.5B is specific to the type of report being discussed in Specification 4.4.2.5. The Liner Plate Surveillance specified in 4.4.2.6, 4.4.2.6.2, 4.4.2.6.3 and 4.4.2.6.4 was only required af ter the first year of initial operation and is no longer applicable.

66. Change _hscdntion: Specification 4.5.1.2B, page 4-28: Delete last i sentence of specification.

Reason: The requirement for verification of locked valves was deleted i

from ASME Code Section XI. Specification is consistent with standard Tech Specs for flow path verification.

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67. ChaDge_Desctlption: Specification 4.5.3.1, page 4-32, and Specification 4.5.3 Bases, page 4-33: Change the combined maximum allowable leakage from components in the Decay Heat Removal System and 1, Reactor Building Spray System from 6.0 gallons per hour (gph) to 0.63 gph.

i Reason: Amendment No. 57 (10/30/84) approved increasing the combined 1 maximum allowable leakage from components in the Decay Heat Removal l System and the Reactor Building Spray System from the original Technical l

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. Facility Change Safety Analysis Log No. 1076 Proposed Amendment No. 171 Page 9 of 13 i

Specification value of 0.63 gph to 6.0 gph. Current radiation dose !

calculations for the Control Room /TSC give acceptable results using 0.63 !

gph leakage, but give unacceptable results using 6.0 gph. Because '

, surveillance experience has found this combined maximum leakage to be

! well below the 0.63 gph value, and in order to achieve acceptable dose ;

calculations for the Control Room /TSC, the District proposes this ;

conservative reduction to 0.63 gph for the combined maximum allowable leakage.

68. Change Descr1 P ilon: Specification 4.5.3.2.B2, page 4-33: Change the ,

term "pt.mp isolation" to "pump suction isolation." i ReAion: The proposed wording change corrects the term, and makes it f

consistent with the wording in Specification 4.5.3.2.81. i i

)' 69. Change

Description:

Specification 4.9, page 4-40: Change the wording ;

"Atomic Energy Commission" to "Nuclear Regulatory Commission."

P.eaton: Title correction. {

1 70. Change _Daicdplion: Specification 4.11.10.1(b), page 4-42a: Add a l comma after "adsorber tray" and "thoroughly." .

j BeAion: Editorial clarification. l I 71. Change

Description:

Specification 4.11.10.2. page 4-42a and 4-43b: I Change the designation "D.1" to "Specification 4.11.10.1," and make a '

, similar change to Specification 4.12.20.2 on page 4-43b. !

Reason: This change is proposed for clarification. l t

72. ChangeJ9.scdplion: Specifications 4.12.28.1, 4.12.28.2, 4.12.28.3, l

, and 4.12.20.2, pages 4-43, 4-43a, and 4-43b: Change the qualifler for '

the flow rates shown as "not exceeding cfm r 10 percent" to "flow rate of cfm i 10 percent."

EcAton: The use of the words "not exceeding" when referring to a i

numerical parameter with a specified tolerance is technically incorrect. !

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73. Change _Deltdpilon: Specifications 4.12.20.1(a) and (b), page 4-43a: '

i 1 Add comma after "thoroughly."

) Reason: Editorial, and for consistency with Change Description No. 70. ,

74. Chauce_Destdpilon: Specification 4.13A.2, Note page 4-45: Change j "exmained" to "examined," and in Specification 4.13C.2, change "ASME l Section XI Code" to "ASME Code,Section XI." !

Reason: Typographical and editorial corrections. l l

75. Change _DesCdption: Table 4.1-1, page 4-47d: Change title to same as i 1 title in index. j Reason: Title on table and index are different. ,

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. Facility Change Safety Analysis Log No. 1076 Proposed Amendment No. 171 Page 10 of 13

76. Change _Deltdplion: Specification 4.17.2, page 4-51: Change "Specification 4.17-3" to "Specification 4.17.3."

Reason: The numerical designation with a hyphen is the prope- format for identifying a figure or a table. The corrected designatioi, is proper for referring to the Specification.

77. Change _D21rdatioD: Specification 4.17.2.d page 4-52: Change "Specification 4.17.4.5" to "Specification 4.57.4a.5."

Reason: Specification 4.17.4a.5 is the proper designation.

78. Change _Deitdpilon: Table 4.17-3, pages 4-57b and 4-57c: Change the tube number in Row 46 from "121" to "119."

Ernon: Tube No. 119 is the correct number of the Special Interest Tube in Row 46.

79. Change _Deltdation: Specification 4.18.1.2, page 4-58: Change "NFPA Standard 720" to "NFPA Standard 720."

Reason: Typographical cc,rrection.

80. Change _Qe1Cdation: Specification 4.18.2.1, pages 4-58 and 4-59:

Delete subparagraph f.1 which states "Verifying that each automatic valve in the flow path actuates to its correct position." Also renumber subparagraphs f.2, f.3 and f 4 accordingly.

~

Reai?o: The only automatic valves in the fire suppression water system are surveillance tested by Specification 4.18.3.1c.1; thus, Specification 4.18.2.1f.1 is deleted because of redundancy.

81. ChaDge_Dessdation: Specification 4.18.2.1h.2, page 4-60: Change "ASTH-D975-74" to "ASTH 0975-78."

Benon: The chance represents the updated version of the ASTM method used at Rancho Sew for diesel oil analysis.

82. Change _DeindatioD: Table 4.20-1, page 4-67: Add reference to Note (6) to Item 3.d., System Effluent Flow Rate Device, Channel Test.

Renon: Inadvertent omission from previous amendment.

83. Change _Destdpilon: Bases for Specification 4.22.4, page 4-79: Add "these" before systems in second sentence.

Realon: Inadvertent omission of word.

84. Change _2c3cr1Etion: Specifications 5.1.1, 5.1.2, 5.1.3, and 5.1.4, page '-1: Change "shall bed to "is."

Renon: Correct wording.

85. Cha' ige _Dettdation: Specification 6.2.2, page 6-1: Change wording in Item c. from "...when the unit is in other than cold shutdown or refueling" to "...for all reactor operational modes from HEATUP-COOLDOHN through POWER OPERATION"; change number of fire brigade members not to be included from minimem shif t crew from "3" to "5."

Reason: Rewording in Item c. more clearly defines operating paramt .s; change in number of fire brigade members ensures no degradation of shift crew necessary to safely shut down the plant or perforo, other essential functions.

l . Facility Change Safety Analysis Log No. 1076 Proposed Amendment No. 171 Page 11 of 13 r

86. Change _Descrignon: Specification 6.2.2, page 6-la: Add the word i "key" in front of radiation protection personnel in first paragraph of i Item g. i Reason: Clarifies and qualifies reference to designated personnel. !

87. Change _

Description:

Specification 6.4.1, page 6-3: Delete reference ;

to Appendix A of IC CFR 55; change qualification references in j

Specifications 6.3.1 and 6.4.1.

Reason: Appendix A of 10 CFR 55 has been incorporated directly into 10 CFR 55. Changes in qualification references are editorial updates. ;

88. Change _DeicdpMob: Specification 6.5.1.6m, page 6-5: Change "Nuclear i

Plant Manager" to "AGH, Nuclear Power Production."

j Reason: Correct position title.

1 89. Change _Deicdelion: Specification 6.5.3, page 6-9: Delete "but shall l be no lower than a division director level" from end of Items b. and c.

Reason: Unnecessary restrictional guideline.

90. Change _Descdstlan: Specification 6.5.4, page 6-10: Add "at least i l once per 12 months" to Item n. ,

j

Reason: Consistent with standard RETS. l l

91. Change _Qticdp11on: Specification 6.8.1(a), page 6-11: Change i 1 "Regulatory Guide 1.33" to "Safety Guide 33." ;

' "Regulatory Guide 1.33, November 1972" should be "safety Guide Reason: ,

33, November 1972." ;

I 92. Change _DeicdaMon: Specification 6.9.2.2.2, page 6-12a: Add i reference to Specification 4.27.

a Reason: Informational enhancement.

93. Change _Descdotion: Specification 6.9.2.2.2, page 6-12b: Add l

< parenthetical reference to Specification 4.30. l j Reason: Informational enhancement.

l 94. Change _Dettdphon: Specification 6.9.2.3.1, page 6-12c: Add l j "Semiannual" to first sentence, third paragraph; change "Radwater" to l "Radwaste" in Gurth paragraph; change "3.20" to "3.18.5" at end of sixth :

+

paragraph.

1 Reason: Editorial / typographical corrections; referenced Specificatior, j 3.20 was deleted in Amendment No. 98.

95. Change _Destdation: Specification 6.9.58 C, page 6-129: Revise l Specification 6.9.5B to make it conform with the reporting requirements :

as stated in Technical Specification Section 4.4.2. Add reference for ,

i item 6.9.5C. t

" Reason: An annual structural integrity inspection is not required v ,

i 10 CFR 50, Appendix J only an inspection prior to an integrated n L

, rate test. A report of the structural integrity inspection is r m u ed .

1 to be included in the Integrated Leak Rate Test Special Report, M only 6

! if the inspection reveals abnormalities. f i

r d

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

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. Facility Change Safety Analysis Log No. 1076 Proposed Amendment No. 171 Page 12 of 13 :

The Liner Plate Surveillance was required before and after the initial pressure test, and within approximately 1 year after that. This surveillance was discontinued after the first year per Specification !

4.4.2.6.4 and therefore is deleted.

96. Changt_Deliriation: Specification 6.9.5E, page 6-12g: Delete "3.14.2.3" and add "14 days (3.14.2.3)." i EtAinn: Specification 3.14.2.3 requires a 14 day report,

97. Change _Qtstriation: Specification 6.9.5F, page 6-129: Add "30 days (3.13.3)." :

Reason: Add reference specification.

98. Change Descr1Rilon: Specification 6.9.5L, page 6-12h: Add "Hetearological Instrumentation, 10 days (3.29)."

Reason: Add report required by Specification 3.29.

99. Change _Deistinilon: Specification 6.9.5H, page 6-12h: Delete "Solid Radioactive Hastes, 30 days (3.21)" and add "Integrated Leak Rate Test, ;

90 days (4.4.1.1.5)." l Reason: Specification 3.21 does not have any reporting requirements.

Add reporting requirements of Specification 4.4.1.1.5.

100. ChangtJ eictintjon: Specification 6.9.50, page 6-12h: Add "(and REMP Monitoring Point Locations)."

Reason: Identify additional reporting requirement in Specification 3.23.

101. Change _Desct]Rilon: Specification 6.10.2a, page 6-13: Change "Final I Safety Analysis Repert" to "Updated Safety Analysis Report."

Reason: The reason for the change is as explained in Change Description No. 14. t t

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.- Facility Change Safety Analysis Log No. 1076 Prv osed Amendment No. 171 Page 13 of 13 NO SIGNIFICANT HAZARDS CONSIDERATION The proposed changes have been evaluated by the District using the standards of 10 CFR 50.92. The conclusion has been reached that the changes constitute no significant hazard to the public. The following is the basis for the District's conclusion:

The changes are purely administrative, consisting of editorial corrections, renumbering and relocating of specifications, updating references to codes and standards, and specification clarifications. The proposed administrative changes have been judged to not constitute a significant hazards consideration because the changes:

a. Would not involve a significant increase in the probability or >

consequences of an accident previously evaluated since the proposed -

administrative changes do not affect existing plant designs or cause )

changes to existing plant operations. !

b. Would not create the possibility of a new or different kind of !

accident from any accident previously evaluated since the proposed administrative changes do not introduce any new operational requirements that could affect plant safety, i

c. Would not involve a significant reduction in the margin of safety l since the purely administrative changes could not cause a reduction in i the conservative nature of the Technical Specifications due to the editorial nature of the proposed changes. ;

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